Symposium Organizers
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Marjorie Bertolus CEA, DEN, DEC/SESC/LLCC
Blas Uberuaga Los Alamos National Laboratory
Robin W. Grimes Imperial College London
A1: Nuclear Fuels I
Session Chairs
Robin Grimes
Karl Whittle
Monday PM, November 28, 2011
Independence W (Sheraton)
9:30 AM - **A1.1
Radiation Resistance of UO2 under Severe Damaging Conditions.
Thierry Wiss 1 , Arne Janssen 1 , Hartmut Thiele 1 , Bert Cremer 1 , Jean-Yves Colle 1 , Dragos Staicu 1 , Vincenzo Rondinella 1 , Rudy Konings 1
1 , European Commission - JRC - ITU, Karlsruhe Germany
Show Abstract The most commonly used nuclear fuel, UO2, is subjected to radiation damage not only during its in-pile irradiation, but also during cooling and storage. Magnitude, rate and conditions of the damage accumulation are different for reactor irradiation and for (long time) storage conditions, but to some extent the damage pattern is very similar. During irradiation in nuclear reactor, each atom in the fuel experiences several thousand displacements from its initial lattice position. A large amount of energy, mainly generated by the fission, is dissipated in the lattice and causes the formation of defects. Driven by power and temperature gradients and as a consequence of radiation damage the properties of the fuel change significantly with increasing burnup. Defects generated in the fuel structure (point and extended defects, micro- and macro-bubbles, solute and segregating impurities) will alter key properties, like e.g. thermal conductivity, density and mechanical properties, which determine the performance and ultimately the safety of the fuel. Future reactor concepts envisage the use of fuel (and materials) up to higher burnup and more severe irradiation conditions; moreoverthey are often characterized by higher Pu- and minor actinide-content, which results in higher alpha-decay damage extent. The fuel after irradiation and during storage is still very radioactive. The long timescale considered for storage in many countries requires understanding of the damage mechanisms and developing suitable tools to predict the fuel evolution. Properties relevant for safe handling/processing of high specific alpha-activity fuels are strongly affected by the build-up of alpha-decay damage and helium. This is the object of a campaign of studies carried out at JRC-ITU, which covers in particular the evolution of thermal transport and mechanical properties as a function of accumulated radiation/decay damage and He.In order to simulate alpha-damage accumulation in UO2 spent fuels aged for periods corresponding up to a few thousand years, samples doped with short-lived alpha-emitters (e.g. 238Pu) have been fabricated and characterized. The alpha-damage accumulation affects many properties of UO2 like thermal diffusivity, lattice parameter, heat capacity, showing a rapidly saturating behaviour. Irradiated fuels have been characterized by different techniques including transmission electron microscopy, X-ray diffractometry, in combination with thermal annealing methods. Comparative analysis of spent fuel and alpha-doped materials allows assessing superimposition of alpha-decay effects after fuel discharge onto radiation damage occurred in-pile. It was shown that the damaged microstructure of irradiated fuel and of UO2 doped with alpha emitters is very similar despite the large difference in the conditions under which the damage occurred. UO2 shows a remarkable ability to maintain its original fluorite structure even under severe irradiation conditions.
10:00 AM - A1.2
Sesquioxide Effect on Thermal Diffusion Processes in UO2.
Simon Middleburgh 1 2 , Robin Grimes 1 , Paul Blair 2 , Karin Oldberg 2
1 Department of Materials, Imperial College London, London United Kingdom, 2 Materials and Fuel Rod Design, Westinghouse Electric Sweden, Vasteras Sweden
Show AbstractThe effects of trivalent cation solution on fission gas release has been studied by calculation of a number of transition states for diffusion processes within UO2. Reduction in the energy required for a vacancy migration to take place has been observed with solution of all trivalent cations, the larger reductions occurring with the smaller cations aluminium and chromium, both suggested fuel dopants. A qualitative comparison of the diffusion co-efficient for chromium doped fuel with undoped fuel has been made, which suggests that higher Cr concentrations will be associated with higher xenon diffusivity (due to enhanced vacancy migration).
10:15 AM - A1.3
Effect of Lanthanide and Actinide Substitution in UO2 Using Atomic Level Simulations.
Rakesh Behera 1 , Chaitanya Deo 1
1 Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering , Georgia Institute of Technology, Atlanta, Georgia, United States
Show AbstractUranium-based fuels are the most common fuel used for commercial nuclear energy generation. The complete fuel cycle based on UO2 fuels generates a large number of transuranic nuclides (Pu, Am, Np, Cm). These fission products influence a variety of properties. While the nuclear fuel cycle is well characterized, the understanding of the physical and chemical properties of the actinides is still limited. This study focuses on characterizing the effect of dilute concentrations of Lanthanides and Actinides on bulk properties of UO2. In particular, the results will include the effect of elastic and electrostatic effects due to the substitution of +4e- (Am, Pu, Ce, Np, U, Th) and +3e- (Gd, Eu, Sm, Am, Nd, Pu, U) ions in the UO2 lattice. The discussions will be based on the experimentally observed concentrations of Lanthanides and Actinides in urania using atomic level simulations.
10:30 AM - A1.4
Effect of Cr Segregation to UO2 Grain Boundaries.
Minki Hong 1 , Simon Phillpot 1 , Blas Uberuaga 2 , Chris Stanek 2 , Susan Sinnott 1
1 MSE, University of Florida, Gainesville, Florida, United States, 2 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThe UO2 fuel pellet has a polycrystalline microstructure and the density and the size of each grain are the key to control the fuel performance particularly by modifying its thermal conductivity. A significant amount of research has been conducted to improve these properties by doping sintering additives and Cr has been suggested as one of the elements that have the capability of grain enlarging especially during the sintering process of UO2 pellet. However the mechanism of the grain enlarging and the effect of Cr on grain boundary behavior under actual operating condition are not well understood. Here, atomic-level simulation methods using empirical interatomic potentials are used to examine segregation of Cr to UO2 grain boundaries and understand its grain enlarging mechanism. In addition, the quantitative energetics of Cr near the grain boundary and its chemical or bonding environment are examined using density functional theory calculations with the Hubbard U approximation. The results indicate that Cr is mostly insoluble in UO2 unless it substitutes uranium under hyper-stoichiometric condition and the segregation energy of Cr to the Σ5 tilt boundary with (310)/(001) plane is about 2.8 eV.
10:45 AM - A1.5
Crack Tip Plasticity in Single Crystal UO2: Atomistic Study.
Yongfeng Zhang 1 , Xiangyang Liu 2 , Bulent Biner 1 , Paul Millett 1 , Michael Tonks 1 , David Andersson 2
1 Fuel Modeling and Simulation, Idaho National Lab, Idaho Falls , Idaho, United States, 2 Structure/Property Relations, Los Alamos National Lab., Los Alamos, New Mexico, United States
Show AbstractThe room temperature fracture behavior of single crystal UO2 is studied using molecular dynamics (MD) simulations with the Basak potential. The cracks are introduced on two low-index charge neutral planes, the (111) and (110), and the mode-I loading is applied normal to the crack planes. At the onset of growth of the cracks, plastic deformations such as dislocation emission and phase transformations are observed at the crack tips. The dislocations are characterized as ½<110> full dislocation gliding on the (001) plane. Two metastable phases are identified as Rutile and Scrutinyite structures, and their formation is confirmed by separate density-functional-theory calculations. The cracks residing on the (111) plane propagate along the high-energy incoherent boundaries between the ground Fluorite and the newly formed metastable phases. In the case of cracks located on the (110) plane, the new phases form coherent boundaries. As a result, the stress at the crack tips is largely reduced; and no crack extension is observed.
11:30 AM - A1.6
Multiscale Fuel Performance Simulation of Metallic Reactor Fuels.
Michael Tonks 1 , Paul Millett 1 , Bulent Biner 1
1 Fuels Modeling and Simulation, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractMetallic fuel is a popular option for Generation IV nuclear reactors. However, the nuclear industry lacks the years of operational experience with metallic fuel that they have with UO2 fuels. Thus, an accurate science-based model of metal fuel performance could be a powerful tool for investigating metal fuel performance in typical and accident conditions. A predictive fuel performance model must account for microstructure evolution, and to develop such a model requires input at the atomistic, meso- and engineering-scales. In this research, atomistic simulation is used to develop important parameters, such as point defect mobilities, and to identify critical mechanisms. Mesoscale phase field models then use this information to predict microstructure evolution due to external conditions, such as loading and radiation damage. The mesoscale model then determines the effect of the microstructure evolution on various bulk material properties, including thermal conductivity and density. These mesoscale-informed properties are used in the engineering-scale fuel performance simulation to predict the thermal and mechanical behavior of metallic fuel during its lifetime in the reactor.
11:45 AM - A1.7
Chemical Behavior of Oxide Nuclear Fuel: Recycle and High Burn-up.
Theodore Besmann 1 , Stewart Voit 1 , Dongwon Shin 1 , Evan Noon 1 , Robert Austin 1
1 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractThermochemical models of oxide nuclear fuel systems containing transuranic and fission product elements are being developed. Specifically, subsystems of major actinides with fission products are being represented by solid solution models such as the subregular model for the 5-metal white phase and the compound energy formalism sublattice approach for variable stoichiometric oxides such as the fluorite-structure fuel phase. Current work has emphasized the behavior of actinides with rare earths as these are important for both fuel recycle where rare earth elements in significant concentrations remain with the actinides, and in-reactor where they influence stoichiometry and oxygen potential. This report will discuss recent experimental and modeling efforts related to the behavior of rare earths in the fuel phase, and the overall complexity and importance of oxygen behavior in fuel.This work was supported by the US Department of Energy Office of Nuclear Energy, Fuel Cycle Research and Development Program.
12:00 PM - A1.8
Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications.
Yang Zhong 1 2 , Robert O'Brien 1 , Steve Howe 1 , Nathan Jerred 1 , Kristopher Schwinn 1 , Amy Kaczmarowski 1 , Joshua Hundley 1 , Laura Sudderth 1
1 , Center for Space Nuclear Research, Idaho National Lab, Idaho Falls, Idaho, United States, 2 Department of Chemical, Materials and Biomolecular Engineering, University of Connecticut, Storrs, Connecticut, United States
Show AbstractThe recent events with the reactors in Fukishima, Japan revealed a need for a high temperature fuel form that will not melt down from decay heat after a loss of coolant accident. Furthermore, the fuel material should contain the fission products from dispersion during a combination of accidental high temperature excursions and steam/hydrogen explosions. Such requirements will necessitate a new robust fuel encapsulation matrix. The Center for Space Nuclear Research has been developing a new fuel form (fuel cermets) for nuclear reactors to be used for space exploration. Fuel Cermets consist of a tungsten-rhenium (W/Re) encapsulating matrix and a ceramic compound (a nuclear fuel such as uranium in its oxide form). Owing to the good thermal conductivity, mechanical strength, hardness, and high melting point of W/Re alloys, as well as their ability to contain fission products, tungsten cermet fuels are highly attractive for applications where enhanced nuclear reactor safety and proliferation resistance is essential. In this study, an analysis of cermet fuels produced via Spark Plasma Sintering (SPS) is provided. SPS processing can greatly reduce the average sintering temperature and minimize the grain growth during production in comparison to traditional sintering techniques. In the examples presented, CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar kinetic properties of these materials, in particular their respective melting points and Gibbs free energies. The densification of the cermets with respect to the volumetric ratios between metal and ceramic, the grain size and the morphology of the starting powder is systematically studied in this work. The sintering kinetics is also investigated using master sintering curve.
12:15 PM - A1.9
Effect of Intergranular Gas Bubbles on Thermal Conductivity.
Karthik Chockalingam 1 , Paul Millett 1 , Michael Tonks 1 , Bulent Biner 1 , Liangzhe Zhang 1 , Yangfeng Zhang 1
1 Nuclear Fuels and Materials, Idaho National Laboratory, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractModel microstructures obtained from phase-field simulations are used to study the effective heat transfer across systems with stationary grain boundary bubble populations. From the analysis it is found that grain boundary bubbles represent a larger impediment to thermal transfer than traditional ‘rule-of-mixture’ theories predict. Additionally, we find that the grain boundary coverage, irrespective of the intergranular bubble radii, is the most relevant parameter to the actually thermal resistance, which we use to derive ‘effective’ Kapitza resistances. Models have been proposed to predict thermal conductivity as a function of porosity, grain size and grain boundary bubble coverage.
12:30 PM - A1.10
Phase-Field Modeling of Pore Migration in Nuclear Fuels Due to a Temperature Gradient.
Liangzhe Zhang 1 , Michael Tonks 1 , Paul Millett 1 , Bulent Biner 1 , Yongfeng Zhang 1 , Karthikeyan Chockalingam 1
1 Fuels Modeling and Simulation, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractSintered UO2 nuclear fuel materials undergo a unique microstructural evolution process during the course of the burn-ups. The evolved microstructure is usually characterized by the columnar grains surrounding a large central void, which mainly results from the migration of the initial pores towards the high temperature regions. A quantitative description of the pore migration process is therefore desirable for better understanding and accurate predictions of the fuel performance. For this purpose, a phase-field model is developed; in which the kinetics of the migration due to both bulk and surface diffusion is formulated by utilizing fourth order Cahn-Hillard (CH) equations. The results indicate that the porosities migrate towards the high temperature region owing to the temperature gradient as the driving force, which are consistent with the experimental observations. Furthermore, it is also seen that a pore can also changes its shape due to the small variations of temperature profile at its surrounding regions.
12:45 PM - A1.11
HIP Bonded U-10Mo Monolithic Fuel Plates with a Modified Fuel to Cladding Interface.
Jan-Fong Jue 1 , Blair Park 2 , Cynthia Breckenridge 1 , Jeffery Hess 1 , Glenn Moore 2 , Dennis Keiser 1 , Daniel Wachs 1
1 Fuel Performance & Design, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Fuel Fabrication, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractUnder the RERTR (reduced enrichment for research and test reactors) program, the nuclear fuels used in research and test reactors are being converted from highly enriched to proliferation-resistant low-enriched uranium (LEU), defined as less than 20% U235 enrichment. One of the fuel designs under development is a monolithic fuel type where the fuel is in the form of a single U-10Mo (uranium - 10 wt% molybdenum) alloy foil. With monolithic fuel, a uranium fuel density of more than 10 g/cm3 can be achieved. The post irradiation results from previous RERTR irradiation experiments indicate that the addition of silicon to the fuel-to-cladding interface resulted in reduced fuel/matrix chemical interaction and increased the stability of the interaction layer during irradiation. This paper provides an update of the developmental effort on the fabrication of monolithic fuel type by the HIP (hot isostatic press) bonding process with a silicon enriched fuel/cladding interface.
A2: Radiation Damage - Ceramics
Session Chairs
Jonathan Hinks
Christopher Stanek
Monday PM, November 28, 2011
Independence W (Sheraton)
2:30 PM - A2.1
Self-Healing Response of Ionic Crystals to Irradiation: Can Damage be Good?
Dilpuneet Aidhy 1 , Dieter Wolf 1
1 , ANL, Argonne, Illinois, United States
Show AbstractMolecular dynamics simulations of irradiated CeO2 (often considered a surrogate for UO2, the most widely used nuclear fuel) reveal the formation of charge-neutral interstitial dislocation loops identical to ones observed recently in experiments. Focusing on the kinetic phase that follows the initial damage cascade, our simulations of the cluster-formation mechanism reveal a self-healing response of the perfect crystal to the radiation-induced defects. Remarkably, the lattice responds to point defects created during irradiation with the spontaneous creation of new point defects. We demonstrate that these new ‘structural defects’, with a negative energy of formation, neutralize the cluster by screening its long-range Coulomb potential, thereby lowering the overall energy and localizing the damage. A similar lattice response was recently identified also in simulations of MgO, although very different types of clusters were formed, suggesting that this self-healing screening response may be an intrinsic reaction of all ionic crystals to irradiation.
2:45 PM - A2.2
Dynamic Recovery in Silicate Apatite Structures under Irradiation and Implications for Long-Term Performance Modeling.
William Weber 1 2 , Yanwen Zhang 2 1 , Haiyan Xiao 1 , Lumin Wang 3
1 Materials Science & Engineering, University of Tennesee, Knoxville, Tennessee, United States, 2 Materials Science & Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 3 Nuclear Engineering & Radiological Sciences, The University of Michigan, Ann Arbor, Michigan, United States
Show AbstractThe irradiation responses of Ca2La8(SiO4)6O2 and Sr2Nd8(SiO4)6O2 with the apatite structure are investigated to predict their long-term behavior as host phases for immobilization of actinide elements from the nuclear fuel cycle. Different ions and energies are used to study the effects of dose, temperature, atomic displacement rate and ionization rate on irradiation-induced amorphization and recrystallization. The dose for amorphization increases with temperature in two stages, below and above 150 K. In the high temperature stage relevant to actinide immobilization, the increase of amorphization dose with temperature exhibits a strong dependence on the ratio of ionization rate to displacement rate for the different ions. Data analysis using a dynamic model for amorphization reveals that ionization-induced processes, with activation energy of 0.15 ± 0.02 eV, dominate dynamic recovery for ions from Ne through Xe. For heavier Au ions or for alpha-recoil nuclei emitted in alpha decay of actinides, ionization becomes less dominant and dynamic recovery is controlled primarily by thermally-driven processes. In post-irradiation annealing studies of amorphous samples, epitaxial thermal recrystallization is observed at 1123 K, and irradiation-enhanced nucleation of nanocrystallites is observed in situ under irradiation with heavier ions. The recrystallization temperature under irradiation decreases with increasing ion mass to a value of ~ 823 K, which also defines the thermally-driven critical temperature for amorphization under irradiation with heavy ions. Some partial recovery due to alpha particle irradiation at 300 K is observed that suggests a self-healing mechanism in apatite phases containing actinides. Based on the results and dynamic model, the temperature and time dependence of amorphization in apatite host phases for actinide immobilization are predicted.
3:00 PM - A2.3
Defects, Minor Phases and Microstructures in O+ and Zr+ Ion Co-implanted Strontium Titanate: A Model Nuclear Waste Form.
Weilin Jiang 1 , Mark Bowden 1 , Zihua Zhu 1 , Libor Kovarik 1 , Bruce Arey 1 , Przemyslaw Jozwik 2 3 , Jacek Jagielski 2 3 , Anna Stonert 3
1 , Pacific Northwest National Laboratory, Richland, Washington, United States, 2 , Institute of Electronic Materials Technology, Warsaw Poland, 3 , The Andrzej Soltan Institute for Nuclear Studies, Otwock Poland
Show AbstractOperations in the nuclear power industry generate spent fuels that contain highly radioactive materials. The fission products of 235U and 239Pu have a high percentage of 90Sr and 137Cs isotopes in the nuclear waste stream from reprocessing of the spent fuels. Long-term storage or permanent disposal of nuclear wastes requires stabilization of the highly radioactive materials in a solid form that degrades very slowly over time, thereby limiting the radionuclide release to the environment. Discovery and development of such forms for immobilization of nuclear wastes are critical for proper management of existing and future spent fuels.Single-crystal strontium titanate (SrTiO3 or STO) is used in this study as a model material to simulate a waste form for disposal of radionuclide 90Sr that decays to 90Y and subsequently to 90Zr. The self-irradiation from decay electrons, charge state change from 90Sr2+ to 90Zr4+ and substantial heat production can significantly affect the structural stability of the host material and potentially lead to phase transformation, phase separation, and/or formation of new phases. In this study, sequential implantation of 16O+ and 90Zr+ ions was performed for STO at 550 K to minimize the charge imbalance and to avoid full amorphization of STO. Each of the implant concentrations of up to 1.5 at% was achieved. Post thermal annealing was conducted in flowing Ar gas environments at temperatures up to 1423 K for 10 hours. Various experimental methods have been employed to characterize the implanted sample, including time-of-flight secondary-ion mass spectroscopy, multiaxial ion-channeling analysis, high-resolution transmission electron microscopy, and micro-beam x-ray diffraction. The results show that, in contrast to the observed mobile Sr interstitials in STO, the implanted Zr does not diffuse noticeably during the ion implantation or thermal annealing up to the highest applied temperature (1423 K) in this study. This behavior is attributed to the formation of strong chemical bonding of the implanted Zr in the structure. A defect concentration was generated in STO and nearly all the implanted Zr was not located exactly at the original lattice site. There are Zr-containing microstructures, precipitates and voids (or oxygen blisters) in the implanted layer. A minor phase with a tetragonal structure was also observed. It survived thermal annealing at temperatures up to 1423 K with only a small decrease in the lattice parameter. Discussion about the results and a general assessment of the model waste form will be provided in this presentation.
3:15 PM - A2.4
Post Irradiation Examination of Neutron Irradiated Inert Matrix Ceramics.
Donald Moore 1 , Cynthia Papesch 2 , Brandon Miller 2 , Pavel Medvedev 2 , Juan Nino 1
1 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractThere is an increasing radiotoxic inventory of nuclear waste requiring safe, ecologically friendly, and economically sensible disposal. A promising approach for reducing waste from spent nuclear fuel and weapons programs is by utilizing an inert matrix fuel (IMF) for the transmutation of waste in light water reactors (LWRs). Implementation of an IMF requires a stable and radiation tolerant inert matrix material with similar thermophysical and neutronic properties as UO2.Several potential inert matrix materials and other ceramic materials including MgO, Nd2Zr2O7, MgO-Nd2Zr2O7 cercer composite, MgAl2O4, MgO1.5Al2O3, and Mg2SnO4 pellets were irradiated in-pile of the Advanced Test Reactor at Idaho National Laboratory. The effects of irradiation temperatures (~350 and ~700°C) and fast neutron fluencies (~1x10^25 and ~2x10^25 n/m2) on the materials properties are currently being investigated. Post irradiation examination includes thermal diffusivity, scanning electron microscopy, and transmission electron microscopy. The radiation induced thermophysical and structural evolution of MgO and MgAl2O4 will be presented. We will discuss the effects of irradiation damage on the thermal diffusivity of MgO and MgAl2O4. Comparison of the different irradiation conditions verse non-irradiated samples gives details of how defects affect the thermal diffusivity.
3:30 PM - **A2.5
Ion Irradiation Induced Defects in Non-Metallics.
Philip Edmondson 1 , Fereydoon Namavar 2 , Robert Birtcher 3 , Jonathan Hinks 4 , Stephen Donnelly 4 , William Weber 5 1 , Yanwen Zhang 1 5
1 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 2 , University of Nebraska Medical Center, Omaha, Nebraska, United States, 3 Materials Science Division, Argonne National Laboratory, Argonne, Illinois, United States, 4 School of Computing and Engineering, University of Huddersfield, Huddersfield, West Yorkshire, United Kingdom, 5 Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee, United States
Show AbstractThe binary oxide ceramics CeO2 and ZrO2 are key engineering materials in nuclear systems; either as structural materials, as possible inert fuel matrices, or as nonradioactive surrogates in studies of nuclear fuel systems. Recently, the nanocrystalline phases of these materials have come under increased interest due to their enhanced properties, and the ability to tailor these properties with grain size. In the work presented here, thin films of both cubic ceria and zirconia have been irradiated with Au+ ions to doses up to 35 displacements per atom (dpa), over a range of temperatures from 160 to 400K. Subsequent examination was performed using a combination of Rutherford Backscattering spectroscopy (RBS), scanning transmission electron microscopy (S/TEM), atom probe tomography (APT) and x-ray diffraction (XRD). In both films, the cubic phase is retained despite the relatively high dpa levels. Grain growth was also observed in all cases and may be attributed to the production of high levels of defects near the grain boundaries. In the zirconia film, the XRD results showed a lattice variation during the irradiation – the extent of which was dependent on the irradiation temperature. This lattice deviation is attributed to the generation and saturation of oxygen vacancies of different charge states being formed during irradiation. The ceria film showed no such lattice variation, and no enhanced oxygen deficiency was observed under ion irradiation. Analysis of the symmetry of the grain boundaries (GBs) showed that the initial film was dominated by asymmetric GBs and that during the irradiation symmetric GBs begin to dominate, reaching a saturation at dpa values of ~5, indicating a reduction in energy of the film. Dark bands of contrast were also formed at the film/substrate interface. STEM-EDS results indicate that the band was an ion-beam induced, chemically-mixed Ce/Zr-Si phase. In addition, amorphization processes in elemental (Si) and ternary (CuInSe2 (CIS)) semiconductors, as studied by TEM during in situ ion irradiation will be discussed. It will be shown that the interstitial-vacancy pair may be the dominant defect formed in Si during ion irradiation. The CIS proved to be resistant to amorphization at temperatures above 200K under the irradiation conditions used. At and below 200 K, amorphization only occurred in the samples that were Cu deficient relative to perfectly stoichiometric CIS.
4:30 PM - A2.6
Pyrochlore-Fluorite Transitions in Y2Sn2-xZrxO7: Implications for Stability.
Massey de los Reyes 1 , Karl R Whittle 1 , Robert G Elliman 2 , Nestor J Zaluzec 3 , Sharon E Ashbrook 4 , Martin R Mitchell 4 , Gregory R Lumpkin 1
1 Materials Engineering, ANSTO, Sydney, New South Wales, Australia, 2 Department of Electronic Materials Engineering, Australian National University, Canberra, Australian Capital Territory, Australia, 3 Materials Science Division, Argonne National Laboratory, Chicago, Illinois, United States, 4 School of Chemistry, University of St. Andrews, St. Andrews, Fife, United Kingdom
Show AbstractThe Y2Sn2-xZrxO7 pyrochlore series undergoes a phase transformation from a cubic pyrochlore structure-type (Fd3m) to defect fluorite (Fm3m) actuated by the increase in Zr content, coupled with thermal annealing above 1500°C. The pyrochlore-fluorite transition is an important factor in determining how materials behave under irradiation whether as a waste form or other nuclear material (e.g., inert matrix fuel, transmutation target, oxygen dispersion strengthened ODS additives). X-ray diffraction analysis reveals the onset of a pyrochlore to defect fluorite transition at Y2Sn1.6Zr0.4O7 with the loss of long range ordering. This is confirmed further by selected area diffraction, illustrating shorter range ordering in the defect fluorite phase incommensurate with unit cell size. However, this transformation occurs at a much higher Zr content than that predicted by classical radius ratio models. The diffuse scattering features observed in electron diffraction patterns of defect fluorite phases indicate some form of longer range ordering involving compositional-displacive structural modulation. The behavior of these materials during irradiation will be discussed and linked with the observed structural parameters (diffuse scattering, unit cell size).
4:45 PM - A2.7
The Role of Sn, Zr and Hf in the Radiation Damage in II,III, IV, and V Pyrochlores.
Karl Whittle 1 , Massey de los Reyes 1 , Yan Gao 1 , Mark Blackford 1 , Nestor Zaluzec 2 , Gregory Lumpkin 1
1 Materials Engineering, ANSTO, Sydney, New South Wales, Australia, 2 Materials Science Division, Argonne National Laboratory, Argonne, Illinois, United States
Show AbstractCeramics based on the general formulation CaLnZrNbO7 (Ln = La, Nd, Sm, Gd and Ho) have been studied as model four-component oxide systems, utilising the observation that they contain 2+, 3+, 4+ and 5+ cations. X-ray and neutron diffraction results show the materials to adopt pyrochlore across the series. Using a combined structural refinement the cation disorder across the A/B cation sites has been determined. The samples have subsequently been irradiated using the IVEM-TANDEM facility at Argonne National Laboratory (ANL), with 1 MeV Kr2+ ions at various temperatures. The results show a decrease in the critical temperature for amorphisation (Tc) from ~ 680 K for CaLaZrNbO7 to < 50K for CaHoZrNbO7. The change in Tc is discussed with references to disorder across cation sites, changes within the structure, and how these affect the radiation damage response. The results are also used to expand further the reliability of predicting those systems which are more tolerant of radiation damage, and how it can be used to develop new waste forms and other nuclear materials. Complementary systems containing Sn and Hf will also be discussed, in both how they respond and how they compare with CaLnZrNbO7 materials.
5:00 PM - A2.8
Structural Features in Fluorite Compounds Relevant for Nuclear Applications.
Gianguido Baldinozzi 1 2 , David Simeone 2 1 , Dominique Gosset 2 1 , Laurence Luneville 2 1 , Lionel Desgranges 3
1 SPMS, MFE, CNRS, Chatenay-Malabry France, 2 DEN, DANS, DMN, SRMA, MFE, CEA, Gif-sur-Yvette France, 3 DEN, DEC, SESC, CEA, St Paul lez Durance France
Show AbstractOxides with fluorite (or fluorite related) structures form a large class of compounds with a high radiation tolerance, somewhat related to their peculiar ability to accommodate a variety of defects and to form nonstoichiometric compounds with a large homogeneity range. Structural modifications are generally observed when the departure from the ideal composition is large. We would like to discuss these structural features using an approach based on the crystal symmetry analysis and to address some of the phase transition mechanisms in compounds relevant for nuclear applications.
5:15 PM - A2.9
Swift Heavy Ion Induced Amorphization of ZrO2.
Fengyuan Lu 1 , Maik Lang 2 , Jianwei Wang 2 , Fereydoon Namavar 3 , Christina Trautmann 4 , Rodney Ewing 2 , Jie Lian 1
1 Mechanical, Aerospace and Nuclear Engineering, RPI, Troy, New York, United States, 2 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States, 3 , University of Nebraska Medical Center, Omaha, Nebraska, United States, 4 , GSI Helmholtzzentrum Schwerionenforsch, Darmstadt Germany
Show AbstractZrO2 is an important engineering material as fuel matrix and nuclear waste forms, and the behavior of ZrO2 upon displacive and ionizing radiations is both scientifically and technologically crucial. Bulk monoclinic ZrO2 displays an excellent radiation tolerance and cannot be amorphized by displacive irradiation. However, we found that the extreme ionizing radiation by swift heavy ion of 1.33 GeV U-238 can induce amorphization of monoclinic ZrO2 with a grain size of ~ 50 nm. A similar amorphization trend was observed in 1.33 GeV U-238 irradiated tetragonal ZrO2. A computational simulation based on the thermal spike model demonstrates that with a very high electronic energy loss of 52.2 KeV/nm, the 1.33 GeV U-238 irradiation causes high transient temperatures in ZrO2 lattice beyond the melting point, resulting in the amorphization of the monoclinic ZrO2. An electronic energy loss threshold can be implied above which the radiation induced amorphization can occur in ZrO2. This work also highlights the potential of controlling ZrO2 phase by varying radiation conditions.
Symposium Organizers
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Marjorie Bertolus CEA, DEN, DEC/SESC/LLCC
Blas Uberuaga Los Alamos National Laboratory
Robin W. Grimes Imperial College London
A3: Nuclear Fuels II
Session Chairs
Majorie Bertolus
Thierry Wiss
Tuesday AM, November 29, 2011
Independence W (Sheraton)
9:15 AM - A3.1
Computational Modeling of Iso- and Aliovalently Doped ThO2 and UO2.
Vitali Alexandrov 1 2 , Niels Gronbech-Jensen 3 , Alexandra Navrotsky 1 4 , Mark Asta 2 1
1 Department of Chemical Engineering and Materials Science and NEAT ORU, University of California, Davis, Davis, California, United States, 2 Department of Materials Science and Engineering, University of California, Berkeley, Berkeley, California, United States, 3 Department of Applied Science, University of California, Davis, Davis, California, United States, 4 Peter A. Rock Thermochemistry Laboratory and NEAT ORU, University of California, Davis, Davis, California, United States
Show AbstractThe thermodynamic and kinetic properties of ThO2 and UO2 based solid solutions are of direct relevance for nuclear-fuel applications. In this talk we present computational results obtained by density-functional-theory (DFT) calculations coupled with cluster expansion and Monte-Carlo simulations. We examine the nature of the defect clusters and the consequences of such defect clustering tendencies for the composition and temperature dependencies of thermodynamic properties. Oxygen stoichiometric solid solutions of both ThO2 and UO2 with isovalent solute additions are shown to have positive mixing enthalpies, with the cation size mismatch being the dominant energy contribution, in quantitative agreement with continuum elasticity theory. ThO2-based solid solutions containing trivalent dopants (Sc, In, Y, Nd, Ce, La) are calculated to have positive enthalpies of formation with respect to constituent oxides and show a tendency to decrease in magnitude as the size and electronegativity of the trivalent dopant decrease. Monte Carlo simulations at elevated temperatures establish that the solid solutions display a significant degree of short-ranged defect clustering. These results are contrasted with those obtained for aliovalently doped UO2.
9:30 AM - A3.2
Electrochemistry of Defects in Irradiated UO2.
Abdel-Rahman Hassan 1 , Thomas Hochrainer 2 , Jianguo Yu 3 , Xianming Bai 3 , Todd Allen 4 , Anter El-Azab 2
1 Materials Science and Engineering Program, Florida State University , Tallahassee, Florida, United States, 2 Department of Scientific Computing , Florida State University, Tallahassee, Florida, United States, 3 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 4 , University of Wisconsin, Madison, Wisconsin, United States
Show AbstractIrradiation alters the local stoichiometry of oxides significantly. The resulting stoichiometric changes play a critical role in the dynamics of defects and microstructure evolution in oxides under irradiation. Stoichiometry in oxides is also sensitive to the surrounding oxygen environment. In general, the levels of point defects and electronic charge carriers in an oxide are sensitive to the oxygen partial pressure in contact with the oxide at hand. We investigate the electrochemistry of defects in UO2 under irradiation, where both the atomic displacements by energetic collision cascades and the exchange of oxygen with the ambient drive stoichiometric changes in the material. The problem is cast in the form of balance laws of lattice and electronic defects under defect generation and diffusion, with boundary conditions dictated by the oxygen partial pressure at the free surface. Inherent to this problem is the electrostatic field resulting from the segregation of charged lattice and electronic defects in the material. Using this model, the scenario of dynamic stoichiometric changes in a UO2 film under ion irradiation will be illustrated in detail. This research was supported as a part of the Energy Frontier Research Center on Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under subcontract # 00091538 from INL to Florida State University.
9:45 AM - A3.3
First-Principles Modeling of Defects Behavior in Ceramic Fuels.
Ying Chen 1 , Hua Y. Geng 2 , Yasunori Kaneta 3 , Jia C. Shang 4 , Motoyasu Kinoshita 5 , Shuichi Iwata 3
1 Department of Nanomechanics, Tohoku University, Sendai, Miyagi, Japan, 2 , Institute of Fluid Physics, Mianyang, Sichuan, China, 3 , The University of Tokyo, Tokyo Japan, 4 , Nuclear Power Institute of China, Chengdu, Sichun, China, 5 , Central Research Institute of Electric Power Industry, Tokyo Japan
Show AbstractUranium dioxide is a most important fuel material used in nuclear reactor, its performance quite relates to the defects behavior under irradiation which arises the deviation from stoichiometric compounds. To investigate the formation, stability mechanism and relevant physical properties of the nonstoichiometric uranium dioxide, comprehensive first principles calculations have been performed using PAW-LSDA+U method for various complex defects clusters of oxygen atoms in UO2. Calculations revealed the stability of the cuboctahedron embedded into the crystal UO2, clarified the ambiguousness remaining for long in structure of nonstoichiometric UO2+x. By incorporating the temperature effect, a pseudo phase diagram of temperature and the oxygen concentration has been constructed, and a new physical model of thermodynamic competition between cuboctahedron and point oxygen interstitials is proposed. The interplay of one main fission products, Xe, and the defect clusters in ceramics fuels has been also investigated.
References
[1] Ying Chen, Hua Y. Geng, Shuichi Iwata, et al., Comp. Mater. Sci. 49 (2010), S364
[2] H. Y. Geng, Ying Chen, Y. Kaneta, M. Kinoshita and Q. Wu, Phys. Rev. B 82 (2010), 094106
[3] H. Y. Geng, Y. Chen, Y. M. Kinoshita, et al., Applied Physics Lett. 93 (2008),201903
[4] H. Y. Geng, Ying Chen, Y. Kaneta and M. Kinoshita, Phys. Rev. B77, 180101 (2008)
[5]. H. Y. Geng, Ying Chen, Y. Kaneta and M. Kinoshita, Phys. Rev. B77, 104120 (2008)
10:00 AM - A3.4
Phase-Field Simulation of Intergranular Bubble Growth and Percolation.
Paul Millett 1 , Michael Tonks 1 , Bulent Biner 1
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractThe production of fission gas products, namely xenon and krypton, in irradiated nuclear fuel elements leads to a variety of phenomena that directly influence fuel performance. Central to the retention and release of fission gases is the evolution of bubbles existing on grain boundaries and grain triple junctions. Here, three-dimensional phase-field simulations of the growth and coalescence of intergranular Xe bubbles in UO2 bicrystal grain geometries will be presented. We investigate the dependency of bubble percolation on three factors: the initial bubble density, the Xe grain boundary diffusivity, and the bubble shape, which is governed by the ratio of the grain boundary energy over the surface energy. The simulations show that variations of each of these factors can lead to large discrepancies in the bubble coalescence rate, and eventual percolation, which may partially explain this observed occurrence in experimental investigations. This research was supported by the NEAMS program within DOE-NE.
10:15 AM - A3.5
Near-Surface Stoichiometry in UO2: A DFT Study.
Jianguo Yu 1 , Xian-Ming Bai 1 , Anter El-Azab 2 , Todd Allen 3 1
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Florida State University, Tallahassee, Florida, United States, 3 , University of Wisconsin, Madison, Wisconsin, United States
Show AbstractNear-surface stoichiometry in uranium dioxide (UO2) is an important issue in microstructural evolution of nuclear fuel under radiation. The mechanisms of oxygen release and uptake by UO2 crystals containing high density of point defects are important for understanding the dynamics of defects and microstructure in these crystals. Few or no previous studies have been performed to understand the surface effects on stoichiometry. In this work, density functional theory (DFT) calculations and temperature-accelerated dynamics (TAD) and thermodynamic analysis are used to investigate the transition of oxygen from the surface into the bulk and vice versa. This investigation will enable the modeling of the oxygen kinetics in irradiated UO2 as a function of temperature and oxygen partial pressure. In an attempt to pin down the mechanisms of oxygen exchange between the gas environment and defected UO2, the results of this work will be compared to available experimental data. This work is supported by the Center for Materials Science of Nuclear Fuel, an Energy Frontier Research Center (EFRC) funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number FWP 1356.
10:30 AM - A3.6
DFT+U Investigation of Oxygen, Uranium and Xenon Transport in Uranium Dioxide Using Electronic Occupancy Control.
Marjorie Bertolus 1 , Boris Dorado 1 2 , David Andersson 2 , Philippe Garcia 1 , Michel Freyss 1 , Blas Uberuaga 2 , Christopher Stanek 2
1 , CEA, DEN, Saint-Paul-lez-Durance France, 2 MST Division , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractUranium dioxide (UO2) attracts much interest due to its technological value as the standard nuclear fuel for pressurized water reactors. Although it has been extensively studied theoretically, its description by first principles remains challenging. The main difficulty lies in the description of the strong correlations between the 5f electrons of uranium atoms, which requires approximation beyond the standard density functional theory (DFT), such as the DFT+U. The use of the latter approximation, however, induces the presence of numerous metastable states which makes it difficult to converge to the ground state [1]. Electronic occupancy control (EOC) has been extensively applied to UO2, as well as to other actinide bearing compounds, and has been shown to tackle effectively the problem of metastable states [2,3,4].We report here results of DFT+U calculations on oxygen, uranium and xenon transport in UO2 using EOC. We have investigated several elementary migration mechanisms and have calculated the associated migration energies. Results for oxygen transport have been compared to a comprehensive range of experimental data: interstitial migration and formation energies, vacancy migration, and Frenkel pair formation energies. A very encouraging correspondence is observed between experimentally determined formation and migration energies and those calculated using the DFT+U approximation with EOC [5]. These results open up the prospect of using first-principles DFT+U calculations as part of a predictive approach to determining transport properties in other actinide compounds.References[1]P. Larson, W. R. L. Lambrecht, A. Chantis, and M. van Schilfgaarde, Phys. Rev. B 75, 045114 (2007)[2]B. Dorado, B. Amadon, M. Freyss, M. Bertolus, Phys. Rev. B 79, 235125 (2009)[3]B. Amadon, F. Jollet, and M. Torrent, Phys. Rev. B 77, 155104 (2008)[4]B. Dorado, M. Freyss, G. Jomard, M. Bertolus, Phys. Rev. B 82, 035114 (2010)[5]B. Dorado, P. Garcia, M. Freyss, G. Carlot, M. Fraczkiewicz, B. Pasquet, G. Baldinozzi, D. Simeone, M. Freyss, M. Bertolus, Phys. Rev. B 83, 035126 (2011)
10:45 AM - A3.7
Simulations of Xe Redistribution in UO2: From Atomistics to Continuum.
David Andersson 1 , Blas Uberuaga 1 , Michael Tonks 2 , Paul Millett 2 , Chris Stanek 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractRedistribution of fission gases such as Xe is closely coupled to nuclear fuel performance. For example, the formation and retention of fission gas bubbles induce fuel swelling, which leads to mechanical interaction with the clad thereby increasing the probability for clad breach. Fission gas bubbles also decrease the thermal conductivity of the fuel. Alternatively, fission gas can be released from the fuel to the plenum, which increases the pressure on the clad walls. Most fission gases have low solubility in the fuel matrix, specifically the insolubility is most pronounced for large fission gas atoms such as Xe, and as a result there is a significant driving force for segregation of gas atoms to grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. The first step of the fission gas redistribution is diffusion of individual gas atoms through the fuel matrix to existing sinks, which is governed by the activation energy for bulk diffusion. Fission gas bubbles are then formed by either separate nucleation events or by filling voids that were nucleated at a prior stage; in both cases the formation and later on growth are coupled to vacancy dynamics and thus linked to the production of vacancies via irradiation or thermal events. In order to better understand bulk Xe diffusion mechanisms in UO2±x we first calculate the relevant activation energies using density functional theory (DFT) techniques. By analyzing a combination of Xe solution thermodynamics, migration barriers and the interaction of dissolved Xe atoms with U vacancies, we demonstrate that Xe diffusion predominantly occurs via a vacancy-mediated mechanism, though other alternatives may exist for high irradiation fields. Since Xe transport is closely related to diffusion of U vacancies, we have also studied the activation energy for this process. Experimental data for the Xe and U activation energies are best reproduced if the active charge-compensation mechanism for intrinsic defects in UO2±x is considered. Due to the high thermodynamic cost of reducing U4+ ions, any defect formation occurring at a fixed composition, i.e. no change in UO2±x stoichiometry, always avoids such reactions, which, for example, implies that the ground-state configuration of an O Frenkel pair in UO2 does not involve any explicit local reduction (oxidation) of U ions at the O vacancy (interstitial). Next a continuum transport model for Xe and U is formulated based on the diffusion mechanisms established from DFT. After combining this model with descriptions of the interaction between Xe and UO2 grain boundaries derived from separate atomistic calculations, we simulate Xe redistribution for a few simple microstructures using finite element methods (FEM), as implemented in the MOOSE framework from Idaho National Laboratory.
11:30 AM - A3.8
On the Deformation of UO2: A Molecular Dynamics Study.
Paul Fossati 1 , Remi Dingreville 3 4 , Laurent Van Brutzel 1 , Jean-Paul Crocombette 2 , Timothy Bartel 4
1 DEN/DPC/SCP, CEA Saclay, Gif-sur-Yvette France, 3 Department of Mechanical and Aerospace Engineering, Polytechnic Institute of New York University, Brooklym, New York, United States, 4 Advanced Nuclear Fuel Cycle Tech., Sandia National Laboratories, Albuquerque, New Mexico, United States, 2 DEN/DMN/SRMP, CEA Saclay, Gif-sur-Yvette France
Show AbstractExemplified by the recent events in Fukushima Japan, increasing concerns about energy security and the environmental impact of energy use have led to intensive interest in nuclear power. In order to safely extend the lifetime of existing nuclear reactors and develop fuels for the next generation of reactors, we need have a fundamental understanding of the mechanical behavior of reactor fuel, uranium dioxide (UO2). Although UO2 pellets are widely used as nuclear fuel, uncertainty still remains about their thermo-mechanical behavior. Most of our knowledge regarding UO2 mechanical behavior is obtained by experiments on unirradiated fuel, or post-mortem analysis on spent fuel. Atomistic models give us a good grasp on what is the behavior of the fuel in conditions inaccessible to current experiments, and the related effects at a larger length-scale.The present investigation considers recent studies on the mechanical properties of UO2 by means of atomistic simulations using empirical potentials. For this study four different rigid ion potentials have been assessed. UO2 elasticity and plastic behavior are of primary concerns in this study. Firstly, the elastic constants of different polymorphs (fluorite, rutile, α-PbO2) have been studied for a wide range of temperatures. Temperature and orientation dependence will be discussed. Secondly, investigations and reflection on the stability and motion mechanisms of dislocations in UO2 will be discussed. Finally, crack initiation and propagation in UO2 single-crystal will be considered in light of the aforementioned results. All of these studies constitute a coherent set of parameters that can be used in mesoscale and continuum models to investigate creep and plastic deformation of nuclear fuel.
11:45 AM - A3.9
Temperature Accelerated Dynamics Simulations of Defect Clustering in UO2.
Xian-Ming Bai 1 , Jianguo Yu 1 , Anter El-Azab 2 , Todd Allen 3 1
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Florida State University, Tallahassee, Florida, United States, 3 , University of Wisconsin, Madison, Wisconsin, United States
Show AbstractThe aggregation of radiation-induced point defects in uranium dioxide (UO2) is a critical step in microstructural evolution, and consequently can have significant effects on material properties such as thermal and mass transport and mechanical properties. Therefore, understanding the kinetic evolution of point defects in UO2 is important for predicting nuclear fuel performance. Here we use temperature accelerated dynamics (TAD) simulations with the Basak potential to investigate the elementary atomistic mechanisms involved in defect clustering processes across different timescales at both 300 K and 1000 K. We investigate the binding and migration energies of different cluster sizes and configurations. Several types of defect clusters are observed including the cuboctahedral oxygen interstitial cluster. The comparison between our TAD simulation results with density functional theory calculations is also discussed. These results give new insight into the initial stages of microstructural evolution necessary for predicting fuel performance. This work is supported by the Center for Materials Science of Nuclear Fuel, an Energy Frontier Research Center (EFRC) funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number FWP 1356.
12:00 PM - **A3.10
Properties of Vacancy Defects Induced in UO2 by Irradiation and Probed by Using Positron Annihilation Spectroscopy.
Marie-France Barthe 1 , Tayeb Belhabib 1 , Stéphanie Leclerc 1 , Laszlo Liszkay 1 , Hicham Labrim 1 , Virginie Moineau 1 , Pierre Desgardin 1 , Gaelle Carlot 2 , Philippe Garcia 2
1 CEMHTI, CNRS, Orleans France, 2 DEN/DEC/SESC, CEA, Saint Paul lez Durance France
Show AbstractThe understanding of the behavior of fission nuclear fuel under irradiation is of first importance to foresee the state of the fuel in reactors and also its evolution after use in storage conditions. The study of this behavior can be carried out by the characterization of the in pile irradiated fuel. This work requires heavy installations (hotlabs..) and long term experiments (decay …) and doesn’t always allow to discriminate between interdependent phenomena which can occur in the material. These studies have to be completed with separate effects experiments associated with multi-scale modeling. This approach allows a more detailed knowledge of the different phenomena and their interconnections. The complete modeling of the behavior of material under irradiation begins with the first stages of damage and requires the detailed knowledge of fundamental data especially concerning the point defects properties. These data can be obtained by calculations and/or from experiments performed in separate effects conditions. Uranium dioxide, as the major component, is used as the model material of the nuclear fission fuel . The damage induced in UO2 by irradiation has been extensively studied by using different techniques such as Channeling Rutherford Backscattering, RX diffraction, Transmission Electron Microscopy and so on. Very few studies have been focused on the direct observation of the point defects and the determination of their properties in the UO2 matrix. In this work, we have used positron annihilation spectroscopy (PAS) to characterize the vacancy defects induced in UO2 by irradiation in different conditions (different particles, energies and fluencies). Both 22Na based Positron Annihilation Lifetime Spectroscopy (PALS) and coincidence Doppler annihilation-ray Broadening Spectrometry (CDBS) and Slow Positron Beam coupled with Doppler Broadening Spectrometry (SPBDBS) have been used to probe these vacancy defects and study their evolution as a function of temperature. In some cases it can be very useful to follow the behavior of gas such as Helium by using Nuclear Reaction Analysis and to identify the interaction between both identities He and vacancy. This work was funded in part by F-Bridge Project which is part of the European Union’s Framework Programme 7 and by the National Research Group MATINEX of the French PACEN Programme.
12:30 PM - **A3.11
Atomistic Simulation of Nuclear Fuels.
Matthias Krack 1
1 , Paul Scherrer Institute, Villigen PSI Switzerland
Show AbstractThe experimental investigation of actinide materials like nuclear fuels is difficult and usually very costly. Therefore a reliable multi-scale modeling of these often hazardous materials starting at the atomistic level is inevitable to gain further insight into this type of materials. The development of new, more advanced simulation methods accompanied by the rapid growth of the available computational resources provided by high-performance computing facilities, allows the modeling of such materials at a new quality level. Also the recent development of the CP2K program package (http://cp2k.berlios.de) has been partially focused on enabling state-of-the-art simulations of actinide materials using classical potential as well as electronic structure methods. The goal is to perform reliable molecular dynamics simulations for actinide materials including advanced simulation techniques like metadynamics. Metadynamics is an accelerated molecular dynamics method allowing for the fast exploration of a system's energy landscape, even if energy barriers are present which are large compared to the typical thermal fluctuations. In this way, rare events can be observed within the time periods accessible by standard molecular dynamics simulation runs. The CP2K program package and some of its first applications to actinide materials, especially uranium dioxide, are presented.
A4: Radiation Damage - Metals
Session Chairs
Tuesday PM, November 29, 2011
Independence W (Sheraton)
2:45 PM - A4.2
Irradiation-Induced Creep in Dilute Nanostructured Cu-W Alloys.
Kaiping Tai 1 , Yinon Ashkenazy 2 , Robert S. Averback 1 , Pascal Bellon 1
1 Department of Materials Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, Illinois, United States, 2 Racah Institute of Physics, Hebrew University of Jerusalem, Jerusalem Israel
Show AbstractThe development of new radiation resistant materials focuses largely on creating high densities of neutral sinks for point defects, such as nanosized inclusions and grain boundaries. While this strategy is appears useful for eliminating point defects and/or trapping He gas atoms, very little is known about the creep properties of these materials under the extreme conditions of an irradiation environment. We report here, new in situ creep measurements on 1.8 MeV Kr+ irradiated Cu and dilute Cu-W nanostructured alloys films as a function of temperature using plane-strain bulge testing. Primary and secondary creeps were observed at all temperatures. The secondary creep rates in the Cu-W alloys were observed to increase with increasing temperature between 300 K and 473 K, and then become constant up to 573K. An activation enthalpy of 0.30±0.05 eV was obtained for Cu93.5W6.5 and Cu99W1 alloys. Subsequent (scanning) transmission electron microscopy analysis revealed a high density of small (2-3 nm) W-rich nanoparticles in these irradiated samples. The precipitates had a BCC structure and were (semi) coherent with the Cu matrix. The alignment of specific crystallographic planes in the W and Cu follow the K-S or N-W orientations. Analysis of kinetic behavior shows that the irradiation-enhanced creep in these materials derives from neutral point defect fluxes to the grain boundaries. A new model of radiation-enhanced creep in ultrafine grained metals, based on molecular dynamics simulations, is presented; it explains how neutral fluxes of point defects can lead to creep deformation.
3:00 PM - A4.3
An Atomic-Level Perspective into the Evolution of Interfaces in Purposed Radiation Tolerant Materials.
Jeffery Aguiar 1 2 , Nigel Browning 1 2 5 , Luke Hsiung 2 , Michael Fluss 2 , Peter Hosemann 3 4 , Sanchita Dey 1
1 Chemical Engineering and Material Science Department, University of California Davis, Davis, California, United States, 2 Condensed Matter and Materials Division, Lawrence Livermore National Laboratory, Livermore, California, United States, 5 Molecular Cell Biology, University of California Davis, Davis, California, United States, 3 Nuclear Engineering, University of California Berkeley, Berkeley, California, United States, 4 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractIn an attempt to address the energy crisis, next generation nuclear technology poses as a serious candidate. Conditionally next generation nuclear power also presents serious issues related to reactor safety and lifetimes due to unknown longevity of reactor containment materials due to heightened reactor environments. In order to address these concerns relating to containment materials, several materials issues are currently and remain to be investigated, ranging from corrosion to synergistic radiation induced effects in a series of purposed damage resistant materials, including oxide dispersion strengthened (ODS) alloys. To address the pressing need for analytical characterization of proposed radiation tolerant materials, we have used the latest the technologically advanced techniques in simultaneous imaging and spectral analysis using aberration corrected (scanning) transmission electron microscopy (S)TEM coupled with high resolution electron energy loss spectroscopy (HR-EELS) and energy dispersive x-ray (EDX) analysis to study a series of ODS alloys. In our work we are focused on the use the use of aberration corrected microscopy and spectroscopy to characterize three purposed radiation tolerant materials, ODS alloys MA-957, K3, and PM-2000, with and without the effects of radiation damage at particle-matrix interfaces. to evaluate for structural and chemical changes, such as radiation induced segregation and swelling. The use of oxide dispersed constituents increases the likelihood of interfacial defect trapping before reaching a grain boundary, but the same interfaces over the lifetime of radiation evolve with the inclusion of trapped defects, complexes, and helium bubbles which requires the uttermost best spatial and energy resolution to decouple. The progression of these atomic-level interfaces in ODS alloys has thereby been studied in great detail using aberration corrected (S)TEM and energy filtered spectral analysis, EDX and EELS, to fundamentally develop an atomic level perspective into the structural and chemical evolution of these three candidate ODS alloys. The talk will therefore focus on the principle capabilities and use of (S)TEM and EELS to study and develop atomic level perspective into the evolution of interfaces in ODS alloys with the uttermost highest achievable spatial and energy resolution.
3:15 PM - A4.4
Cavity Formation in Multi-Ion-Beam Irradiated ODS Ferritic Steel.
Luke Hsiung 1 , Scott Tumey 1 , Michael Fluss 1
1 Physical and Life Sciences, Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractOne of the major challenges in designing fusion reactors is to develop the high performance structural materials for first wall and breeding-blanket components, which will be exposed to high fluxes of high-energy (14 MeV) neutrons from the deuterium-tritium fusion and helium and hydrogen from (n, α)- and (n, p)-transmutation reactions. Although significant progress has been made recently to understand the processing-microstructure-property relationships of ferritic and martensitic (F/M) steels and oxide dispersion strengthened (ODS) F/M steels, it remains to understand the role of helium and hydrogen transmutation gases on the cavitational swelling of F/M and ODS F/M steels. Since no prototype fusion reactors currently exist, it is difficult to directly evaluate the high-energy neutron damage environment expected to prevail in the first wall of a fusion reactor. One technique commonly used to study the evolution of defect structures and the nucleation and growth of voids utilizes transmission electron microscopy (TEM) examinations of specimens simultaneously bombarded by heavy ions and helium and/or hydrogen ions through so called "dual-beam" and "triple-beam" experiments. The heavy ions create atomic displacements while the gas ions lead to the effects of the transmutation gases, helium (10 appm/dpa) and hydrogen (40 appm/dpa). We have recently conducted HRTEM studies to compare radiation effects on Fe-14Cr alloy and Fe-16Cr ODS steel using (He + Fe) dual-beam and (H + He + Fe) triple-beam techniques. Important results will be presented to address the effects of nanoparticles on the suppression of radiation-induced cavitational swelling. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
3:30 PM - **A4.5
In Situ TEM Studies of Microstructure Evolution under Ion Irradiation for Nuclear Engineering Applications.
Djamel Kaoumi 1 , Jimmy Adamson 1 , Arthur Motta 2 , Brian Wirth 3 , Aaron Kohnert 3 , Mark Kirk 4
1 Mechanical and Nuclear Engineering, University of South Carolina, Columbia, South Carolina, United States, 2 Mechanical and Nuclear Engineering, Penn State , University Park, Pennsylvania, United States, 3 Nuclear Enggineering, University of Tennessee, Knoxville, Tennessee, United States, 4 Materials Science Division, Argonne National Laboratories, Argonne, Illinois, United States
Show AbstractOne of the difficulties of studying processes occurring under irradiation (in a reactor environment) is the lack of kinetics information since usually samples are examined ex situ (i.e. after irradiation) so that only snapshots of the process are available. Given the dynamic nature of the phenomena, direct in situ observation is invaluable for better understanding the mechanisms, kinetics and driving forces of the processes involved. This can be done using in situ ion irradiation in a TEM at the IVEM facility at Argonne National Laboratory. To predict the in reactor behavior of alloys, it is essential to understand the basic mechanisms of radiation damage formation (loop density, defect interactions) and accumulation (loop evolution, precipitation or dissolution of second phases…). In-situ ion-irradiation in a TEM has proven a very good tool for that purpose as it allows for the direct determination of the formation and evolution of irradiation-induced damage and the spatial correlation of the defect structures with the pre-existing microstructure (including lath boundaries, network dislocations and carbides) as a function of dose, dose rate, temperature and ion type Using this technique, different aspects of microstructure evolution under irradiation were studied, such as defect cluster formation and evolution as a function of dose in advanced Ferritic/Martensitic (F/M) steels, the irradiation stability of precipitates in Oxide Dispersion Strengthened (ODS) steels, and irradiation-induced grain-growth. In this paper we will emphasize the work done on model F/M steels (Fe12Cr0.1C and Fe-9Cr-0.1C) which were irradiated with 1 MeV Kr ions at 50K, 180K, 298K, 473K, 573K to doses up to 10 dpa in-situ in a TEM. The microstructure evolution under irradiation was followed and characterized at successive doses in terms of defect formation and evolution, black dot density, and stability of as-fabricated microstructure using weak-beam dark-field imaging and g.b analysis for comparison with computations made using a spatially dependent rate theory model of cluster evolution in both compositional and geometric spaces under conditions of high energy ion irradiation. More specifically, in the model the concentrations of interstitial loops and voids are calculated as a function of time, number of interstitials/vacancies, spatial position, dislocation densities, temperature, dose and dose rate, impurities and so on. This presentation will focus on the experimental in-situ TEM observations.
A5: Metallic Systems - Modelling
Session Chairs
Tuesday PM, November 29, 2011
Independence W (Sheraton)
4:30 PM - A5.1
Modelling Radiation Effects in ODS Steels.
Roger Smith 1 , Tomas Lazauskas 1 , Steven Kenny 1
1 , Loughborough University, Loughborough, Leicestershire, United Kingdom
Show AbstractODS materials are promising candidates for use in both fission and fusion reactors. It has been suggested that such materials can be more radiation resistant and stronger than other steels and can minimise the effect of inert gas bubble accumulation. Here we present the first results arising from a UK-India nuclear collaboration project of the simulation of radiation in these materials. Yttria particles are embedded in a bcc Fe matrix with a size distribution corresponding to those that are experimentally observed in a typical ODS material. The system is modelled using a fixed charge potential for the Y-O interactions and an embedded atom type potential for the Fe-Fe interactions. Collision cascades at various energies are initiated in the Fe matrix and the effect of the embedded nanoparticles on the cascade development is reported.It is shown that under certain circumstance the nanoparticles deflect that moving Fe atoms with a tendency to trap defects at the interface between the nanoparticles and the matrix.
4:45 PM - A5.2
Structure and Properties of the Y2O3/ Fe Interface from First Principles Calculations.
Samrat Choudhury 1 , Christopher Stanek 1 , Blas Uberuaga 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractNanostructured ferritic alloys (NFAs) (with a typical composition of 12–14 wt% Cr, 0.25 wt% Y and 0.5 wt% Ti) are considered excellent candidate materials for structural applications in nuclear reactors as they exhibit exceptionally high creep strength and radiation tolerance due to the presence of highly stable nanometer sized Y-Ti-O precipitates (NPs). It is believed that these properties result from the characteristics of the particle and ferritic matrix interface. Y2O3 has also been shown to form nanoprecipitates in Fe and is a simpler surrogate for the Y-Ti-O precipitates. In this work, we will present the behavior of the interface between the ferritic matrix and Y2O3 using density functional theory. In particular, the role of alloying elements and orientation relationship on the atomic structure of the particle-matrix interface, segregation energies of the alloying elements, electronic structure at the interface, and calculated interfacial energy will be discussed. These results form the basis of a phase-field model that will examine the nucleation and growth of Y2O3 precipitates in Fe.
5:00 PM - A5.3
Evolutions of Oxide Particles and Grain Morphology of 12 Cr ODS Steel.
Jinsung Jang 1 , Tae Kyu Kim 1 , Xiaodong Mao 1 2 , Chang Hee Han 1 , Young Soo Han 1 , Kyu Hwan Oh 2
1 Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon Korea (the Republic of), 2 Dept. of Materials Science & Engineering, Seoul National University, Seoul Korea (the Republic of)
Show AbstractOxide dispersion strengthened (ODS) steel is one of good candidate materials for in-core components of Generation IV nuclear systems due to its good high temperature mechanical strenghth as well as the excellent neutron radiation resistance. 12Cr ODS steel samples were prepared by mechacanical alloying(MA) of the elmental metal powders along with 20-30 nm yttria (Y2O3) particels as the strengthening dispersoids. MA powders were reduced by hydrogen mixture gas during degassing process, and then consolidated by hot isostatic pressing(HIP) and hot rolling(HR).Evolutions of yttrium containing oxide particles such as YTaO4 or Y3TaO7 as well as the grain morphology after each step are investigted using SEM/EBSD(Scanning Electron Microscopy /Electron Backscattered Diffraction), TEM(Transmission Electron Microscopy), and the particle size distributions are estimated by SANS(Small Angle Neutron Scattering) and compared with those by other analytical techniqes.
5:15 PM - A5.4
An Atom Probe Study of Radiation-Induced Segregation/Depletion in a Fe-14.25wt%Cr Ferritic Steel.
Rong Hu 1 , George Smith 1 , Emmanuelle Marquis 2
1 Department of Materials, University of Oxford, Oxford United Kingdom, 2 Department of Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractFerritic chromium steels are important structural materials for future nuclear fission and fusion reactors due to their advantages over traditional austenitic steels, including low swelling rates, better thermal fatigue resistance, and lower thermal expansion coefficients. Radiation-induced segregation or depletion (RIS/RID) of solute atoms at grain boundaries is considered to be a potentially significant phenomenon for structural materials because of its potentially detrimental role in affecting microstructure and furthermore mechanical properties. However, the behaviour of Cr at grain boundaries in ferritic steels is not well understood. Both segregation and depletion of Cr at grain boundary under irradiation have been previously observed and no clear dependency on irradiation condition or alloy type has been presented. To understand the Cr behaviour at grain boundaries in ferritic steels under irradiation, a systematic approach combining EBSD, FIB specimen preparation and atom probe tomography analysis has been applied on a Fe-14.25wt%Cr to investigate the effect of grain boundary orientation, irradiation depth, impurities and other factors to get a better understanding of RIS/RID phenomenon. Both low sigma boundaries and high sigma boundaries have been investigated in detail and systematic differences between the behaviour of different classes of boundaries will be reported.
5:30 PM - A5.5
Calculation of Reaction Constants for Vacancy Migration in α-Iron Using Transition Path Sampling with Lyapunov Bias.
Massimiliano Picciani 1 , Manuel Athenes 1 , Mihai-Cosmin Marinica 1
1 DEN/DMN/SRMP, CEA Saclay, Gif-sur-Yvette France
Show AbstractPredicting the microstructural evolution of radiation damage in materials requires handling the physics of infrequent-events, in which several time scales are involved. The reactions rates which characterize those events are the main ingredient for simulating the kinetics of materials under irradiation over large time scales and high irradiation doses. Here we propose an efficient, finite temperature method to compute reaction rate constants of thermally activated processes.
The method consists of two steps. Firstly, rare reactive trajectories in phase-space are sampled using a Transition Path Sampling[1] algorithm supplemented with a Lyapunov bias[2] favoring diverging trajectories. This enables the system to visit transition regions separating stable configurations more often, and thus enhances the probability of observing transitions between stable states during relatively short simulations. Secondly, reaction constants are estimated from the unbiased fraction of reactive trajectories[1] , yielded by an appropriate statistical data analysis tool, the Multistate Bennett Acceptance Ratio package[3].
We apply our method to the calculation of reaction rates for the migration of vacancies and divacancies in an α-Iron crystal, testing different Embedded Atom Model potentials[4], for temperatures ranging from 250 K to 900 K. Vacancy diffusion rates associated with activation barriers at finite temperature are then evaluated, showing a significant difference from values obtained using the standard harmonic approximation. Finally, the calculated diffusion constants are employed as input parameters in a first passage kinetic Monte Carlo (FPKMC)[5] code in order to model the migration and clustering of defects in resistivity recovery experiments[6].
[1] C. Dellago,P.G. Bolhuis, D. Chandler, J. Chem. Phys. 110 6617 (1999)
[2] M. Picciani, M. Athènes, M.-C. Marinica, in preparation
[3] M.R. Shirts, J.D. Chodera, J. Chem. Phys. 129 124105 (2008)
[4] L. Malerba, M.C. Marinica, N. Anento, C. Bjorkas, H. Nguyen, C. Domain, F. Djurabekova, P. Olsson,K. Nordlund, A. Serra, D. Terentyev, F. Willaime, C.S. Becquart, J. Nucl. Mat. 406 19 (2010)
[5] T. Oppelstrup, V.V. Bulatov, G.H. Gilmer, M.H. Kalos, B. Sadigh, Phys. Rev. Lett. 97 230602, (2006)
[6] C.C. Fu, J.Dalla Torre, F. Willaime, J.L. Bocquet, A. Barbu, Nature Materials 4 68 (2004)
5:45 PM - A5.6
Effects of Li on Zirconium Alloy Corrosion – Li Insertion, and Ion Migration in ZrO2.
Mostafa Youssef 1 , Bilge Yildiz 1
1 Nuclear Science and Engineering Department, MIT, Cambridge, Massachusetts, United States
Show AbstractIt is known that the corrosion resistance of the zirconium alloys is greatly diminished when there is a high concentration of Li in water. Accelerated corrosion of zirconium alloys pose safety and operational challenges as they serve as nuclear fuel cladding. The mechanisms and kinetics that lead to Li incorporation into the zirconium oxide, the passive layer on zirconium metal in water corrosion, and the corresponding acceleration of corrosion are described only macroscopically and empirically. An understanding of the structure and the corrosion characteristics of lithiated zirconium oxide at a fundamental level would allow for the predictive assessment of corrosion kinetics, and strategies against Li trapping. In this study, we investigate the structure of zirconium oxide with Li insertion at interstitial and at substitutional sites, identify the diffusion barriers of Li and of oxygen in the lithiated structure, and the relative stabilities of the monoclinic and tetragonal zirconia phases when Li is inserted, all of which affect the protective characteristics of the zirconium oxide, and thus, the corrosion kinetics. We use first principles methods based on density functional theory calculations in our analysis. Our initial results show that the transition from a protective tetragonal phase to a less protective monoclinic phase is not driven by the presence of interstitial Li atoms. Correlations of Li presence and ion transport barriers are assessed and discussed in connection to corrosion kinetics.
Symposium Organizers
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Marjorie Bertolus CEA, DEN, DEC/SESC/LLCC
Blas Uberuaga Los Alamos National Laboratory
Robin W. Grimes Imperial College London
A6: Nuclear Ceramics
Session Chairs
Blas Uberuaga
Yanwen Zhang
Wednesday AM, November 30, 2011
Independence W (Sheraton)
9:45 AM - A6.2
Synergy between Electronic and Nuclear Stopping in Producing Ion Irradiation Damage in Amorphous SiO2.
Marie Backman 1 2 , Flyura Djurabekova 2 , Kai Nordlund 2 , Yanwen Zhang 1 3 , Marcel Toulemonde 4 , William Weber 1 3
1 Department of Materials Science and Engineering, University of Tennessee, Knoxville , Tennessee, United States, 2 Helsinki Institute of Physics and Department of Physics, University of Helsinki, Helsinki Finland, 3 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 4 CIMAP-CEA-CNRS-ENSICAEN, University of Caen, Caen France
Show AbstractIon irradiation in SiO2 can cause damage by elastic collisions between the ion and target nuclei at low ion energy (high nuclear stopping), or by energy deposition to the electrons, which leads to melting within an ion track at high ion energy (high electronic stopping). It has recently been shown experimentally [1] that there is a synergy between the two damage mechanisms in amorphous SiO2 in the intermediate energy regime. In this work we model ion irradiation in the energy regime where nuclear and electronic stopping powers are comparable in magnitude in order to better understand this synergy. Using molecular dynamics (MD) simulations we examine the structural modification of amorphous SiO2 due to one Au ion impact. The ion energy is varied between 0.6 MeV (Sn = 3.3 keV/nm, Se = 0.36 keV/nm) and 77 MeV (Sn = 0.31 keV/nm, Se = 11.5 keV/nm). The nuclear energy deposition is simulated by recoils obtained by binary collision approximation (BCA) calculations and the electronic energy deposition to the atoms from the ion is calculated using the thermal spike model [2]. The simulations provide a means for isolating the contributions from nuclear and electronic stopping to the observed radiation damage and can thus provide further insight to experimental observations of the synergy effect.
[1] M. Toulemonde, W. J. Weber, G. Li, V. Shutthanandan, P. Kluth, T. Yang, Y. Wang, and Y. Zhang, Phys. Rev. B 83, 054106 (2011)
[2] P. Kluth, C.S. Schnohr, O.H. Pakarinen, F. Djurabekova, D. J. Sprouster, R. Giulian, M.C. Ridgway, A.P. Byrne, C. Trautmann, D.J. Cookson, K. Nordlund, and M. Toulemonde Phys. Rev. Lett. 101, 175503 (2008)
10:00 AM - A6.3
The Effect of Structure on Threshold Displacement Energy: A Case Study of TiO2 Polymorphs.
Marc Robinson 1 , Nigel Marks 1 , Karl Whittle 2 , Greg Lumpkin 2
1 Nanochemistry Research Institute, Curtin University, Perth, Western Australia, Australia, 2 , Australian Nuclear Science and Technology Organisation, Sydney, New South Wales, Australia
Show AbstractThe threshold displacement energy (Ed) is a fundamental quantity that is pivotal in defining the radiation tolerance of a material. It is of significant importance in analytical models of radiation damage, such as SRIM, where it governs overall defect production. Historically, values of Ed have been extrapolated from experimental studies, which fail to capture the onset of defect production and the related defect formation mechanisms. Computer simulation methods have enabled insight into these areas, yet substantial computational effort is required to ensure statistically sound results are achieved. In this work, an extensive, systematic approach to calculating Ed has been adopted using molecular dynamics simulation. Our approach employs a large ensemble of Primary Knock-on Atoms (PKAs) in which directions are drawn from a dense sampling of the unit sphere and kinetic energies are finely spaced. This methodology will have general utility and is particularly useful for complex crystal types where symmetry cannot be exploited. Here we demonstrate our method in the context of three low-pressure polymorphs of TiO2, that is rutile, anatase and brookite. Consistent with in-situ irradiation experiments and thermal spike simulations, we find that Ed for rutile is significantly higher than the other two polymorphs. An observation common from all simulations is the creation of replacement chains on the oxygen sublattice, generated by both O and Ti PKAs. These chains are readily formed and play a key role in governing the formation of stable defects. This study highlights the difficulty in defining a single value of Ed, suggesting a definition based on defect formation probability is more suitable.
10:15 AM - A6.4
Radioparagenesis: Robust Nuclear Waste Form Design and Novel Material Discovery.
Boris Dorado 1 , Blas Uberuaga 1 , Christopher Stanek 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractNearly thirty years ago, Gray [1] and Vance et al. [2] posed the question: if a major constituent of a crystalline compound is an unstable isotope, what is the effect of the transmutation of that isotope to a chemically different element on the structure and stability of the compound? These researchers were explicitly interested in the impact of this transmutation on the performance of ceramic nuclear waste forms. However, since the half-lives of the so-called “short-lived” isotopes in the nuclear waste stream (e.g. 90Sr and 137Cs) is approximately 30 years, corresponding experiments require more than 100 years. As such, a systematic answer to the transmutation question has remained elusive.The transmutation question has recently been revisited with modern computational materials science tools. Using density functional theory (DFT) calculations, we have attempted to reproduce the evolution of a simplified example crystalline waste form, 109CdS, during transmutation of 109Cd to 109Ag via γ-decay, i.e. Cd(1-x)AgxS. By surveying the lattice energies of all possible crystal structures for a range of x, we predicted that the rock salt structure is thermodynamically favored (over the CdS wurtzite structure) at x≈0.6, and this preference increases with increasing x. Based on this result (and similar results for other systems), we have introduced the concept of radioparagenesis, which we define as the formation of compounds, often unconventional (e.g. rocksalt AgS), due to the chemical transmutation that occurs during in situ radioactive decay.In this presentation, implications of radioparagenesis on nuclear waste form design, unconventional defect chemistry and novel materials discovery will be discussed in the light of electronic structure calculations based on DFT. Also, recently devised experiments aimed at confirming radioparagenesis by significantly accelerating the transmutation process, using small quantities of highly radioactive samples, will be discussed. Results of these experiments will be compared to DFT results.[1] W.J. Gray, “Fission product effects on high-level radioactive waste forms,” Nature, 296 (1982) 547-549.[2] E.R. Vance, R. Roy, J.G. Pepin and D.K. Agrawal, “Chemical mitigation of the transmutation problem in crystalline nuclear waste radiophases,” Journal of Materials Science, 17 (1982) 947-952.
10:30 AM - **A6.5
Radioparagenesis: The Effect of Transmutation on Crystal Structure and Stability, and Implications for Robust Nuclear Waste Form Design.
Chris Stanek 1 , Blas Uberuaga 1 , Boris Dorado 1 , Kurt Sickafus 1 , Laura Wolfsberg 1 , Russell Feller 1 , Brian Scott 1 , Wayne Taylor 1 , Meiring Nortier 1 , Nigel Marks 2
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , Curtin University, Perth, Western Australia, Australia
Show AbstractNearly thirty years ago, Gray and Vance, Roy, et al. posed the question: if a major constituent of a crystalline compound is an unstable isotope, what is the effect of the transmutation of that isotope to a chemically different element on the structure and stability of the compound? These researchers were explicitly interested in the impact of this transmutation on the performance of ceramic nuclear waste forms. However, since the half-lives of the so-called “short-lived” isotopes in the nuclear waste stream (e.g. 90Sr and 137Cs) is ~30 years, corresponding experiments require more than 100 years. As such, a systematic answer to the transmutation question has remained elusive.We have recently revisited the transmutation question with modern computational materials science tools. Using density functional theory calculations, we have attempted to reproduce the evolution of a simplified example crystalline waste form, 137CsCl, during transmutation of 137Cs to 137Ba via β- decay, i.e. 137Cs1-x137BaxCl. By surveying the lattice energies of all possible crystal structures for a range of x, we predict that the rocksalt structure is thermodynamically favored (over the CsCl structure) at ~x=0.2, and this preference increases with increasing x. Based on this result (and similar results for other systems), we have introduced the concept of radioparagenesis, which we define as the formation of compounds, often unconventional (e.g. rocksalt BaCl), due to the chemical transmutation that occurs during in situ radioactive decay.In this presentation, implications of radioparagenesis on nuclear waste form design, unconventional defect chemistry and novel materials discovery will be discussed. Also, recently devised experiments aimed at confirming radioparagenesis by significantly accelerating the transmutation process, using small quantities of highly radioactive samples, will be discussed. These experiments require a compromise between sample size and sample activity – that is, enough sample must be produced (typically via an accelerator-based process) that it can be characterized, but the amount must be kept small in order to allow for safe handling. Furthermore, choice of isotope (chemically distinct daughter, half-life conducive to experimentation, etc) is an important consideration. Specifically, experiments involving Cd-109 (which decays to Ag-109) and Be-7 (which decays to Li-7) will be discussed. Results of these experiments will be compared to DFT calculations.
11:30 AM - A6.6
The Role of Antisite Disorder on Pre-Amorphization Swelling in Titanate Pyrochlores.
Blas Uberuaga 1 , Yuhong Li 2 , Chao Jiang 3 , Samrat Choudhury 1 , James Valdez 1 , Maulik Patel 1 , Jonghan Won 1 , Yongqiang Wang 1 , Ming Tang 1 , Doug Safarik 1 , Darrin Dyler 1 , Ken McClellan 1 , Igor Usov 1 , Thomas Hartmann 4 , Gianguido Baldinozzi 5 , Kurt Sickafus 1
1 , Los Alamos Natl Lab, Los Alamos, New Mexico, United States, 2 , Lanzhou University, Lanzhou China, 3 , Central South University, Changsha China, 4 , University of Nevada, Las Vegas, Las Vegas, Nevada, United States, 5 , Ecole Centrale Paris, Chatenay-Malabry France
Show AbstractIon irradiation experiments and atomistic computer simulations were used to demonstrate that irradiation-induced lattice swelling in a complex oxide, Lu2Ti2O7, is due initially to the formation of cation antisite defects. X-ray diffraction (XRD) revealed that cation antisite formation correlates directly with pre-amorphization lattice swelling. XRD indicates that the volume per cation antisite pair is approximately 12 cubic angstroms. First principles calculations revealed that lattice swelling is best explained by near-neighbor antisite defects. Temperature accelerated dynamics simulations indicate that cation Frenkel defects are metastable and decay to form neighboring antisite defects.
11:45 AM - A6.7
Computer Simulation of Glass-Crystal Interfaces in Nuclear Waste Glasses.
Michael Rushton 1 , Robin Grimes 1 , Scott Owens 2 1
1 Department of Materials, Imperial College London, London United Kingdom, 2 , National Nuclear Laboratory, Risley United Kingdom
Show AbstractMixed alkali borosilicate glasses show considerable promise as host materials for long-term immobilisation of high level nuclear waste. Glass wasteforms can contain crystalline phases due either to incomplete dissolution of the powdered waste calcine, or partial crystallization of the glass (an effect enhanced by the heating effect of radionuclide decay). Under repository conditions, vitrified waste is expected to immobilise waste elements for millennia and thus, the manner in which interfaces might affect wasteform performance must be considered as part of a responsible nuclear-waste programme. As a first step, an improved understanding of interface structures in these glass systems is required.Interfaces between sodium lithium borosilicate glasses and the (100) and (110) surfaces of MgO and CaO crystals were simulated using a melt-quench procedure, that employed classical pair potentials and molecular dynamics. The density of network forming species within the glass at these interfaces was considered as a function of distance from the plane of the interface and the positions of network formers were calculated in relation to sites in the crystal surface.In the immediate vicinity of the glass-crystal interfaces, a strong correlation was observed between the position and orientation of the boron and silicon coordination polyhedra within the glass volume and the particular positions of ions in the crystal surface. Examination of the glass, oxygen density, profiles normal to the interface, revealed a sequence of consistently spaced layers of increased density, for all systems considered. This indicates that a degree of order had been imposed on the glass network by the crystal even at relatively large distances from the interface. In addition, strong segregation of network modifying species towards the interface was observed. Understanding, these structural and compositional effects could have a significant effect on helping to improve the performance and durability of glass wasteforms.
12:00 PM - A6.8
Radioparagenesis in 90SrF2: A DFT Study of Structural Stability in ZrF2.
Michel Sassi 1 , Nigel Marks 1 , Blas Uberuaga 2 , Christopher Stanek 2
1 Nanochemistry Research Institute, Curtin University, Perth, Western Australia, Australia, 2 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractIn the 1970's and 80's large quantities of Cs-137 and Sr-90 were extracted from spent fuel at Hanford and synthesized into capsules of CsCl and SrF2. Accounting for around one-third of the radioactivity at the Hanford site, these capsules are reportedly the most lethal source of radiation in the US outside of a reactor. Stored underwater due to the radiation field, monitoring primarily consists of 'clunk-tests' to assess swelling associated with radiolytic and/or phase decomposition. Here we present density-functional-theory (DFT) calculations on radioparagenetic ZrF2 arising from the beta-decay of the parent 90SrF2 phase. By following imaginary phonon modes from fluorite ZrF2 and also considering analogous materials (ZrH2 and BaF2) we assess the stability of potential ZrF2 daughter phases. We find two probable structures, one with monoclinic symmetry, the other with orthorhombic symmetry. In contrast with the analogous ZrH2, both of our ZrF2 structures are semiconductors, possessing small band-gaps (0.1-0.6 eV) and showing signs of covalent bonding. In each case the volume is significantly less than the parent SrF2, 26% for the monoclinic phase and 12% for the orthorhombic phases. Surprisingly, neither structure resembles the only experimental literature data for ZrF2 which dates from 1964 and assesses ZrF2 to be stable up to 800 ○C. In contrast, both of our ZrF2 phases are significantly less stable (~1 eV/f.u.) than a phase-decomposed mixture of ZrF4 and Zr metal. While this disproportionation reaction may be kinetically hindered due to the extensive mass transport required, it does cast doubt on the reported stability of ZrF2 and the nature of the structure. This in turn has implications for the reliability of official reports on the Hanford 90SrF2 capsules which base their temperature and structural stability assessments on the 1964 article.
12:15 PM - A6.9
Atomistic Simulation of the Structural, Thermodynamic and Elastic Properties of Li2TiO3.
Samuel Murphy 1 2 , Philippe Zeller 2 , Alain Chartier 2 , Laurent Van Brutzel 2 , Robin Grimes 1
1 , Imperial College London, London United Kingdom, 2 , CEA, DEN, LM2T, Gif-sur-Yvette France
Show AbstractLithium based ceramics, such as lithium metatitanate, have been proposed for adoption in the breeder blanket region of a fusion reactor. In this presentation we report a combination of empirical and DFT simulations employing “on-the-fly” pseudo-potentials for Li2TiO3. The smoothing parameters of the plane-wave pseudo-potentials were optimised to ensure an appropriate level of precision for determination of structural, thermodynamic and elastic properties. As the elastic properties of lithium metatitanate are not well known the efficacy of the DFT simulations employing the new pseudo-potentials was explored using Li2O and TiO2 where experimental data is available. These pseudo-potentials are then used to investigate the three intermediate temperature phases of Li2TiO3 (ie. C2/c, C2/m and P3112). Next, we examine the elastic properties of Li2TiO3 using both DFT and an empirical potential model and find it to be, irrespective of space group, the most resistant to deformation of the materials suggested for adoption as a breeder material. Finally we examine defect species in Li2TiO2.
12:30 PM - **A6.10
Utilizing Radioelement Compound Synthesis for Nuclear Materials.
Ken Czerwinski 1 , Frederic Poineau 1 , William Kerlin 1 , Erik Johnstone 1 , Alfred Sattelberger 2 , Phil Weck 2 , Boris Narboux 3
1 Chemistry, University of Nevada, Las Vegas, Las Vegas, Nevada, United States, 2 , Argonne National Laboratory, Argonne, Illinois, United States, 3 , Chimie Paris Tech, Paris France
Show AbstractFor many nuclear applications synthetic routes are based on well established methods. In future fuel cycles the ability to exploit materials containing technetium and the actinides requires compound synthesis and characterization. Development of advanced fuels and waste forms can leverage innovative synthetic techniques that are utilized in the laboratory and non-nuclear industry. In particular methods that use novel reactions with common starting materials can be applied to produce fuels and waste forms with suitable attributes for advanced fuel cycles. High temperature and hydrothermal reactions with technetium are presented as a means to develop tailored waste forms. Reactions of elemental forms of technetium and halides in flowing gas and sealed tubes yielded a range of compounds. Many of these compounds represent new binary species. Characterization of the resulting compounds indicated trends with technetium oxidation state that can be exploited for waste forms, including species with iodine. Hydrothermal techniques have demonstrated the ability to produce low valent technetium compounds from pertechnetate starting species, a common form of technetium in solution. The ability to tailor the properties and morphologies of technetium metal and dimers is possible from variations in the hydrothermal techniques. This nuclear derived method can also impact new areas and is an instance of extending radioelement research into new domains. An example for nuclear fuels is provided based on the formation of uranium mononitride from dinitride starting material. Uranium dinitride is air stable and can be produced from oxide starting material. Uranium dinitride pellets can be formed in air and then sintered under inert atmosphere to produce uranium mononitride. The unique method for the nitride synthesis can be coupled with established sintering techniques to produce fuel. These waste form and fuel illustrations exemplify the utility synthesis reactions can play in the future fuel cycles.
A7: Corrosion
Session Chairs
Michael Fitzpatrick
Djamel Kaoumi
Wednesday PM, November 30, 2011
Independence W (Sheraton)
2:30 PM - A7.1
Atomistic Simulations of Thermodynamics of CRUD Deposition.
Avinash Dongare 2 , Christopher O'Brien 2 , Jacob Eapen 1 , Donald Brenner 2
2 Materials Science and Engineering, NC State University, Raleigh, North Carolina, United States, 1 Nuclear Engineering, NC State University, Raleigh, North Carolina, United States
Show AbstractCRUD deposition on fuel clad surface, which is accelerated by subcooled nucleate boiling, results in anomalous flux depressions (AOAs). Structural alloys chemically react with coolant at high temperatures which results in diffusion of oxygen into the base metal. Attendant release of metal ions into the water facilitates the growth of CRUD layers on the outer clad surface. A mechanistic description of the deposition of CRUD requires a fundamental understanding of the driving force for deposition of the relevant CRUD species (most often a class of mixed nickel–ferrite oxides). We evaluate the Gibbs free energy of solvation (ΔGsolv) of different CRUD species in water through electronic structure and atomistic simulations. The stability of the solvated species is then assessed through the Gibbs free energy.
2:45 PM - A7.2
Hydrolysis of Hydroxamic Acid Complexants in the Presence of Non-Oxidizing Metal Ions.
Scott Edwards 1 , Fabrice Andrieux 1 , Colin Boxall 1 , Robin Taylor 2
1 , Lancaster University, Lancaster United Kingdom, 2 , National Nuclear Laboratory, Seascale United Kingdom
Show AbstractAll commercial nuclear fuel reprocessing plants use the hydrometallurgical PUREX process to separate U and Pu from used nuclear fuel. These are recycled as new fuels whilst the remaining highly radioactive liquid containing, inter alia, fission products and minor actinides, is calcined into glass and stored pending disposal. Substantial international effort is being expended to improve on PUREX for GenIV fast reactor fuel cycles. Whilst this will be reviewed in detail in this paper, the drivers for this may be summarised to include: improving its proliferation resistance; management of the highly radiotoxic minor actinides; reducing the heavy metal disposal burden. With regard to the latter, the first stage of PUREX extracts 4 and 6 valent actinides using tributyl phosphate (TBP) followed by selective stripping of Pu and Np. Reductants such as Fe(II) sulfamate are used to control extraction by conditioning actinide redox states. This, however, adds undesirable inorganic salts to the high-level aqueous waste.The last 10 years has seen significant research into modified PUREX schemes such as UREX+, NUEX and Advanced PUREX, centring round the use of simple hydroxamic acids (XHAs) such as acetohydroxamic acid (AHA). XHAs are salt-free organics that, due to their capacities for complexing hard cations and participating in redox, can effectively strip Np(VI) and Pu(IV) from TBP. U(VI) cations are unaffected and remain in the organic phase. This allows for the generation of a high purity bulk U product and co-processed U/Pu product. XHA hydrolysis as the free or metal-bound ligand is known to occur and may have negative implications in advanced nuclear fuel processing. There is a lack of data pertaining to hydrolysis of bound XHAs. Using the Fe(III)-AHA system as a non-active surrogate system, we have developed a model that allows for the study of the hydrolysis of both 1:1 and 2:1 AHA-Fe(III) complexes in aqueous solution. Application of this model to data recorded over 293–323K in the presence of the 1:1 complex allows determination of the activation energy EA = 100 kJ mol-1 and collision factor A = 2.03 × 1010 dm3 mol-1 s-1 for complex hydrolysis. Studies of solutions containing both 1:1 and 2:1 complexes over 293–323K suggest values of EA=92.0 kJ mol-1 and A = 6.30 × 1011 dm3 mol-1 s-1 for the hydrolysis of the 2:1 complex. The lower EA and higher collision factor of the 2:1 complex implies greater susceptibility to hydrolysis than the 1:1 complex.Using this model, analysis of analogous data pertaining to the Np(IV)-AHA system shows that hydrolysis of the 1:1 and 2:1 complexes at 293K proceeds with rate parameters of k1 = 1.93 × 10-4 dm3 mol-1 s-1 and k2 = 3.76 × 10-4 dm3 mol-1 s-1. Thus, as well as an overview of PUREX, this paper will also discuss the development of our modified PUREX model, and consider the implications of our findings for the GenIV fuel cycle requirements for this widely used separations scheme.
3:00 PM - A7.3
First-Principles Assessment of the Boric Acid Reaction Mechanisms on NiO (001) and ZrO2 (-111) Surfaces.
Priyank Kumar 1 , Michael Short 2 , Sidney Yip 2 , Jeffrey Grossman 1 , Bilge Yildiz 2
1 Materials Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States, 2 Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States
Show AbstractBoron incorporation into the corrosion-related deposits (CRUD) on nuclear fuel cladding can lead to power distribution problems, termed as the Axial Offset Anomaly (AOA), which poses safety and operation concerns in light water reactors. The mechanisms and kinetics that lead to boron incorporation into CRUD thus far are described only macroscopically. An understanding of these processes at a fundamental level would allow for the assessment of boron build-up rates, enabling novel surface chemistry design strategies against boron trapping. The present study investigates the adsorption and dissociation reaction pathways of boric acid, B(OH)3, and the reaction kinetic descriptors on the stable surfaces of two such CRUD oxides, NiO (001) and ZrO2 (-111), using first principles methods based on density functional theory and ab-initio molecular dynamics. Strong electron correlations in the case of NiO are included using the DFT + U approach. Adsorption of boric acid on clean ZrO2 (-111) is found to be more favorable compared to that on NiO (001), in agreement with prior experiments. Dissociative adsorption is observed to dominate over molecular adsorption in the case of ZrO2, while NiO (001) favors molecular adsorption. The most stable configuration for B(OH)3 on NiO (001) is a hydrogen-bonded molecular structure, Nis-(OH)B(OH)(OH)...Os, with a binding energy of 0.74 eV, while on ZrO2 (-111), a single O-H dissociated structure, Zrs-(O)B(OH)(HO)-Zrs + Os-H, with a binding energy of 1.61 eV is the most stable. The kinetics of dissociation reactions is being explored through the Nudged Elastic Band (NEB) calculations. We demonstrate that, taken together, such computational studies can help understand boron deposition mechanisms, predict oxides that can significantly repel boron and provide crucial guidance to better control of coolant chemistry.
3:15 PM - A7.4
Mechanisms of Fixed Contamination of Commonly Engineered Surfaces.
Rebecca Woodhouse 1 , Colin Boxall 1 , Richard Wilbraham 1
1 Engineering, Lancaster University, Lancaster United Kingdom
Show AbstractDuring the lifetime of a nuclear facility, radioactive material may become deposited onto commonly engineered process surfaces. A large quantity of this ‘plate out’ material will be loosely bound by weak electrostatic interactions. However, some material will be rigidly bound e.g. by being held within oxide layers formed over many years of use, or, in the case of highly concentrated acidic liquors, metal-oxide layers that have formed in-situ, molecularly binding radioactive material to the solid surface. During decontamination, these radioactive deposits are removed by (electro-) chemical or mechanical techniques. Mechanical techniques, such as water jets or abrasives, are generally used on structural materials (concrete). Chemical techniques such as use of mineral acids, and electrochemical techniques such as electropolishing using nitric acid as an electrolyte are used to decontaminate metallic surfaces, including steels, which are commonly used on nuclear sites.These techniques are extremely destructive to not only the surface layer but also to the underlying material, creating large volumes of secondary waste. A greater understanding of the mechanisms of how contaminant radionuclides interact with and attach to process engineering materials in nuclear plant environments is required, enabling informed decisions to be made about the most effective application of decontamination techniques, reducing secondary waste output. There is limited literature relating to radionuclide sorption mechanisms on engineering metals. Key studies have found that sorbed contamination is almost entirely located in the outermost 0.5-1µm of the oxide layer formed at the steel surface. Thus, a molecular level investigation of contaminant uptake during induced oxide growth would be greatly assistive to decontamination strategy design.We describe work carried out on electrochemically accelerated oxide growth on, and corrosion of, steels 304L, 316L, 347, SS2343 (a 316L analog) and the Ni-Cr-Mo alloy, hastelloy in nitric acid – all commonly employed in the nuclear industry in process streams and pipework. Combined potentiodynamic and electrochemical impedance spectroscopy (EIS) measurements allow for the identification of active, passive, high voltage passive, transpassive and secondary passivation regimes. Corrosion potential measurements indicate the participation in the overall corrosion process of an autocatalytic reduction of nitric acid at [HNO3] > 4.5 mol dm-3, a reaction suppressed under flow conditions. EIS indicates that net growth of surface oxide is found to occur in the high voltage passive regime, an area we will use for the study of surrogate contaminant uptake. In this regime we have directly measured rates of oxide layer growth by microgravimetric means using SS2343 coated quartz crystal microbalance piezoelectrodes and correlated these with both CV and Raman spectroscopy measurements to determine layer composition and mode of contaminant uptake.
3:30 PM - **A7.5
Mechanisms of Oxidation of Fuel Cladding Alloys Revealed by High Resolution APT and TEM Analysis.
Chris Grovenor 1 , Na Ni 2 , Daniel Hudson 1 , George Smith 1 , Sergio Lozano-Perez 1 , John Sykes 1
1 Materials, Oxford University, Oxford United Kingdom, 2 Department of Materials, Imperial College , London United Kingdom
Show AbstractZirconium alloys were selected for use as nuclear fuel cladding and structural fuel assembly components in nuclear reactors in the early 1950s because of an attractive combination of low thermal neutron absorption cross-section, adequate corrosion resistance in high temperature water and reasonable mechanical properties. However, aqueous corrosion and hydrogenation of zirconium alloys have now become the major factors limiting the high fuel burn-up or prolonged fuel campaigns in nuclear plant, and there are many groups working on the fundamental mechanisms that control these processes. It is reasonably well established that the oxidation rate is governed by oxygen anion diffusion through a “barrier layer” at the metal-oxide interface but there is still considerable confusion as to the nature of this barrier layer. Studies using SEM, TEM and electrochemical impedance measurements have been interpreted as showing a dense inner-most oxide layer, and an increased thickness of the layer has been correlated to a better corrosion resistance. It has also been inferred that its thickness varies with the stage of oxidation process. Many authors have reported, usually from TEM studies, that an ‘intermediate layer’ at the metal oxide interface has a complex structure or/and stochiometry different to that of both the bulk oxide and bulk metal, sometimes claimed to be a suboxide phase. Diffraction evidence has suggested the presence of both cubic ZrO and rhombohedral Zr3O phases, and compositional analysis has revealed similar variations in local oxygen stoichiometry. In part the difficulty in building up a clear picture of what is present at the interface at any particular stage of the complex oxidation process characteristic of commercial zirconium alloys is because the observations represent isolated ‘snapshots’ in time of a phenomenon that all the experimental evidence shows to be highly dynamic. We have been carrying out a systematic investigation of the structure and chemistry of the metal/oxide interface in samples of commercial ZIRLO corroded for many different times up to 180 days. We have developed new experimental techniques for the study of these interfaces both by Electron Energy Loss Spectroscopy (EELS) analysis in the Transmission Electron Microscope (TEM) and by Atom Probe Tomography (APT), and exactly the same samples have been investigated by both techniques. Our results show the development as the initial oxidation rate slows down of a clearly defined suboxide layer of stoichiometry close to ZrO, and the subsequent disappearance of this layer at the first of the characteristic ‘breakaway’ transitions in the oxidation kinetics. We can correlate this behaviour with changes in the structure of the oxide layer, and particularly the development of interconnected porosity that links the corroding interface with the aqueous environment.
4:30 PM - A7.6
Modeling the Speciation of Actinide Ions in Aqueous Solution and Interfaces.
Raymond Atta-Fynn 1 , Donald Johnson 1 , Eric Bylaska 1 , Gregory Schenter 1 , Wibe De Jong 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractA thorough understanding of the speciation of actinide ions in aqueous solution and mineral/aqueous interfaces is of significant importance to the long-term storage, disposal, and management of high level nuclear waste. In this regard, there is a plethora of experimental works but only a handful of theoretical work. This talk focuses on the application of computer simulations to model the speciation of the Curium(III), Uranium(IV), Uranium(V), and Uranium(VI) ions in aqueous solution and iron oxide surfaces using quantum molecular dynamics, classical molecular dynamics, and mixed quantum/classical molecular dynamics. The simulated coordination shell structures, hydrolysis constants, and EXAFS spectra will be compared to available experimental data. Predictions on the hydrogen bonding network, electronic structure, and atomic scale ligand-ion dynamical properties such as mean residence times and exchange mechanisms will also be presented.
4:45 PM - A7.7
Surface Decontamination by Photocatalysis.
Richard Wilbraham 1 , Colin Boxall 1 , Robin Taylor 2
1 Engineering Department, Lancaster University, Lancaster, Lancashire, United Kingdom, 2 B170 Central Laboratory, National Nuclear Laboratory, Seascale, Cumbria, United Kingdom
Show AbstractCurrently in the nuclear industry, surface contamination in the form of radioactive metal or metal oxide deposits is most commonly removed by chemical decontamination, electrochemical decontamination or physical attrition. Physical attrition techniques are generally used on structural materials (concrete, plaster), with (electro)chemical methods being used to decontaminate metallic or painted surfaces. The most common types of (electro)chemical decontamination are the use of simple mineral acids such as nitric acid or cerium (IV) oxidation (MEDOC). Use of both of these reagents frequently results in the dissolution of a layer of the substrate surface increasing the percentage of secondary waste which leads to burdens on downstream effluent treatment and waste management plants. In this context, both mineral acids and MEDOC can be indiscriminate in the surfaces attacked during deployment, e.g. attacking in transit through a pipe system to the site of contamination resulting in both diminished effect of the decontaminating reagent upon arrival at its target site and an increased secondary waste management requirement. This provides two main requirements for a more ideal decontamination reagent: Improved area specificity and a dissolution power equal to or greater than the previously mentioned current decontaminants.Photochemically promoted processes may provide such a decontamination technique. The use of photocatalytic nanoparticles to drive the photochemical reduction of metal ion valence states to aid in heavy metal deposition has already been extensively studied, with reductive manipulation also being achieved with uranium and plutonium simulants (Ce). Importantly photooxidation of a variety of solution phase metals, including neptunium, has also been achieved. Here we briefly review existing decontamination techniques and report on the potential application of photocatalytic nanoparticle promoted oxidation technologies to metal dissolution (including process steels) and to the dissolution of adsorbed actinide contaminants.
5:00 PM - A7.8
Investigation of Water Adsorption on Metal Oxide Surfaces under Conditions Representative of PuO2 Storage Containers.
Patrick Murphy 1 , Colin Boxall 1 , Robin Taylor 2
1 Engineering, Lancaster University, Lancaster, Lancashire, United Kingdom, 2 , National Nuclear Laboratory, Seascale, Cumbria, United Kingdom
Show AbstractStandardised packaging and storage of plutonium oxide powders involves sealing the materials in welded, stainless steel containers. Pressurization of these containers arises from decomposition of adsorbed water contained in and on the surface of hygroscopic PuO2 (1). In an attempt to remove water from the surfaces of these oxides, and thus minimise pressurisation, PuO2 samples are calcined at temperatures as high as 700°C (2). The potential of PuO2 to generate a water vapour derived pressure in a storage can headspace is directly related to its capacity for H2O adsorption. Water adsorption on PuO2 has previously been investigated by measuring headspace pressure, as a function of temperature within a closed system containing a fixed quantity of PuO2 in the presence of varying amounts deliberately added water (1). This involves making a number of assumptions relating to the PVT behaviour of the headspace of the closed system, usually based on the behaviour of an ideal gas, in order to estimate the mass of water adsorbed at the PuO2 surface.We have developed a QCM (Quartz Crystal Microbalance) based method for direct gravimetric determination of water adsorption on PuO2 surrogate surfaces, especially CeO2, under conditions representative of those in a typical PuO2 storage can. In this application, the method of transduction of the QCM relies upon the linear relationship between the resonant frequency of piezoelectrically active quartz crystals and the mass adsorbed on the crystal surface. In our system, using sol gel and spatially constrained hydrolysis methods, we have coated QCM crystals with a PuO2 surrogate so as to measure directly the mass of water adsorbing at the surrogate surface. However, as well as responding to adsorbed mass, quartz crystals also respond to changes in temperature, exhibiting changes in both resonant frequency and mass sensitivity. For mass changes in a system with a dynamically changing temperature regime, this thermal effect can be compensated for by modelling the temperature dependence of the frequency response of a QCM crystal in the absence of water (3). By extension, this approach can also be used for modelling the temperature response of PuO2 surrogate-coated crystals in the absence of water adsorption, so allowing for the gravimetric measurement of water adsorption at the surrogate surface as a function of temperature. Water adsorption isotherms for CeO2 have been generated to high temperatures, allowing for the extraction of key adsorption isotherm parameters. Gravimetric data will be used to compliment PVT data for water adsorption on PuO2 generated at the UK National Nuclear Laboratory. 1.M. T. Paffett et al, J. Nuc. Mat., 322 (2003) 452.US Department of Energy, DOE/DP-0123T (1994)3.D. Wang et al, Colloids and Surfaces, 268 (2005) 30
5:15 PM - A7.9
Thermochemistry of Monomeric and Nanocluster Alkali Uranyl Peroxides.
Christopher Armstrong 1 , May Nyman 2 , Tatiana Shvareva 1 , Ginger Sigmon 3 , Peter Burns 3 , Alexandra Navrotsky 1
1 , UC Davis, Davis, California, United States, 2 , Sandia National Laboratories, Albuquerque, New Mexico, United States, 3 , University of Notre Dame, South Bend, Indiana, United States
Show AbstractUsing high temperature oxide melt solution calorimetry, enthalpies of formation from the oxides, ΔHf,ox, and from the elements, ΔHf,el, were obtained for isostructural alkali uranyl peroxide phases (M4[UO2(O2)3](H2O)9; M = Li, Na, K) and for a nanocluster compound, the U60 with fullerene topology (Li40K20[UO2(O2)(OH)]60(H2O)214). In contrast to the mineral studtite, (UO2)O2(H2O)4, which is stable only in the presence of peroxide in solution, ΔHf,ox for all of these materials is highly exothermic, with the sodium-bearing phase being the most stable (ΔHf,ox = -515.5 ± 8.9 kJ mol-1, see Table 1). In addition, the formation of the U60 nanosphere from studtite and the alkali monomers ((Li4[UO2(O2)3](H2O)9 and K4[UO2(O2)3](H2O)9) is thermodynamically favorable, suggesting that the monomers may be viable building blocks of U cluster materials in the solid state. Current work is focused on modeling and predicting stability relationships between alkali peroxide monomers, clusters, and common uranyl containing minerals under aqueous conditions.
5:30 PM - A7.10
X-Ray Absorption Spectroscopy Study of Co2+ Ion Adsorption on Fe3O4 Nanoparticles in Supercritical Aqueous Fluids.
Hao Yan 1 , Robert Mayanovic 1 , Alan Anderson 2 , Joseph Demster 1 , Peter Meredith 2
1 Dept. of Physics, Astronomy & Materials Science, Missouri State University, Springfield, Missouri, United States, 2 Department of Earth Sciences, St. Francis Xavier University, Antigonish, Nova Scotia, Canada
Show AbstractThe reactivity of ferrite (Fe3O4) nanoparticles with cobalt ions under hydrothermal conditions was studied using in-situ x-ray absorption spectroscopy (XAS). Co K-edge XAS spectra were measured from Fe3O4 nanoparticles (~24 nm in diameter) in a 0.05 m Co(NO3)2 aqueous solution to 500 °C. The XAS measurements were made in fluorescence mode using our modified hydrothermal diamond anvil cell on the ID20-B beam line at the Advanced Photon Source. The XAS data show that the Co2+ ions begin to react with Fe3O4 nanoparticles at 250 °C, which is at least 100 °C lower than the reaction of Ni2+ or Zn2+ ions with Fe3O4 nanoparticles. The pre-edge feature of the x-ray absorption near edge structure (XANES) data exhibits a single peak from 25 to 200 °C, indicating that the Co2+ aqua ion species predominates in the aqueous system in this temperature range. Conversely, the pre-edge of the XANES exhibits a double-peaked feature from 250 to 500 °C, indicating adsorption of the Co2+ ion on the surface of Fe3O4 nanoparticles in the near and supercritical region of the aqueous solution. The pre-edge peaks are associated with dipole-allowed transitions of Co 1s-electrons to unoccupied states having mixed 3d (Co) and p (Co, O) character. The double-peaked feature most likely reflects the crystal field and/or molecular-orbital splitting of the Co 3d orbital sub-bands, upon reaction and subsequent bonding of the Co2+ ion with the Fe3O4 nanoparticles. A detailed analysis was made of the background-subtracted pre-edge feature using Gaussian fitting. Both the lower energy (A1) and the higher energy (A2) peaks of the pre-edge feature are shifted to lower x-ray-absorption energy with increasing temperatures. After the onset of adsorption of the Co2+ ions on Fe3O4 nanoparticles, the A1 and A2 peaks increase monotonically in area with increasing temperatures. The increasing size of the A1 and A2 peaks is correlated with a reduction in electronic population of the 3d orbitals associated with the electronic 1s transitions. Analysis of extended x-ray absorption fine structure (EXAFS) using a spinel-motif structure model shows that the Co2+ ion adsorbs preferentially on octahedral sites of the surface of the nanoparticles in aqueous fluids above 250 °C. The Fourier transforms of the EXAFS and the XANES of the Co-Fe3O4 nanoparticle aqueous system at high temperatures (250 – 500 °C) are qualitatively consistent with those of the CoFe2O4 bulk material measured at room temperature.
A8: Poster Session: Material Challenges in Current and Nuclear Fuel Technologies
Session Chairs
Majorie Bertolus
Robin Grimes
Blas Uberuaga
Karl Whittle
Thursday AM, December 01, 2011
Exhibition Hall C (Hynes)
9:00 PM - A8.1
Fundamental Insights into Fission Product Sorption on Graphite.
Alejandro Londono-Hurtado 1 , Dane Morgan 1 , Izabela Szlufarska 1
1 Materials Science and Engineering, University of Wisconsin-Madison, Madison, Wisconsin, United States
Show AbstractUnderstanding of sorption of radioactive fission products to nuclear graphites is essential for prediction of release rates from fuel particles that may occur during accident conditions. One challenge in evaluating sorptivities is that different grades of nuclear graphites exhibit large variations in the microstructures (e.g., porosity, degree of graphitization) and these microstructures have often not been characterized in great detail. In addition sorption mechanisms in graphites are currently unknown, which makes it challenging to predict sorption isotherms even if the detailed structural information was available. Here we present a systematic study of the sorption behavior of two fission products, Sr and Cs, in different carbon-based structures. Density functional theory is employed to calculate trends in binding energies on pristine and defected graphite, amorphous carbon, and diamond. These materials were chosen to cover the range of most common hybridizations of carbon. Our study shows that fission products bind strongly to sp3 and amorphous structures, while binding to pristine graphite is relatively weak. In the presence of defects, binding energies in graphite become stronger, which is consistent with the experimental observation that radiation increases fission product sorptivity. Calculated binding energies to certain types of sites show good agreement with those extracted from experimental sorption isotherms, allowing identification of likely sorption sites. We will also present analysis of trends in graphite sorptivities in terms of surface area and degree of graphitization, which trends have been extracted from published in experimental data.
9:00 PM - A8.10
Low Temperature Sintering (930-1050°C) of SiC for Inert Matrix Fuels Using a Polymer Precursor.
Roberto Esquivel 1 , Donald Moore 1 , Juan Nino 1
1 , University of Florida, Gainesville, Florida, United States
Show AbstractA polymer precursor route was followed in order to sinter β-SiC at 930 and 1050°C. The thermal analysis (thermogravimetric and differential thermal analysis), Raman spectroscopy and X-ray diffraction of allylhydridopolycarbosilane confirmed the formation of amorphous SiC at 930°C. Fired pellets made of 90 wt% polycrystalline β-SiC powder and 10wt% polymer precursor, showed a higher average density of 2.42 g/cm3 (77% of theoretical), by utilizing a binary mixture of fine and coarse particles. The hardness, fracture strength and fracture toughness were determined. For the 1050°C fired samples, values of 4.38 GPa, 93.3 MPa and 2.63 MPa-m(1/2) were obtained, respectively. These results were compared with those of UO2, MOX (mixed oxide) and other inert matrix material candidates for light water reactors. The thermal diffusivity, thermal conductivity and heat capacity were measured from 100 to 900°C. The thermal conductivity was found to be higher than UO2 and MOX in all ranges of temperature, and higher than other candidates at elevated temperatures. The obtained density and measured thermophysical and mechanical properties make this process suitable for an inert matrix fuel open porosity concept.
9:00 PM - A8.11
Identifying the Energetics of He-Point Defect Interactions in Fe through Coordinated Experimental and Modeling Studies of He Ion Implanted Single Crystal Fe.
Xunxiang Hu 1 , Donghua Xu 2 , Brian Wirth 2
1 Nuclear Engineering, University of California, Berkeley, Berkeley, California, United States, 2 Nuclear Engineering, University of Tennessee, Knoxville, Tennessee, United States
Show AbstractHelium effects on the microstructural evolution and mechanical properties of structural materials are among the most challenging issues facing fusion materials research. In this work, we combine thermal helium desorption spectroscopy (THDS) with positron annihilation spectroscopy and a spatially-dependent cluster dynamics model to investigate the kinetics of helium-point defect interactions of helium implanted single-crystalline iron. The combination of the modeling and thermal desorption measurements allows identification of possible mechanisms (e.g., shrinkage of He3V2 clusters) responsible for measured Helium desorption peaks. Furthermore, the model predicts the depth dependence of the Helium and Helium – defect clusters as a function of time and temperature during the desorption measurement. This provides an opportunity for additional microstructural characterization methods to ensure the self-consistency of the modeling predictions with the experimental results and validation of the identification of helium – point defect interactions. In this work, we have thus combined the THDS measurements as a function of He implantation energy in the range of 10-40 keV with fluence levels of 1E14 to 1E15 He/cm2 and as a function of the time-temperature annealing protocol, with select positron annihilation spectroscopy measurements at a variety of material conditions. The positron lifetime measurements provide an indication of the He-vacancy cluster size. This presentation will provide an overview of the self-consistency of the model predictions based on the assumptions made in terms of the He-point defect binding and interaction energies, and diffusivities.
9:00 PM - A8.12
Bonding Strength of Tungsten Coating on the First Wall Structure Material.
Manabu Satou 1
1 , Hachinohe Institute of Technology, Hachinohe Japan
Show AbstractTungsten is one of the candidates as the first-wall plasma facing material for a magnetic confinement fusion power reactor. One of the important properties of the candidate structure materials is joining or bonding to the first-wall material to maintain their appropriate properties such as high-heat load capability and strength. Bonding strength between the structure material and tungsten layer was evaluated by a laser shock spallation method, which uses a pulse laser to generate a shock wave to create tensile stress inside the specimen. The prepared tungsten layer of quantitative strength is discussed from a viewpoint of their application to the first-wall.
9:00 PM - A8.13
Fuel Swelling Due to Dopant and Fission Product Concentration.
Simon Middleburgh 1 2 , Robin Grimes 1 , Samuel Murphy 1 , Paul Blair 2
1 Department of Materials, Imperial College London, London United Kingdom, 2 Materials and Fuel Rod Design, Westinghouse Electric Sweden, Vasteras Sweden
Show AbstractCalculations using empirical interatomic potentails have been used to understand the effect of various fission products and dopants on the UO2 lattice parameter. Results suggest that the likely accommodation mechanism for divalent and trivalent cations in UO2 is to create an oxidized uranium ion cluster {2UU●:DU''} for divalent ions and {UU●:TU'} for the trivalent ions. Small ions such as Cr3+ and Mg2+ ions have the largest contracting effect on the UO2 lattice whilst larger ions, La3+ and Ba2+ ions have the largest expansive effect on the lattice, although the majority of extrinsic species studied contract the lattice.Extrinsic tetravalent cations, substituting onto the uranium sublattice have also been calculated for and a simple relationship between cation size and the effect on the UO2 crystal volume has been suggested.
9:00 PM - A8.14
OFET Sensors with Poly 3-Hexylthiophene and Pentacene as Total Dose Detectors for Ionizing Radiation.
Harshil Raval 1 , V. Ramgopal Rao 1
1 Centre of Excellence in Nanoelectronics, Department of Electrical Engineering, Indian Institute of Technology Bombay, Mumbai, Maharashtra, India
Show AbstractPentacene and poly 3-hexylthiophene (P3HT) are the most promising p-type organic semiconducting materials for fabrication of organic field effect transistors (OFETs). OFETs with aforesaid organic semiconducting materials have been demonstrated as sensors for ionizing radiation, wherein the changes in the electrical characteristic parameters were used as a measure of ionizing radiation dose. P3HT based OFET sensor has shown OFF current sensitivity of 0.5 pA/Gy/1µm-width of the sensor while pentacene based OFET sensor has shown an OFF current sensitivity of 1.4 pA/Gy/1µm-width of the sensor. Changes in the electric band structures of the P3HT and Pentacene upon exposure to ionizing radiation have been observed using UV-visible spectroscopy, X-ray photoelectron spectroscopy (XPS) and photoluminescence study. UV-visible spectroscopy and photoluminescence study of P3HT and pentacene thin-films exposed to increasing doses of ionizing radiation have shown generation of new oxidized states with different electric band structure resulting in increased conductivity of the materials. The results were also studied with XPS of the thin-films of the materials. Role of interface between gate dielectric and organic semiconductor has been studied for linear response from the sensors considering the interface trap generation due to ionizing radiation. Increasing number of interface traps have been extracted from the changes in subthreshold swing and their role in the linearity of response of the sensors for identical doses of ionizing radiation is investigated.
9:00 PM - A8.15
Copper Phthalocyanine Based Field Effect Transistors as Total Dose Sensors for Determining Ionizing Radiation Dose.
Harshil Raval 1 , V. Ramgopal Rao 1
1 Centre of Excellence in Nanoelectronics, Department of Electrical Engineering, Indian Institute of Technology Bombay, Mumbai, Maharashtra, India
Show AbstractWell known organic semiconducting materials - phthalocyanines and their derivatives have been explored for numerous sensing applications as electrical properties of Phthalocyanine thin-films change with their exposure to oxidizing or reducing gases or environment. However, in practical applications air-stability of these materials is a serious concern. In this work, use of copper phthalocyanine (CuPc) thin-film with a thin layer of silicon nitride (~ 50 nm), deposited at room temperature using hot-wire CVD process, as a passivation layer, is proposed for determining ionizing radiation doses. Silicon nitride passivation has shown more than 200% reduction for thin-film degradation for ambient conditions. Effect of ionizing radiation on the electrical properties of thin-films of CuPc have been studied by observing a change in resistance of the film which resulted in a 110X change for a total of 20 Gy radiation dose using Cobalt-60 (60Co) radiation source. Moreover, significant changes in the electrical characteristics of an organic field effect transistor (OFET), with CuPc as an organic semiconductor, have been observed with increasing doses of ionizing radiation. Surface morphology of CuPc thin-films on the substrates have been studied using AFM. Generations of oxidized states in the thin-films of CuPc have been studied using UV-visible spectroscopy for increasing doses of ionizing radiation. Experiments with an OFET (W/L = 19350 µm / 100 µm and tox = 100 nm) as a sensor resulted in ~110X change in OFF current for a total of 20 Gy dose of ionizing radiation providing 0.8 nA/Gy of sensitivity. Moreover, CuPc based OFET sensors have various advantages such as low cost of fabrication on flexible substrates and large area coverage. Also, changes in various other characteristic parameters such as ON current, current ratio, subthreshold swing, field effect mobility, increasing number of interface states, shift in threshold voltage etc. extracted from the electrical characterizations of the irradiated OFET proved it a better choice for determining ionizing radiation employing simple technique of electrical measurement.
9:00 PM - A8.16
Temperature and Time-Dependent Ion Diffusion across Stainless Steel-Zirconium Film Interface Studied by Scanning Surface Potential.
Marilyn Hawley 1 , Kendall Hollis 1 , Pat Dickerson 1
1 Materials Science and Technology, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThis work was driven by the need to fabricate a high-density diffusion barrier coating for the development of new low-enriched uranium based fuel reactors. This thin diffusion barrier serves two purposes: to protect the fuel from directly interacting directly with the 6061 aluminum cladding, which could result in swelling and cracking of the fuel, and to prevent accumulation of fission generated gases. As part of this effort, thin 304 stainless steel foil substrates were used as a surrogate for the fuel material. Zirconium films, grown by plasma spraying, were used for the diffusion barriers. A reverse polarity current applied to the substrate continuously cleaned the surface and heated the substrate. The zirconium films were applied sequentially, first to one side of the substrate followed by deposition to the second side. This resulted in the first applied film being experiencing twice the heating time as long as the second film. Two different currents were used for the samples studied here. Samples were studied in cross section and the scanning probe images revealed an increasing grain size from the interface to the outer film surface. However, 304 stainless steel, unlike uranium oxide, contains a number of different chemical species (Fe, Ni, Cr, and Mn in decreasing amounts). The surface potential images clearly showed evidence of major diffusion from the stainless steel through the zirconium film. Further, the appearance of an additional interface within the film indicated different rates of diffusion for difference ions. Higher temperature and time results correlated with greater diffusion distances. Corresponding atom probe data, collected on the same sample, will be presented for comparison with the surface potential results.
9:00 PM - A8.17
Ion Irradiation Damage in Glassy Polymeric Carbon for TRISO Fuel Application.
Malek Abunaemeh 1 2 , Mohamed Seif 3 , Daryush Ila 4
1 Physics, Alabama A&M University, Madison, Alabama, United States, 2 Center for Irradiation of Materials, Alabama A&M University Reseach Instiute, Normal, Alabama, United States, 3 Mechanical Engineering, Alabama A&M University Reseach Instiute, Normal, Alabama, United States, 4 Reseach, North Carolina Fayetteville State University, Fayetteville , North Carolina, United States
Show AbstractThe TRISO fuel that is planned to be used in the Generation IV nuclear reactor consists of a fuel kernel of UOx coated in several layers of materials with different functions. We are looking at the ion irradiation induced structural modifications of the glassy polymeric carbon (GPC) microstructure and their effect on the mechanical and physical properties. We irradiated GPC samples with 5 MeV Ag and Au ions. During the nuclear fission of 235U, the fission fragment mass distribution has two maxima around 98 and 137 that would best fit Rb and Cs However, both ions are hard to produce from our SNICS source therefore we chose Ag (107 amu) and Au (197 amu) as best replacements. A look at the Young’s Modulus and the Hardness along with the transmission electron microscopy (TEM) of the GPC samples before and after Irradiation will lead the consideration for GPC to become a potential replacement for the pyrolytic carbon coatings, with a function of diffusion barrier for the fission products
9:00 PM - A8.2
Swift Heavy Ion Induced Modifications in Gd2Ce2O7.
Maulik Patel 1 , Jonghan Won 1 , James Valdez 1 , Jean-Claude Pivin 2 , Isabelle Monnet 3 , Kurt Sickafus 1
1 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , Centre De Spectrometrie Nucleaire Et De Spectrometrie De Masse , Orsay France, 3 , Centre de Recherche sur les Ions, les Matériaux et la Photonique, Caen France
Show AbstractThe radiation stability of fluorite derivative complex oxides, such as pyrochlore and bixbyite, have received considerable attention [1, 2], with a futuristic view to using them as nuclear waste forms or as inert matrix fuels [3]. These applications impact the disposal or transmutation of actinides such as 239Pu. In order to understand irradiation-induced structural modifications in materials containing 239Pu, Ce is often used as a surrogate for Pu. Also, fuel and waste form radiation damage effects are often simulated using energetic ion irradiation.In the present work, a complex Ce-bearing oxide, Gd2Ce2O7, was synthesized in order to simulate Pu4+ in a fluorite derivative oxide lattice. X-ray diffraction (XRD) analysis using Rietveld refinement of the pristine material showed that Gd2Ce2O7 crystallizes in a, bixbyite (C-type cubic) crystal structure (Ia-3). In order to simulate crystal structure damage due to fission fragments, we performed swift heavy ion (SHI) irradiations of Gd2Ce2O7 using 92 MeV Xe to various ion fluences, ranging from 1011 - 1014 Xe/cm2. XRD analyses of these samples revealed an irradiation-induced transformation from the C-type cubic bixbyite (Ia-3) structure to a cubic, disordered fluorite (Fm-3m) structure. This order-to-disorder (O-D) phase transformation was confirmed based on transmission electron microscopy (TEM) observations. A more detailed analysis of the structure of a single ion track is currently being carried out using TEM. Cross-sectional Raman spectroscopic measurements are also being performed in order to better understand the nature of the O-D transformation, especially as a function of penetration depth of the incident energetic ion.
9:00 PM - A8.3
Development of High Performance Ni-Base Superalloys for Innovative Reactor Systems.
Y. Gu 1
1 , National Institute for Materials Science, Tsukuba, Ibaraki Japan
Show AbstractNew materials are key for optimizing Generation II and III light water reactors and as well as to meet Generation IV nuclear system requirements. High performance Ni-base superalloy is a promising class of structure materials for fuel cladding, core structures, reactor cooling systems and components, and power conversion systems in innovative reactor systems.The High Temperature Materials Center (HTMC) in the National Institute for Materials Science (NIMS) has been leading the research and development of wrought superalloys for the applications beyond 700 degree centigrade, which are now expected as key-materials in supercritical water reactor (SCRW). The new generation of wrought superalloys, a kind of nickel-coble-base superalloys processed by a normal cast and wrought (C&W) route and named as TMW alloy, can withstand temperature in excess of 725 degree centigrade, a 50-degree increase over C&W alloy U720LI currently in operation.In this presentation, we show the design idea, workability and properties of these TMW superalloys. Furthermore, the evaluation of the processing and microstructure on a full-scale component made of TMW superalloys are described, which demonstrated the advantages and possibility of the TMW superalloys at the component level.
9:00 PM - A8.4
Atomistic Simulation of Radiation Damage in β-Eucryptite (LiAlSiO4).
Badri Narayanan 1 , Ivar Reimanis 1 , Hanchen Huang 2 , Cristian Ciobanu 3
1 Department of Metallurgical and Materials Engineering, Colorado School of Mines, Golden , Colorado, United States, 2 Department of Mechanical Engineering, University of Connecticut, Storrs, Connecticut, United States, 3 Division of Engineering, Colorado School of Mines, Golden, Colorado, United States
Show AbstractLithium aluminum silicates are an important class of engineering materials mainly because to their low (near-zero or slightly negative) thermal expansion coefficient and exceptional thermal stability. β-eucryptite (LiAlSiO4) is a prominent member of this class with applications ranging from heat exchangers to ring laser gyroscopes and telescope mirror blanks. Recently, solid solutions based on β-eucryptite have been recognized to be promising materials for blankets and fuel coatings in breeder reactors. It is, therefore, essential to gain a fundamental understanding of the damage induced by incident radiation in β-eucryptite. It is well known that LAS ceramics become amorphous under irradiation; however, the nature and extent of the structural modification depends both on the type of material and the type of incident radiation. In the present work, we have employed classical molecular dynamics to study collision cascades in β-eucryptite by imparting initial kinetic energy (up to 10 keV) to a primary knock-on atom (PKA). A reactive force field (ReaxFF) for Li-Al-Si-O systems, developed based on density functional theory calculations, was used to describe interatomic interactions. We determined the threshold energies required to cause atomic displacements for each atom species (i.e., Li, Al, Si and O) and compared them with those available from the literature for Al2O3. Higher energy cascades were then investigated to study the nature of radiation-induced defects, their accumulation and the mechanisms underlying structural modifications under irradiation. The results have been discussed in the context of developing radiation tolerant materials based on β-eucryptite
9:00 PM - A8.5
Extended X-Ray Absorption Fine Structure Studies of Radiation Damage-Tolerant Nanocomposites.
Simerjeet Gill 1 , Avishai Ofan 1 , Lynne Ecker 1 , Amit Misra 2
1 Nuclear Sciences and Technolgy, Brookhaven National Lab, Upton, New York, United States, 2 Center for Integrated Nanotechnologies , Los Alamos National Lab, Los Alamos, New Mexico, United States
Show AbstractNanocomposites such as Oxide Dispersion Strengthened (ODS) steels have metal-oxide nanoparticles dispersed in a ferritic matrix and utilize the metal-oxide interface as a sink for radiation-induced vacancies and interstitials. They are promising fuel cladding materials for future reactors. One important challenge in ODS steels is that the atomic environment of the metal-oxide interface is difficult to access due to the geometric complexity of the internal interfaces. This geometric complexity of ODS steels lead to the development of model material systems to explore the physics of defect interactions at interfaces. The model systems are nanocomposite layered materials with large, regularly spaced interfaces between materials. In order to understand the fundamental mechanisms of radiation-induced defect evolution and annihilation at interfaces, model nanocomposite systems with multilayer geometry are being investigated. Nanocomposite layered systems with both metal-oxide/metal (Y2O3/Fe, TiO2/Fe) and metal-metal (Cu/Nb) interfaces are reported. Nanocomposite layered systems have demonstrated high strength and good thermal stability, but local molecular structure and lattice distortions induced at the interface are not clearly understood. In the present studies, Extended X-Ray Absorption Fine Structure (EXAFS) is used to investigate changes in molecular structure and lattice distortions induced at interface in nanocomposite layered systems. In addition the difference in elemental distribution at the interface in nanocomposite layered systems is reported.
9:00 PM - A8.6
Effects of Carbide and Strain Induced Martensitic Transformation on the Cavitation Erosion Resistance of Fe-Cr-C-X(X=Mo, Si) Alloys.
Min Ho Shin 1 , Myung Chul Park 1 , Jae Yong Yun 1 , Seon Jin Kim 1
1 , Hanyang University, Seoul Korea (the Republic of)
Show AbstractThe influences of carbide and strain induced martensitic transformation on the cavitation erosion resistance of Fe-10Cr-1C-X(X=0, 1Si, 1Mo, 1Mo-1Si) alloys has been studied. The amount of carbide was almost same in each alloy. EPMA showed that Mo was mainly contained in the carbide and Si was constant all over the phases. During cavitation erosion test, the alloys exhibited a strain induced martensitic transformation. Among all of the alloys, the best result of the cavitation erosion has been found from 1Mo-1Si alloy with the lowest critical strain energy (CSE) and the results complied with CSE up to 20h in all the alloys. However, this relation could not be established after 20h due to more acceleration of mass loss of 1Mo-1Si alloy after the longest incubation time. It was found that the mass loss rate became almost same between 1Mo and 1Mo-1Si alloy. The fractography near the carbide was also identical. Thus, it was considered that a strain induced martensitic transformation could be more beneficial at the early stage of cavitation erosion while the mass loss rate originated from the erosion mechanism related with carbide in the acceleration period.
9:00 PM - A8.7
Monte Carlo Simulation of Phonon Transport in UO2.
Ryan Deskins 1 , Anter El-Azab 1
1 Department of Scientific Computing, Florida State University, Tallahassee, Florida, United States
Show AbstractMotivated by the need to understand thermal transport in irradiated UO2 at the mesoscale, we present a preliminary Monte Carlo simulation of phonon transport in this important material. The simulation scheme aims to solve the Boltzmann transport equation for phonons within the relaxation time approximation of that equation. In this approximation the Boltzmann transport equation is simplified by assigning time scales to each scattering mechanism associated with phonon interactions. Unlike most previous works on solving this equation by Monte Carlo method, the momentum and energy conservation laws for phonon-phonon interactions are treated exactly; in doing so, the direction and magnitude of possible wave vectors and frequency space are all discretized and a numerical routine is then implemented to consider all possible phonon-phonon interactions and choose those interactions which obey the conservation laws. The simulation scheme accounts for the acoustic and optical branches of the dispersion relationships of UO2. Using periodic boundary conditions, we present results illustrating the ballistic and diffusion limits of phonon transport in UO2 single crystals, and compute the thermal conductivity of the material in the diffusion limit based on the detailed phonon dynamics. The temperature effect on conductivity is predicted and the results are compared with experimental data and Molecular Dynamics simulation results available in the literature. This research was supported as a part of the Energy Frontier Research Center on Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under subcontract # 00091538 from INL to Florida State University.
9:00 PM - A8.8
Effects on PWR Environments Change of Fatigue Crack Growth Rate of Austenitic Stainless Steel.
Ki-Deuk Min 1 , Dae-Whan Kim 2 , Bong-Sang Lee 2 , Seon-Jin Kim 1
1 Materials Science and Engineering, Hanyang University, Seoul Korea (the Republic of), 2 , Korea Atomic Energy Research Institute, Daejeon Korea (the Republic of)
Show AbstractThe austenitic stainless steels have been used as the structural material of Nuclear Power Plants due to the excellent corrosion resistance and the mechanical property. Nb stabilized type 347 stainless steel is used for coolant pressurizer surge line of Korea Standard Nuclear Power Plant (KSNPP). Surge line of PWR nuclear reactor are damaged by thermal fatigue due to thermal gradient during heat-up and cool-down, mechanical fatigue due to mechanical stress, and corrosion fatigue due to nuclear reactor water environment. Fatigue is an important factor which limits the life of structure. PWR nuclear power plant environment, high temperature and high pressure the direct result of the experiment is difficult to obtain. ASME Section ΧΙ fatigue crack growth rate curves of low alloy steel in the air and water are presented in. However, austenitic stainless steel curves are presented only in the air. Fatigue crack growth rate curves in nuclear reactor environment are needed to evaluate the integrity of nuclear reactor structure but that result is not sufficient. In this study, Type 347 austenitic stainless steels in PWR nuclear power plant environment, was to obtain fatigue crack growth curve. And according to the change of dissolved oxygen, frequency and pH to evaluate the fatigue crack growth rate were analyzed. The material used in an experiment is Type 347 commercial stainless steel. The specimen geometry is the thickness 5mm and width 25.4mm CT type specimen. The test environment 150atm, 316°C, dissolved hydrogen(DH) 30cc/Kg, dissolved oxygen(DO) 0.1ppm~5ppb, Boron 1000ppm, Lithium 2ppm and load ratio was 0.1, frequency was 1Hz, 10Hz. Fatigue crack growth rate tests were conducted under load control. The crack length was measured by Direct Current Potential Drop(DCPD) method. Dissolved oxygen increases to 0.1ppm, the fatigue crack growth rate was lower than at room temperature. However, rapid fatigue crack growth rate showed in dissolved oxygen 5ppb. Frequency from 10Hz to 1Hz the reduction seemed to increase the fatigue crack growth rate. And Born, Lithium addition of fatigue crack propagation characteristics were changed.
9:00 PM - A8.9
Effect of Grain Boundary Character Distribution in Austenitic Stainless Steel under Ion Irradiation.
Christopher Barr 1 , Greg Vetterick 1 , Joseph Hsieh 1 , Khalid Hattar 2 , Mitra Taheri 1
1 Material Science and Engineering, Drexel University, Philadelphia, Pennsylvania, United States, 2 Radiation Solid Interaction, Sandia National Laboratory, Albuquerque, New Mexico, United States
Show AbstractAdvanced energy systems such as nuclear power plants require material advancements that provide stability over extended periods of time to a severe, local environment that include high irradiation dose, high temperature and corrosion. The role and interaction of grain boundaries with the localized extreme environment is critical to understanding how to develop new materials and processing routes for advanced materials. Grain boundary engineering, associated with specific thermomechanical processing routes that yield a high fraction of twin and twin-variant grain boundaries, could provide insight into the role of specific grain boundary structure under irradiation environments. Grain boundary character distributions in Fe- and Ni-based alloys that have consisted of a high fraction of these lower than average energy boundaries have previously been associated with improved resistance to grain boundary failure mechanisms including hydrogen embrittlement and intergranular stress corrosion cracking.We employ an experimental method to examine specific grain boundary character and grain boundary character distributions under extreme conditions of hydrogen embrittlement and ion-irradiation. First, a strain-annealing grain boundary engineering method was used to increase the fraction of Σ3n (n=1, 2, 3) coincidence site lattice (CSL) grain boundaries in a 316L austenitic stainless steel. A suite of characterization methods was used to qualitatively understand the role of grain boundary engineering on the local microstructure including triple junction distributions and local grain boundary plane information from the five parameter stereology grain boundary distribution. Following exposure to simulated extreme environments from nuclear energy systems, we employ a site-specific sample preparation technique to investigate the interaction of irradiation induced defects with twin-related boundaries by transmission electron microscopy (TEM). We will provide information on the interaction and dependence of grain boundary character distribution on extreme environments.
Symposium Organizers
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Marjorie Bertolus CEA, DEN, DEC/SESC/LLCC
Blas Uberuaga Los Alamos National Laboratory
Robin W. Grimes Imperial College London
A9: Metallic Systems - Structural Materials I
Session Chairs
Lyndon Edwards
William Weber
Thursday AM, December 01, 2011
Independence W (Sheraton)
9:30 AM - A9.1
Atomic-Scale Modeling of the Migration of Symmetrical Tilt Grain Boundaries in Alpha-Iron.
Jinbo Yang 1 , Yuri Osetsky 1 , Roger Stoller 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractGrain boundary migration plays an important role in plastic deformation of polycrystalline materials in applications. In this work a series of symmetrical tilt grain boundaries in alpha-iron has been constructed on the basis of the Frank-Bilby equation and then relaxed in atomic-scale modeling. In the fully relaxed boundaries the line direction, Burgers vector and average spacing between dislocations are consistent with Frank’s formula. These boundaries are revealed to be capable of migrating through a dislocation mechanism. It is found that the orientation of the crystal on one side of boundary can be converted to that of the adjacent crystal through the boundary migration, and additionally the total lattice sites are conserved and the shape deformation is a simple shear parallel to the boundary plane, in agreement with previous experimental observations and theoretical predictions. Some other controversial topics are discussed according to the results obtained here, e.g. the variation of boundary mobility with misorientation angle and the stress fields near the boundary.
9:45 AM - A9.2
Microstructure Evolution of Heterogeneous Systems under Vacancy Supersaturation.
Enrique Martinez 1 , Jeffery Hetherly 1 , Alfredo Caro 1 , Michael Nastasi 1
1 MST-8, LANL, Los Alamos, New Mexico, United States
Show AbstractA new hybrid Molecular Dynamics-kinetic Monte Carlo (MD-kMC) algorithm has been developed in order to study the microstructure evolution of heterogeneous systems in a vacancy-supersaturated environment. The algorithm takes into account both chemical and stress fields. Migration barriers are calculated using a linear approximation in which final and initial energies are obtained from MD. The approximation makes the algorithm fast enough to be able to handle hundreds of vacancies. Therefore, we are able to observe void formation in twist boundaries in Fe and Cu, characterized by different sets of screw dislocations. We have captured as well the jog formation in an edge dislocation dipole in Fe. Incoherent interfaces have been also studied. The de-mixing process after a displacement cascade event for the case of the Cu-Nb system in the presence of a Kurdjumov-Sachs interface is reported.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, and by the LANL LDRD Office.
10:00 AM - A9.3
Effect of He on the Mechanical Properties of Cu-Nb Interfaces.
Abishek Kashinath 1 , Michael Demkowicz 1
1 Department of Materials Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States
Show AbstractWe use atomistic modeling to show that He concentrations of several atomic percent may be stored at Cu-Nb interfaces without forming bubbles. The He aggregates into platelet-shaped clusters that preferentially form at interface misfit dislocation intersections. The influence of these clusters on the shear resistance and cohesive strength of Cu-Nb interfaces is investigated and the implications to He-induced embrittlement discussed. This work was supported by the LANL LDRD program.
10:15 AM - A9.4
Towards an Equation of State of He Bubbles in a Metal Matrix.
Alfredo Caro 1 , Jeffery Hetherly 1 , Enrique Martinez 1 , Mike Nastasi 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThe evolution of Helium bubbles in metals under irradiation is usually described by one of two major approaches, namely rate equations or phase field. In both cases, quantitative parameters describing the energy change related to an emission or absorption of a He atom by a bubble are needed.A number of assumptions about this process can be found in the literature, among them, assuming He as an ideal gas, or using the pure bulk He EOS, neglecting the metal-He interaction energy.In this talk we present a systematic study of the energetics of He bubbles in bulk Fe and at a particular grain boundary, based on computer simulations of bubbles in different environment situations. We describe the effects of the He-metal interaction in altering the energetics, volume and pressure of He atoms depending on the location inside the bubble, creating a core –mantle structure. Additionally we report on the mechanisms of bubble growth by matrix yield.From the ensemble of data we express the pressure and energy of He in a bubble as a function of He density, radius of the bubble, temperature, and radial position inside the bubble. These equations of state provide a readily accessible and accurate value to feed coarser scale models with information that captures the He-metal interaction in detail, together with the effects of interfaces. 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, and by the LANL LDRD Office.
10:30 AM - **A9.5
3-Dimensional, High-Resolution Modeling of Nuclear Fuel Performance: Pellet Clad Interaction.
Brian Wirth 1 , Derek Gaston 2 , Jason Hales 2 , Richard Martineau 2 , Robert Montgomery 4 , Joseph Rashid 3 , Chris Stanek 5
1 Nuclear Engineering, University of Tennessee, Knoxville, Tennessee, United States, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 4 , Pacific Northwest National Laboratory, Richland, Washington, United States, 3 , ANATECH, San Diego, California, United States, 5 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThe environment experienced by fuel rods inside a nuclear fission reactor is among the most extreme encountered by any functioning materials system. Fission processes in the uranium dioxide ceramic fuel pellets generate high temperatures (and corresponding large thermal gradients due to the relatively low thermal conductivity of the ceramic) and produce high fluxes of energetic neutrons. Outside of the fuel, the zirconium alloy cladding is exposed to a highly corrosive environment, in addition to numerous mechanical and chemical interaction forces as it serves as the first engineered barrier against the release of radioactive fission products. This presentation will first introduce the inherently multiscale nature of irradiation effects in nuclear fuels and cladding materials and then describe an engineering-scale three dimensional based framework for modeling nuclear fuel performance, with an emphasis on pellet clad interaction. This effort is being performed within a new Energy Innovation Hub for nuclear energy that is funded by the Department of Energy, the Consortium for the Advanced Simulation of Light Water Reactors (CASL). The materials and fuels performance research into pellet clad interaction-based fuel failures is focused on developing the Peregrine thermal-mechanical model for nuclear fuel performance. Peregrine will be described in detail and contrasted to the current state of the art in the numerical simulation of fuel performance. Following a description of the important materials degradation phenomena and the current best effort models to describe them, the presentation will focus on an example demonstrating the capability to the model stress state in a Zirconium alloy fuel clad surrounding a fuel pellet with a missing pellet surface during an operational transient.
11:30 AM - A9.6
Atomic-Scale Mechanisms of Radiation Induced Strengthening Metals.
Yury Osetskiy 1 , Roger Stoller 1
1 Materials Science and Technology, ORNL, Oak Ridge, Tennessee, United States
Show AbstractAccumulation of radiation-induced microstructure leads to specific behavior for irradiated materials under deformation such as hardening, strengthening, loss of ductility, etc. In addition, significant microstructure modification may occur during deformation. We present here a review of results of extensive atomic-scale modeling devoted to the study of the interaction between moving screw and edge dislocations in bcc and fcc metals and irradiation-specific defects such as voids, He-filled bubbles, stacking fault tetrahedra and interstitial dislocation loops. All reactions between dislocations and individual defects were separated into two groups that is dislocation-type reactions (with dislocation loops and SFT) and inclusion-type reactions (voids, precipitates, bubbles etc.). Mechanisms of strengthening and microstructure changes due to these reactions involve a spectrum of effects, including complete elimination or restoration of defects, their mutual recombination, and change of size and/or structure (shear, Burgers vector change, etc.). We show how the reactions may be classified and discuss them in the light of experimental observations and continuum theory approach.
11:45 AM - A9.7
Strain Rate Effects on the Mechanism of Dislocation Interactions with Self-Interstitial Atom Clusters in Zirconium.
Yue Fan 1 , Sidney Yip 1 2 , Bilge Yildiz 1
1 Nuclear Science & Engineering, MIT, Cambridge, Massachusetts, United States, 2 Material Science & Engineering, MIT, Cambridge, Massachusetts, United States
Show AbstractIrradiation creep is an important long-term macroscopic degradation phenomenon in nuclear cladding materials that involves dislocation interactions with microstructural obstacles. The study of these interactions by conventional molecular dynamics (MD) simulations has been hampered by the difficulty of reaching realistic strain rates. We employ a new, alternative approach, the Autonomous Basin Climbing (ABC) method (Kushima et al., JCP, 130, (2009)), to construct the atomic trajectories and the corresponding potential energy landscape associated with the microstructural evolution. We investigate the climb of a prismatic edge dislocation in interactions with irradiation induced self-interstitial atom (SIA) clusters in Zr, obtaining results spanning a much wider range of strain rates than is possible with traditional MD simulations. For static stress-strain behavior we find of the prismatic edge dislocation in overcoming a 5-SIA cluster on its glide plane. At low stresses dislocation pinning by a 5-SIA cluster on its glide plane leads to stress relaxation. For shear stress beyond the critical value of 250 MPa, the dislocation passes through the cluster and both the dislocation and the cluster fully recover their original structure, consistent with prior findings (Voskoboynikov et al Mater. Sci. Eng. A, 400-401, (2005)). Results for strain rates from 103s-1 to 105s-1 and several temperatures are presented to show appreciable effects on the interaction paths, energy barriers and critical stresses. A critical strain rate is identified below which the SIA cluster is eventually absorbed by the dislocation forming a jog structure; at higher rates the interaction follows the same full recovery mechanism as in the static behavior. Temperature and strain rate dependence of the transitions between these two interaction mechanisms will be described. Present capability to extend atomistic simulations to low strain rate is expected to lead to new insights into dislocation mechanisms in irradiation creep.
12:00 PM - A9.8
Radiation Growth of HCP Metals under Cascade Damage Conditions.
Stanislav Golubov 1 , Alexander Barashev 1 , Roger Stoller 1
1 , ORNL, Oak Ridge, Tennessee, United States
Show AbstractModels of radiation growth have developed up to date are all based on the assumption that the primary damage produced via neutron irradiation takes place in the form of single point defects. These models do not account for the most important feature of cascade damage: intra-cascade clustering of self interstitial atoms (SIAs) and their one dimensional diffusion. During the last twenty years, a ‘Production Bias Model’ (PBM) has been developed, which shows that the damage accumulation in BCC and FCC metals crucially depends on cascade properties. Since the cascade properties in hcp, e.g. zirconium, are found to be similar to those in cubic crystals one may expect that the PBM can provide a realistic framework for the hcp metals as well. An objective for the work is to present such a model in application to low temperature (below 600K) radiation growth of single Zr crystals.
12:15 PM - A9.9
Electrostatic and Elastic Effects in the Theory of Void Growth under Irradiation.
Thomas Hochrainer 1 , Srujan Rokkam 2 , Anter El-Azab 1
1 Department of Scientific Computing, Florida State University, Tallahassee, Florida, United States, 2 Mechanical Engineering Department, Florida State University, Tallahassee, Florida, United States
Show AbstractNucleation and growth of voids is a key technical problem in irradiated materials. Classical studies of void growth under irradiation treat the growth process within the framework of a homogenized (effective medium theory) chemical rate theory. Over the past few years, a new spatially resolved approach for modeling void nucleation and growth emerged, which is based on the phase field framework. Development of quantitative phase field models requires connecting specific parameters of the diffuse interface model to the physical parameters of a more explicit ‘sharp interface’ formulation of void surface motion. We present a sharp interface model for void surface motion driven by fluxes of electro-statically and elastically interacting point defects. The model is based on transition state theory and features the void surface as a material discontinuity which can act both as defect sink and source. After validating the model at simple example problems we present two-dimensional results on coupled defect and surface evolution under irradiation in various geometries. Special emphasis is put on discriminating the influence of electrostatic and elastic interactions and discussing the consequences of void growth modeling in oxide fuels. Furthermore, we discuss the implications of the model for the development of diffuse interface phase field models. This research was supported as a part of the Energy Frontier Research Center on Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under subcontract # 00091538 from INL to Florida State University.
12:30 PM - A9.10
Simulation of H-He-Vacancy Defect Clusters in W.
Niklas Juslin 1 , Brian Wirth 1
1 Nuclear Engineering, University of Tennessee, Knoxville, Tennessee, United States
Show AbstractHydrogen and helium will be present in fusion reactor materials, due tobombardment from the plasma, as well as transmutation reactions causedby high energy neutron bombardment. Of particular interest is the effectof the high heat load of low-energy (1-100 eV) hydrogen isotope and heliumbombardment of the divertor, for which tungsten is a candidate material.Hydrogen and helium are known to modify bulk and surface morphology andmaterial properties of metals, e.g. bubble and blister formation, swellingand change in ductile to brittle transformation temperature. Continued lowsputtering, good thermal properties and structural strength and during intenseirradiation are key issues for a divertor material, and further understandingof the effect of hydrogen and helium is needed.Molecular dynamics (MD) and Monte Carlo (MC) simulations are important toolsto study hydrogen and helium in tungsten on an atomistic level, which is noteasily achieved in experiments, and on longer length and time scales than thoseavailable in ab initio calculations. A multiscale approach is used by basingsemi-empirical inter-atomic potentials on experimental and density-functionaltheory (DFT) results, using MD to study mechanisms and rates and MC forlonger time scales. In order to identify the correct mechanisms andtrustworthy qualitative and quantitative results, the inter-atomic potentialsused in the simulations are of utmost importance.We discuss the currently available potentials for the W-H-He system, as wellas ongoing and needed development work for improved description of theinteractions. We present MD results for the formation of hydrogen-helium-vacancydefect clusters in tungsten, and the binding of defects to the clusters. Forsmall clusters a comparison with literature density functional theory data ismade. In addition the difference between clusters in bulk and near the surfaceis studied. The binding and clustering of defects are of importance for thebuild-up of larger bubbles and changes in bulk and surface morphology.
12:45 PM - A9.11
Stress Effects on Void Growth under Irradiation.
Srujan Rokkam 1 , Karim Ahmed 2 , Anter El-Azab 2 3
1 Mechanical Engineering, Florida State University, Tallahassee, Florida, United States, 2 Materials Science and Engineering Program, Florida State University, Tallahassee, Florida, United States, 3 Department of Scientific Computing, Florida State University, Tallahassee, Florida, United States
Show AbstractWe investigate the effect of stress on the nucleation and growth of voids in irradiated single component materials with a phase-field approach. The formulation of the problem couples point defect dynamics and void microstructure evolution with the elastic effects arising due to the non-uniform lattice relaxation of defects and the inhomogeneity introduced by voids. The phase-field problem requires the solution of two Cahn-Hilliard equations for evolution of vacancy and interstitial concentration fields and an Allen-Cahn equation for void nucleation and growth. In addition, the formalism requires the solution of a subsidiary stress equilibrium problem, which is cast in the form of an eigenstrain problem. Analysis of the coupled phase-field/elastic problem shows that the model recovers exactly the corresponding sharp interface formalism of point defect diffusion and void growth under applied stress, with zero traction boundary condition over the interior void surfaces. In the case of gas bubbles, a traction boundary condition in terms of the gas pressure is recovered. We present the overall formalism of the problem and discuss results showing the effect of applied stress on the defect migration, void growth and void-void interactions. Effect of stress on the anisotropy of diffusion of point defects and on the formation of self-organized void ensembles will also be illustrated. This research was supported as a part of the Energy Frontier Research Center on Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under subcontract # 00091538 from INL to Florida State University.
A10: Metallic Systems - Structural Materials II
Session Chairs
Thursday PM, December 01, 2011
Independence W (Sheraton)
2:30 PM - A10.1
In Situ and Ex Situ TEM Ion Irradiation Experiments for Cladding Materials.
Khalid Hattar 1 , Brad Yates 1 2 , Thomas Buchheit 1 , Blythe Clark 1 , Luke Brewer 1 3
1 , Sandia National Labs, Albuquerque, New Mexico, United States, 2 , University of Florida, Gainesville, Florida, United States, 3 , Naval Postgraduate Institute, Monterey, California, United States
Show AbstractThe alloys used for the cladding of both current nuclear reactors and those proposed for potential future reactors undergo a progression of microstructural changes during reactor operation that can significantly impact the material integrity. As many of the processes are driven or influenced by the irradiation environment, it is important to understand the changes in microstructure, such as formation of defects, precipitate dissolution, and chemical segregation, as a function of radiation dose. This work will highlight two programs at Sandia to use ion irradiation to simulate neutron damage and to directly observe this damage using a newly developed in-situ ion irradiation transmission electron microscope. The first program uses ion beams to simulate damage in a 316L and HT9-based diffusion couples with a variety of refractory elements to predict lifetimes of advanced Generation IV reactor claddings. The second program aims to increase understanding of microstructural changes as a function of burnup in Zr-based alloys currently used in boiling water and pressurized water reactors. In addition, the current status of a feasibility study to combine ion irradiation and vapor/liquid cell experiments within the transmission electron microscope for studying ion-assisted corrosion mechanisms will also be discussed.This work is supported by the Division of Materials Science and Engineering, Office of Basic Energy Sciences, U.S. Department of Energy both at Sandia and under grant DE-FG02-07ER46443. Sandia National Laboratories is a multi-program laboratory operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin company, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.
2:45 PM - A10.2
Cluster Dynamics Modeling of Point Defect Cluster Evolution in Ferritic/Martensitic Iron Chrome Alloys.
Aaron Kohnert 1 , Brian Wirth 1 2 , Nathan Capps 2 , Djamel Kaoumi 3 , Arthur Motta 4 , Cem Topbasi 4
1 Nuclear Engineering, University of California, Berkeley, California, United States, 2 Nuclear Engineering, University of Tennessee, Knoxville, Tennessee, United States, 3 Mechanical Engineering, University of South Carolina, Columbia, South Carolina, United States, 4 Mechanical and Nuclear Engineering, Pennsylvania State Univeristy, University Park, Pennsylvania, United States
Show AbstractCluster dynamics modeling is used to investigate the mechanisms controlling microstructural evolution in iron chrome alloys following high dose radiation exposure over a wide range of temperatures. The primary goal of this modeling is to investigate the development of defect clusters during exposure to heavy ion irradiation as observed at the IVEM facility at Argonne National Lab. The cluster dynamics method used in this study is an expansion of simpler rate theory approaches with added consideration of the possibility of defect reactions to form arbitrarily large clusters and to explicitly include spatially dependent defect sinks, including the free surfaces of the thin foil in-situ irradiation samples. Primary damage production is implemented with a multiscale approach that uses results from a database of molecular dynamics simulations of displacement cascades to implant damage in the form of defect clusters in addition to Frenkel pairs. Particular attention is given to determining the appropriate diffusivity for clusters of various sizes in light of the differences between experimental and computational studies of interstitial cluster mobility. The role of cascade-defect interactions in determining the ultimate mobility of defects is considered as well. The results demonstrate the important effects of defect trapping as well as the interaction between displacement cascades and trapped defect clusters to explain the lack of a strong temperature dependence in the experimental measurements of defect cluster density and size.
3:00 PM - **A10.3
Application of Advanced Experimental Techniques to the Study of Materials for Nuclear Power Plant Applications.
Mike Fitzpatrick 1
1 Materials Engineering, The Open University, Milton Keynes United Kingdom
Show AbstractThe next generation of nuclear power plant will be designed for longer operating lives with increasingly challenging operating conditions for the materials in the reactor core and heat exchangers. Understanding the mechanisms that underpin the behaviour of materials is critical in the development of accurate models for life prediction.The development and application of experimental methods such as neutron diffraction to the study of materials for nuclear power plant has provided significant advances in the characterization of the fundamental behaviour of materials for applications in nuclear power plant. Specific examples are the micromechanics of deformation of materials for fuel cladding, and the development of internal strains in austenitic steels as a consequence of load and temperature cycling. The ability to monitor the strain evolution in individual grain families during simulated processing or operating conditions is a powerful tool for understanding subsequent behaviour. Zirconium alloys, for example, are processed to ensure a beneficial crystallographic texture, and the evolution of texture and intergranular strain can be monitored directly by in situ loading in a neutron beam. Structural steels are required to have long-term creep resistance, but the effects of thermal and load cycling during plant operation on creep mechanisms are relatively poorly understood, and again neutron diffraction can provide valuable insight. There has also been a step change in our ability to determine residual stresses in complex welded geometries, which is highly-beneficial given the problems arising from effects such as reheat cracking of welds. A key development has been the integration of advanced experimental simulation tools with neutron instrumentation, that has improved our capability of analysing residual stresses in components with complex geometries, with high spatial accuracy. Such measurements are essential in the development of accurate models for prediction of structural integrity of power plant.
3:30 PM - **A10.4
The Role of Advanced Weld Modelling in Ensuring the Structural Integrity of Present and Future Welded Plant in the Nuclear Industry.
Lyndon Edwards 1 , Ondrej Muransky 1 , Cory Hamelin 1 , Phil Bendeich 1
1 Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Sydney, New South Wales, Australia
Show AbstractThe continued safe operation of present nuclear power plant is dependent upon structural integrity assessment of pressure vessels and piping. Furthermore, structural failures most commonly occur at welds so the accurate design and remnant life assessment of welded plant is critical. The residual stress distribution assumed in defect assessments often has a deciding influence on the analysis outcome, and in the absence of accurate and reliable knowledge of the weld residual stresses, the design codes and procedures use assumptions that yield very conservative assessments that can severely limit the economic life of some plant. However, recent advances in both the modelling and measurement of residual stresses in welded structures and components open up the possibility of characterising residual stresses in operating plant using state-of–the–art fully validated Finite Element simulations. Indeed such approaches are now beginning to be codified in failure assessment methodologies such as that used in the UK’s R6 methodology. This paper describes recent international multidisciplinary research undertaken to predict residual stresses in welded structures in order to provide validated reliable, accurate Structural Integrity assessments that can e used to assess the life of present nuclear power plant components and systems.History has shown us that is not unknown for new failure modes to be discovered only when a component or structure has entered service in the nuclear industry, because full-scale testing under reactor operation conditions for design lifetimes is simply not feasible. To combat this, the industry is very conservative with present design codes (ASME 2001; ref. 2), currently only including a relatively small number of materials, which are well characterized under present reactor conditions that were developed during the last half of the twentieth century. Future reactors whether Gen IV or Fusion pose greater challenges because of the even more extreme conditions in which they will operate. These will only be met using the present radiation tolerant materials currently under development. However, in many cases we still to develop manufacturing methods and joining methods to construct the complex geometries that make up both the reactor systems and their associated heat exchangers and generators. The paper will outline what needs to be accomplished before we can use today’s novel radiation resistant materials in future state of the art, long life, high efficiency nuclear power plant.
A11: Carbides
Session Chairs
Massey de los Reyes
Samuel Murphy
Thursday PM, December 01, 2011
Independence W (Sheraton)
4:30 PM - A11.1
Silicon Carbide Phase Stability in a Fusion Environment.
Magdalena Serrano de Caro 1 , Alfredo Caro 1 , Christopher Stanek 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractProgress in developing fusion reactor concepts is constrained by materials technology. The materials selected for the first wall should be able to stand high thermal loading and high fast neutron fluence, in the presence of corrosive cooling environments. Materials near the area of the first wall will be exposed to high fast neutron fluence (~10^23 n/cm2), accumulate a large dose (~ 100 dpa) and be subjected to large gas and solid transmutation effects.Silicon carbide ceramics and SiC/SiC composites are considered promising first wall materials in various fusion design concepts because of their excellent high-temperature properties, good corrosion resistance and low activation. However, the effect of solid transmutants on SiC swelling and irradiation stability is a major concern. A recent review paper [1] shows that in either Inertial (IFE) or Magnetic (MFE) fusion reactor concepts, the percent burnup (atoms of Si and C in SiC) is high; up to 10% loss of atoms after reaching the end of the first wall expected 5-year lifetime. Transmutation reactions in SiC generate H, He and up to 1% Mg after an exposure of 5 years from fast neutron (n,α) reactions in Si-28 and Si-29 leading to the formation of stable isotopes Mg-25 and Mg-26, respectively.Following a path similar to that described in [2], we study the transmutation defect reactions in which the daughter product resides in the Si site, examine the evolving chemistry and investigate the effect of lattice stability of a large concentration of an insoluble solute looking for the formation of new/rare phases, in particular MgC, a crystalline structure with limited information available in the literature. In this work, we apply density functional theory (DFT) combined with the pseudo-potential approximation and a basis set of numerical atomic orbitals as implemented in the program SIESTA [3] to explore the presence of possible new phases, e.g. MgC, Mg3C2 and MgC2. The USPEX code [4] is used in combination with SIESTA to predict crystal structures for MgC at several temperature conditions. We show that a phenomenon similar to that of radioparagenesis could lead to the formation of rare crystal structures in Si when exposed to the extreme solid transmutation conditions occurring at the first wall of the fusion power reactor. This material is based upon work supported by the LANL LDRD Office.[1] L.L. Snead et al., Journal of Nuclear Materials xxx (2011) xxx–xxx, article in press.[2] B.P. Uberuaga et al., Nucl. Inst. and Meth. in Phys. Res. B 268 (2010) 3261.[3] J. Soler et al., J. Phys.: Condens. Matter 14 (2002) 2745.[4] C. W. Glass et al., Computer Phys. Comm. 175 (2006) 713.
4:45 PM - **A11.2
Modeling of Radiation Damage in Nanocrystalline Silicon Carbide.
Izabela Szlufarska 1 , Dane Morgan 1 , Narasimhan Swaminathan 2 , Ming-Jie Zheng 2
1 Materials Science and Engineering and Materials Science Program, University of Wisconsin-Madison, Madison, Wisconsin, United States, 2 Materials Science and Engineering, University of Wisconsin-Madison, Madison, Wisconsin, United States
Show AbstractThe excellent mechanical properties of ceramics, such as silicon carbide, make them promising structural materials for next generation nuclear fission and fusion reactors. Further enhancements of mechanical properties of these materials are possible through reduction of grain size to the nanometer regime. However, understanding of the effect of grain refinement on radiation resistance is currently limited. In this talk we will discuss the current state of knowledge on radiation effects in nanocrystalline silicon carbide, including the results of our multi-scale simulations. Specifically, using molecular dynamic simulations we have shown that primary defect production in SiC is not affected by the grain refinement, but that grain boundary structures are affected by the radiation. This latter effect is particularly pronounced for low-energy grain boundaries. Building on a large number of recent ab initio studies we constructed a rate-theory model for SiC’s long-term point defect evolution and predict how its amorphization couples to temperature and grain size. Using this ab initio based continuum model we have also demonstrated that in addition to migration barriers, recombination barriers for defects can play an important role in the resistance of SiC to radiation-induced amorphization. Although these barriers are critical to determining whether grain refinement reduces or increases radiation resistance, contradictory reports regarding these barriers have been published in the literature. Our ab initio molecular dynamics simulations have resolved these discrepancies by providing a detailed energy landscape for interstitial-vacancy pairs migrating toward each other from large distances and recombining. Comparison between the predicted trends in radiation-induced amorphization as a function of grain size in SiC and the preliminary experimental data obtained by collaborators at the University of Wisconsin will also be discussed.
5:15 PM - **A11.3
Effects of Displacing Radiation on Mrozowski Cracks in Graphite Observed Using in situ Transmission Electron Microscopy.
Jonathan Hinks 1 , Abbie Jones 2
1 Electron Microscopy and Materials Analysis, University of Huddersfield, Huddersfield United Kingdom, 2 School of Mechanical, Aerospace and Civil Engineering, University of Manchester, Greater Manchester United Kingdom
Show AbstractGraphite is important to the United Kingdom’s fleet of advanced gas-cooled reactors (AGRs) as both a moderator and structural component. Under the temperatures and displacing radiation fluxes present in the cores of AGRs, graphite demonstrates significant changes to its dimensions and physical properties.In order to continue to operate AGRs safely and to extend their lifetimes, it has become necessary to develop a better understanding of this problem in order to predict the changes in the graphite from which they are constructed. The nuclear graphites used are highly inhomogeneous materials. One of the features in these materials are Mrozowski cracks which have been observed to grow under displacing radiation and to shrink under annealing using in situ transmission electron microscopy at the Microscope and Ion Accelerator for Materials Investigations (MIAMI) which has recently moved to the University of Huddersfield.
5:45 PM - A11.4
Silver Behavior on and near the SiC Surface.
Haiyan Xiao 1 , Yanwen Zhang 2 1 , William Weber 1 2
1 Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee, United States, 2 Materials Science & Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractThe high-temperature gas-cooled nuclear reactor (HTGR) is one of several advanced reactor concepts being pursued world-wide. Most HTGR concepts are based on tristructural isotropic (TRISO) coated particles as the nuclear fuel, in which SiC is the most important coating for structural integrity and fission product containment. However, the volatile fission-product metals (e.g., Ag, Cs, Sr) have been reported to pass through the dense SiC coating and be released from intact TRISO particles. The release of radionuclides, especially 110mAg, a strong γ-ray emitter with a half-life of 253 days, from the core during normal operation is of particular importance to direct-cycle, gas-turbine HTGR designs because of their impact on safe and economic operation of the nuclear reactors.In order to effectively reduce or control the silver release, it is of crucial importance to develop a fundamental understanding of the possible diffusion and release mechanisms. Despite numerous experimental studies of the transport behavior of silver in SiC, significant discrepancies exist in the diffusion coefficients and the proposed diffusion mechanisms, including volume diffusion (i.e., solid phase transport through bulk SiC), grain boundary diffusion, and vapor or surface diffusion through structural imperfection such as nano-pores or nano-cracks. An integrated experimental and computational study of silver behavior has been carried out using ab initio calculations and ion beam techniques (ion implantation and ion beam analysis). In the calculations, low-index cubic β-SiC planes are considered. The energies for silver adsorption/absorption on and at the SiC surfaces are calculated and the preferred adsorption (or absorption) sites are determined. To gain a more clear insight into silver behavior, silver diffusion on the surface, diffusion from the surface into interior layers, and diffusion away from the surface are studied to determine the energy barriers and diffusion pathway. The simulation results are compared with well-designed/controlled experiments in crystalline SiC. Experimentally, energetic silver ions were implanted into SiC single crystals at temperature above the critical temperature for amorphization to introduce a detectable amount silver into crystalline SiC. Two energies were used: lower energy to produce a silver profile peaked at the surface, and higher energy to create a buried silver profile. Thermal treatment was performed to study possible silver redistribution or release from surface using complimentary ion beam analysis techniques. The synergetic study based on well-integrated ab initio calculations and ion beam methods suggests that surface diffusion through mechanical structural imperfection, such as vapor transport through cracks in SiC coatings, may be a dominating mechanism accounting for silver release from the SiC in the nuclear reactor.
Symposium Organizers
Karl R. Whittle Australian Nuclear Science and Technology Organisation
Marjorie Bertolus CEA, DEN, DEC/SESC/LLCC
Blas Uberuaga Los Alamos National Laboratory
Robin W. Grimes Imperial College London
A12: Metallic Systems
Session Chairs
Friday AM, December 02, 2011
Independence W (Sheraton)
9:45 AM - A12.2
Direct Conversion Neutron Detection Using Functionalized Boron Nitride Nanotubes.
Jacob Eapen 1 , Brahmananda Chakraborty 1
1 Nuclear Engineering, NC State University, Raleigh, North Carolina, United States
Show AbstractSolid state neutron detection with indirect conversion involves two distinct materials – a neutron sensitive material for capture reactions (such as boron) and a semiconducting material for electron–hole pair generation (such as silicon). Advancements in the recent years have focused on optimization of indirect conversion geometries, transport and modern processing techniques for conventional semiconductors such as silicon and germanium. In contrast, neutron capture and charge generation are facilitated by the same material in a direct conversion neutron detector. Traditional choices for the detector material such as B5C, and uranium and gadolinium rich materials are not optimal for neutron capture, as well as generation, separation and collection of charge carriers. The limiting factor thus far originates from the rather poor semiconducting properties of direct conversion materials.Functionalized nanotubes provide an exciting alternative where the electronic properties can be enhanced by several orders in magnitude through appropriate doping while maximizing the neutron capture reactions. One of the nanostructured materials which is relevant to neutron detection is the boron nitride (BN) nanotube which has several interesting properties including high neutron capture cross–section, ion radiation stability, thermal stability, and chemical inertness. Our work explores doped BN nanotubes as a direct conversion detector material that can potentially improve the spatial, temporal and energy resolution of neutron detectors. Using density functional theory (DFT) simulations we have investigated the electronic structure of BN nanotubes doped with group-IV elements and transition metals. Our results show that the band gap can be tuned by varying the doping element and concentration. Our approach thus can potentially lead to the development of a new class of solid–state neutron detectors using nanostructures for improved sensitivity and range.
10:00 AM - A12.3
Plasticity-Induced Oxidation Reactivity on Ni(100) Studied by Scanning Tunneling Spectroscopy.
William Herbert 1 , Bilge Yildiz 2 , Krystyn Van Vliet 1
1 Materials Science and Engineering, MIT, Cambridge, Massachusetts, United States, 2 Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States
Show AbstractEnhanced reactivity at defects during incipient plasticity is likely to influence the early stages of degradation mechanisms such as stress-corrosion cracking. Here we study for the first time the effect of individual dislocations on the surface electronic structure of nickel. Using in-situ indentation with a customized diamond tip we show that highly localized mechanical deformation can be coupled with structural and electronic characterization in the scanning tunneling microscope. Individual dislocations were topographically imaged and systematically probed by scanning tunneling spectroscopy to assess their effect on local surface electronic structure. Compared to undamaged, flat terraces, regions around dislocations exhibited a significant increase in local density of states around, and in particular below, the Fermi level. We interpret these results using a well-established "d-band" model of surface reactivity and suggest that dislocations emerging at the surface serve to locally accelerate adsorption-driven chemical reactions with species such as oxygen.
10:15 AM - A12.4
Modelling Radiation Damage in Ni Based Alloys.
Zainab Al-Tooq 1 , Steven Kenny 1
1 Mathematical Sciences, Loughborough University, Loughborough United Kingdom
Show AbstractNi based materials are of interest as possible materials for generation IV reactors, due to their higher operating temperatures than present reactors. One aspect of the in service behaviour of Ni based materials that is not fully understood is segregation. This study looks at both grain boundary segregation and the formation of Ni2Cr precipitates during the aging of these materials, both due to thermal diffusion and due to the effects of radiation. In particular the depletion Cr at grain boundaries caused by the inverse Kirkendall effect driven by vacancies created by radiation damage is studied. Furthermore, the formation of Ni2Cr precipitates both due to thermal diffusion and radiation enhanced diffusion is studied.The system has been modelled at an atomistic level by a combination of molecular dynamics and on-the-fly kinetic Monte Carlo (otf-KMC). In the otf-KMC work the transition pathways have been discovered using the RAT method, whilst the barriers having been calculated using the climbing image nudged elastic band method. Once the possible diffusion pathways have been calculated then their rates are computed and a move is selected using the standard kinetic Monte Carlo approach. Utilising this approach allows us to explore timescales that are far beyond those accessible via molecular dynamics, but does not require the diffusion pathways to be pre-specified. The local environment around a defect is properly taken into account in this approach, thus giving insight into local composition based effects.
10:30 AM - A12.5
Atomistic Simulations of Radiation Damage in Amorphous Metals.
Richard Baumer 1 , Michael Demkowicz 1 , Tomas Oppelstrup 2 , Vasily Bulatov 2
1 Materials Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States, 2 Condensed Mater and Materials Division, Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractWe synthesize a 474 million atom amorphous Cu50Nb50 system via massively parallel classical molecular dynamics melting and quenching simulations and subsequently study the passage of a 475keV Nb atom through the as-quenched material. By resolving atomic-level stresses, we observe the formation of a cylindrical shockwave, axisymmetric with the ion trajectory, and study the effect of its propagation on the material. We additionally probe for ion-induced local melting and measure the spatial distribution of liquid-like regions. This work yields insight into the nature of the radiation response of a model amorphous solid.
10:45 AM - A12.6
Probing Low-Level Radiation Damage Using Thin Film Capacitors and Dielectric Measurements.
Alexander Smith 1 , Y. Zhang 2 , W. Weber 2 , S. Shannon 3 , Jon-Paul Maria 1
1 Materials Science and Engineering, North Carolina State University, Raleigh, North Carolina, United States, 2 Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee, United States, 3 Nuclear Engineering, North Carolina State University, Raleigh, North Carolina, United States
Show AbstractAs nuclear reactor designs improve they utilize new materials for their improved properties under extreme conditions. Investigation into radiation-solid interactions of advanced reactor materials must provide quantitative measurements of parameters like defect accumulation rate, defect annihilation rate, and defect mobility while also providing identification of metastable phases that are only stable in extreme environments. Here we present initial findings of a study where refractory nuclear reactor materials are prepared in the form of thin film capacitors, a geometry providing rapid and uniform irradiation while allowing for post-dose and in situ monitoring by conventional dielectric measurement to detect damage at defect levels that are transparent to most ex situ techniques. Sparse populations of point defects are the initial manifestations of irradiation, and provided they are charged with respect to the host lattice, will be easily detected electrically at PPB levels and lower.The material of choice for this study is CeO2, which can be prepared using thin film techniques with dense and tunable microstructures featuring large columnar grains or nanoscaled equiaxed grains. Films spanning these characteristics were prepared using magnetron sputtering and were characterized using x-ray diffraction and AFM image analysis. Standard CeO2 specimens were prepared at room temperature and feature columnar 110-oriented grains with average diameters of 100 nm. All films were prepared to a thickness of 500 nm. The CeO2 films were deposited on two substrates: (001) silicon and (0002) pyrolytic graphite. The Si substrates were coated with 10 nm of Al prior to ceria deposition. The graphite was used in the as received state. 10 nm of Al was deposited on the CeO2 surfaces to complete Al-CeO2-Al and C(g)-CeO2-Al metal-insulator-metal capacitor structures. Dielectric characterization of these structures revealed highly insulating ceria with a permittivity of approximately 24 and negligible dispersion in the frequency range between 0.1 kHz and 100 kHz. The films could withstand electric fields of 500 kV/cm with no increase in dielectric loss.Experiments that expose our CeO2 capacitors to MeV Au radiation are underway and we will report the relationship between radiation exposure and dielectric properties. The initial intent of this research is identifying the damage threshold visible to dielectric characterization, and to compare the changes in electrical properties that occur at damage levels that can be seen using conventional techniques, specifically x-ray diffraction.
11:30 AM - A12.7
Monte Carlo Study of Equilibrium Properties in Iron-Chromium Alloys: From Solubility Limits to Free-Energy Surfaces.
Gilles Adjanor 1 , Manuel Athenes 2 , Jocelyn Rodgers 3
1 Groupe Métallurgie, EDF, Moret-sur-Loing France, 2 SRMP, CEA, Gif-sur-Yvette France, 3 Smit Group, LBL, Berkeley, California, United States
Show AbstractModeling the thermodynamic and kinetic properties of chromium ferritic/martensitic steels is important for better understanding the effect of ageing on the behaviour of structural materials to be used in Generation IV reactors. Here, we present a Monte Carlo study of the equilibrium properties of FeCr alloy. Simulations consist of firstly sampling a transmutation ensemble via gradual transformations of an iron atom into a chromium atom and secondly estimating the difference of chemical potentials as a function of Cr composition and temperature [1]. The free energy surface is directly reconstructed from the chemical potential differences which gives access to the solubility and spinodal limits and to the interface free energies. [1] Waste-recycling Monte Carlo with optimal estimates: application to free energy calculations in alloys, arXiv:1105.3874v1.
11:45 AM - A12.8
Crystal Plasticity Finite Element Modeling of Stress Localization in BCC Polycrystals during Creep.
Zhe Leng 1 , David Field 1
1 School of Mechanical and Materials Engineering, Washington state university, Pullman, Washington, United States
Show AbstractFerritic/martensitic steels such as HT9 steel, is used for structural components in nuclear power plants because of its high strength and good swelling resistance. Understanding the creep behavior of these steels is quite important, since it will affect the strength and creep life of the component. In this study, a finite element polycrystal plasticity model is embedded in ABAQUS to carry out the modeling of stress localization in BCC polycrystals. A thermal activation energy theory is used to calculate plastic flow and self and latent hardening and dislocation densities are tracked explicitly in the model. Experimental observations of HT9 in creep at elevated temperature (about 0.5 Tm), along with microstructural observations were used to study creep damage that occurred during creep. The modeling result exhibits good agreement with experimental observation with stress concentrations at specific grain boundaries leading to cracks and voids. The model also shows that the stress concentrations at grain boundaries depend upon the grain boundary character.
12:00 PM - A12.9
Freeze-Cast Stainless Steel Scaffolds for Thermally Conductive, Mechanically Supportive Nuclear Fuel Pins/Pellets.
Jonadan Ando Burger 1 , Clarissa Yablinsky 2 , Thao Le 1 , Amanda Lang 2 , Cale Buxton 1 , Todd Allen 2 , Ulrike Wegst 1
1 Materials Sci.+Eng., Drexel University, Philadelphia, Pennsylvania, United States, 2 Engineering Physics, University of Wisconsin - Madison, Masdison, Wisconsin, United States
Show Abstract†Freeze casting is a versatile process for creating tunable size and structure aligned porosity in polymer, ceramic, metallic, or composite structures. One can tune pore size and structure and shape via changes in processing conditions or slurry additives and hence exhibit numerous degrees of control over scaffold properties. Stainless steel scaffolds are of particular interest for nuclear fuel pins/pellets, particularly to enhance the thermal conductivity and temperature distribution within a fuel unit, hopefully to enhance burn-up throughout the fuel as well as heat transfer to cladding/coolant and hence increase plant efficiency. Moreover, such scaffolds may provide an internal means for fission gas pressure relief/release and strain tolerance to fuel swelling while stabilizing the fuel unit’s overall shape. Pore size/structure characterization and trends based on additives are helpful for ceramic/fuel infiltration as well as mechanical and thermal property correlations and neutronics calculations and modeling. Mechanical testing and thermal conductivity analysis are useful for initial performance assessments for fuel scaffold optimization as well as structure-property-processing correlations which are important for fundamental understanding of thermal, nuclear, and lifetime behaviors, optimal fuel pellet design, and potential scalability of the freeze casting process potentially to a developmental or industrial level.
12:15 PM - A12.10
Grain Boundary Analysis of Crept Alloy 617.
Fan Zhang 1 , David Field 1
1 Mechanical and Material Engineering, Washington State University, Pullman, Washington, United States
Show AbstractAlloy 617, a high-temperature creep-resistant, nickel-based alloy, is being considered for the primary heat exchanger for the Next Generation Nuclear Plant. For heat exchangers, one of the major damage types during their service life is thermal creep, a time-dependent plastic deformation of materials subjected to constant or varying stress at elevated temperature. The main objective of this study is imitating the work condition of heat exchanger for alloy 617 samples and to inspect the damaged grain boundaries, explain the relationship of creep void nucleation and grain boundary characters and determine which boundaries are susceptible to damage and which are more resistant, in order to help improve its creep resistance in future manufacture. For this purpose, samples which had been crept at elevated temperatures were examined and quantitatively characterized for damage and grain boundary structure. Electron backscatter diffraction and orientation image mapping were used to measure the proportions of each boundary by observing and analyzing these crept microstructures. The grain-boundary distribution can be expressed in terms of five parameters that include: three parameters that describe the lattice misorientation across the boundary and two parameters that describe the orientation of the grain-boundary plane normal. Thus, a complete description of the grain-boundary character distribution of crept alloy 617 was obtained. The majority of grain boundaries are Σ3 boundaries. Three conditions were analyzed: the original material, metal that was annealed without stress, and one that was crept at 1000 C at 20 MPa and 25 MPa for various times. The crept sample resulted in an increase of Σ3 tilt boundaries and low angle boundaries (0 to 15 degree), which could be explained by the fact that these boundaries are susceptible to preferential creep damage under high stress. By observing the position of voids, it is found that the voids seldom occur at low angle grain boundaries, again confirming that low angle boundaries are more resistant to void formation.