Philip Edmondson, University of Oxford
Christopher Stanek, Los Alamos National Laboratory
Marjorie Bertolus, CEA/DEN
Heather McLean Chichester, Idaho National Laboratory
Symposium Support CEA DEN MINOS
EE2: Ceramics I
Monday PM, December 02, 2013
Hynes, Level 3, Room 309
2:30 AM - *EE2.01
Radiation Damage Evolution in Oxide Heterocomposites
Blas Uberuaga 1
1Los Alamos National Laboratory Los Alamos USAShow Abstract
It is well established that interfaces and grain boundaries can act as efficient sinks for radiation-induced defects. Exactly how interfaces interact with defects and how this interaction depends on both the structure of the interface and the radiation conditions, however, are still uncertain. Here, we examine coherent heterointerfaces in oxide thin film bilayers to determine how radiation-induced defects interact with those interfaces and modify the radiation tolerance of the material. While these particular interfaces are often nearly fully coherent, with no thermodynamic trap states at the interface, the interface nevertheless greatly influences how the materials on both sides respond to the produced defects. Both enhancement and degradation of radiation tolerance is observed in experimental studies of model oxide heterocomposites. This behavior is rationalized using atomistic calculations and mesoscale simulations via which differences in chemical potential and bulk migration properties of defects in each phase are hypothesized to be the controlling factors. We identify different regimes of defect evolution in irradiated composites that may provide new opportunities for developing radiation tolerant nanocomposites. We discuss the implications for composites more generally, such as nanostructured ferritic alloys (NFAs), that have potential applications in nuclear energy systems.
3:00 AM - EE2.02
Analysis of the Structure of Heavy Ion Irradiated Perovskites Using X-Ray Absorption Spectroscopy
Martin C Stennett 1 Amy S Gandy 1 Neil C Hyatt 1
1The University of Sheffield Sheffield United KingdomShow Abstract
Crystalline ceramics are one of a number of candidate materials for the immobilisation of radio-nuclides arising from the nuclear fuel cycle. In particular, ceramics have been suggested as the most promising option for containment of transuranics such as U, Pu and Am. Transuranic elements undergo decay by alpha particle emission and recoil of the parent nucleus. These recoil events causes disruption of the crystal lattice and after sufficient events many crystalline materials can be rendered amorphous. Little is known about the structure of the amorphised material and the subsequent effect on key wasteform properties. This research seeks to investigate radiation damaged in crystalline wasteform materials using a combination of spectroscopic techniques. Previous work by the authors has shown that the local environment of cations in titanate based ceramics changes significantly as a result of radiation induced damage, particularly the Ti, which was shown to change from six- to five-fold coordination. This contribution expands the study to investigate the behaviour in iron based materials specifically LaFeO3 and LaSrFeO4. Our approach involved heavy ion implantation of bulk ceramic samples, to simulate heavy atom recoil, combined with grazing angle X-ray absorption spectroscopy (GA-XAS) to characterise the resulting amorphised surface layer. Quantitative analysis was performed on the GA-XAS data to determine the change in valence and local co-ordination environment of cations in the amorphised surface layer.
3:15 AM - EE2.03
Thermal and Radiation Stability of Iodine-Bearing Vanadate Apatite Structure
Fengyuan Lu 1 Jinling Xu 1 Tiankai Yao 1 Rodney C Ewing 2 Jie Lian 1
1Rensselaer Polytechnic Institute Troy USA2University of Michigan Ann Arbor USAShow Abstract
The immobilization of the long-lived radiotoxic fission product I-129 into a durable waste form is important for effective nuclear waste management as iodine is highly mobile and has significant environmental and health concerns. The consolidation into a dense waste form without significant iodine loss is also a challenge due to the highly volatile nature of iodine. In this work, iodine bearing apatite Pb10(VO4)6I2 nanopowder was synthesized by high energy ball milling at room temperature with a high iodine loading (9 wt%). Dense iodine-apatite pellets were fabricated by spark plasma sintering (SPS) at 700 °C with a very short duration less than 3 minutes. The thermal stability of the iodine-bearing apatite powder and SPS densified pellets was studied by post-thermal annealing, and the iodine loss was investigated by thermal gravimetric analysis (TGA). The iodine apatite power is stable annealed at 200 and 300 °C with improved crystallinity and larger grain size. Phase decomposition occurred for apatite powder annealed at 400 °C, leading to significant iodine loss. In contrast, the SPS densified pellets are stable without phase decomposition or iodine loss at temperature up to 670 °C.
The radiation stability was investigated by energetic ion beam irradiations using 1 MeV Kr2+ irradiation under in-situ TEM observation. The as-milled Pb10(VO4)6I2 nanocrystals embedded in amorphous matrix can be easily amorphized at room temperature. The iodine-bearing Pb10(VO4)6I2 annealed at 300 °C exhibits enhanced radiation tolerance with a lower critical amorphization temperature (Tc) of 242 °C, as compared with lead/calcium vanadate fluorapatite (PbxCa1-x)10(VO4)6F2. SPS process further improves the radiation stability of Pb10(VO4)6I2 and the critical temperature for SPS densified pellets is reduced to 229 °C. The greater radiation tolerance of the iodine-bearing apatite is consistent with the enhanced crystallinity upon thermal annealing and SPS densification. These results indicate that SPS-fabricated Pb10(VO4)6I2 is a promising waste form for I-129 immobilization with greatly enhanced thermal and radiation stability and iodine confinement.
3:30 AM - EE2.04
A Many-Body Potential Approach to Modelling the Thermomechanical Properties of Actinide Oxides
Michael William Donald Cooper 1 Michael Rushton 1 Robin Grimes 1
1Imperial College London London United KingdomShow Abstract
UO2 has been studied widely since it is the basis of conventional reactor fuels. It can be blended with other actinide oxide powders, in particular PuO2, to form what is commonly called mixed oxide fuel (MOX). Alternative fuel cycles are being studied based on other combinations, notably with ThO2, since the rate of higher actinide breeding is much lower. Furthermore, the higher actinides are problematic for waste forms as they often have long half lives, therefore, the incorporation of minor actinides such as AmO2, CmO2 and NpO2 with UO2 or ThO2 is desirable so that these species can undergo transmutation in a reactor or accelerator driven system. The difficulty in reproducing the many-body effects of these actinide oxides (such as the Cauchy violation) using an empirical pairwise description of ionic interactions was proven a stumbling block for previous atomistic simulation studies, particularly when investigating thermomechanical properties, such as bulk modulus, over a broad temperature range.
In this work we present a novel approach to simulating actinide oxides by including many-body effects using the Embedded Atom Method. This ensures a good description of a range of thermophysical properties (lattice parameter, bulk modulus, enthalpy and specific heat) between 300 K and 3000 K for AmO2, CeO2, CmO2, NpO2, ThO2, PuO2 and UO2. The oxygen-oxygen interactions are fixed across the actinide oxide series to enable the simulation of MOX fuels. The new potential is also used to predict Schottky and Frenkel pair energies.
3:45 AM - EE2.05
Stabilizing Nanocrystalline Grains in Ceramic-Oxides
Dilpuneet Aidhy 1 Yanwen Zhang 1 2 William Weber 2 1
1Oak Ridge National Laboratory Oak Ridge USA2University of Tennessee Knoxville USAShow Abstract
The inherent grain-growth problem in nanocrystalline ceramic-oxides renders their highly attractive properties practically unusable, and controlling the nano-grain sizes continues to be an uphill task. We elucidate a framework to design dopant-pinned grain boundaries that prevent this grain growth. Using atomic simulations, we show that effective grain boundary pinning depends upon dopant-oxygen vacancy interactions, i.e., (a) dopant migration energy in the presence of oxygen vacancy, and (b) dopant-oxygen vacancy binding energy. Our prediction agrees with and explains previous experimental observations. This new concept is in complete contrast to the dopant-host atomic size mismatch concept prevalent in metallic systems, and elucidates that nanograin stabilizing concepts are not inter-transferable between metallic and ceramic-oxide systems.
This work was supported as part of the Materials Science of Actinides, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences. The computer simulations were performed at the National Energy Research Scientific Computing Center at Lawrence Berkeley National Laboratory.
4:30 AM - EE2.06
Accelerated Chemical Aging Studies to Assess the Impact of Daughter Product Formation on Crystalline Stability
Chris Stanek 1 Blas Uberuaga 1 Laura Wolfsberg 1 Wayne Taylor 1 Brian Scott 1 Nigel Marks 2
1Los Alamos National Laboratory Los Alamos USA2Curtin University of Technology Perth AustraliaShow Abstract
The effect of transmutation of radionuclides, especially “short-lived” Sr-90 and Cs-137, to chemically distinct daughter products (Zr and Ba respectively) will impact nuclear waste form stability. Due to the technical challenges associated with this studying problem, the topic of transmutation has received limited attention during the past 30 years of waste form development. In order to develop a predictive capability to design radiation tolerant and chemically robust nuclear waste forms, we must first address a fundament materials science question: What is the impact of daughter product formation on the stability of solids comprised of radioactive isotopes? To answer this question, a multidisciplinary approach integrating first principles modeling with the synthesis and characterization of small, highly radioactive surrogate samples has been instigated. We present the details of this approach as well as recent results for a range of materials systems, including: 109Cd1-xAgxS, 55Fe2-xMnxO3 and 177Lu2-xHfxO3.
4:45 AM - EE2.07
Synthesis and Characterization of 177Lu2-xHfxO3
Jeffery Aguiar 1 Laura Wolfsberg 1 Brian L Scott 1 Wayne A Taylor 1 Rob Dickerson 1 Christopher Stanek 1
1Los Alamos National Laboratory Los Alamos USAShow Abstract
Transmutation of constituents may offer a novel approach to synthesize compounds far from equilibrium conditions - a phenomenon we refer to as radioparagenesis. Especially in the cases isotopes that decay via beta - or electron capture, transmutation leads to significant changes in the valence and radius of the transmuting ion, often resulting in a daughter product that is incompatible with the original parent crystal structure. However, exploring these issues for “short-lived” fission products of interest is not feasible due to the ~30 year half-life, and previous experimental approaches to accelerate the process focused on former isotopes with shorter half-lives via neutron activation were inconclusive. In this work, we present a new accelerated chemical aging approach, which combines density functional theory calculations with experiments on isotopically pure samples to investigate radioparagenesis under well-defined. We will specifically present recent experiments using aberration corrected transmission electron microscopy (TEM), energy dispersive X-ray and electron energy loss spectroscopies to study the synthesis and characterization of 177Lu2O3. 177Lu decays via beta minus to 177Hf with a 6 day half-life. We present experimental results of the impact of Hf formation on the structural stability of bixbyite Lu2O3. These results are compared to complementary DFT calculations, which ultimately will allow for predictions of structural stability as function of compositional evolution.
5:00 AM - EE2.08
Understanding Structure-Property Relationships in beta;-eucryptite through Atomistic Simulations
Badri Narayanan 1 Ivar E Reimanis 1 Cristian V Ciobanu 2
1Colorado School of Mines Golden USA2Colorado School of Mines Golden USAShow Abstract
The study of materials with unusual properties offers to provide new insights into structure-property relationships and promise in the design of novel composites with tailored properties. In this spirit, we have chosen to study β-eucryptite, a technologically relevant lithium aluminum silicate that exhibits negative thermal expansion (NTE), radiation tolerance and pressure-induced amorphization (PIA) under moderate applied pressures. These exotic physical properties make β-eucryptite suitable for various specific applications like heat exchangers, ring laser gyroscopes, and nuclear breeder reactors. Using density functional theory calculations [Narayanan et al., Phys. Rev. B 81, 104106 (2010)], we found that the linear compressibility of β-eucryptite along the c-axis is positive consistent with recent ultrasonic experiments, as opposed to a negative value reported by earlier direct measurements. More importantly, this finding indicated that the NTE behavior in β-eucryptite occurs due to tetrahedral tilting and cation disordering rather than elastic effects arising from negative compressibility. Recently, our reactive force field (ReaxFF) molecular dynamics [Narayanan et al., J. Appl. Phys. 113, 033504 (2013)] showed that at radiation doses below 0.21 displacements-per-atom or less, β-eucryptite retains its long-range crystalline order while exhibiting tetrahedral tilting, change in atomic coordination around Al/Si and disordering of Li atoms. Furthermore, upon thermal annealing, most of the under-coordinated Si-polyhedra formed during radiation regained their tetrahedral coordination via a mechanism involving tilting of Al-, and Si-centered polyhedra. Our metadynamics simulations based on ReaxFF revealed that β-eucryptite begins to amorphize under moderate pressure ~3 GPa close to empirically known transition pressure [Narayanan et al., submitted to Appl. Phys. Lett. (2013)]. We also identified the atomic scale mechanisms underlying PIA in β-eucryptite that consist of (a) progressive tetrahedral tilting that eventually results in change in O-coordination around several Al atoms (~41.7%) while keeping SiO4 intact, and (b) spatial disordering of Li atoms forming Li-Li, Li-O and Li-O-Li linkages. We show that the atomic-scale processes in β-eucryptite induced by thermal, radiation, and pressure environments arise from the inherent flexibility of the three-dimensional network of corner-sharing AlO4 and SiO4 tetrahedra. These results will be discussed in the context of a possible trend between NTE, radiation tolerance and PIA under moderate pressure in flexible framework structures.
5:15 AM - EE2.09
Effect of Neutron Irradiation on Select Mn+1AXn Phases
Darin Joseph Tallman 1 Elizabeth Hoffman 2 Gordon Koshe 3 Robert L Sindelar 2 Michel W Barsoum 1
1Drexel University Philadelphia USA2Savannah River Site Aiken USA3Massachusetts's Institute of Technology Cambridge USAShow Abstract
Gen IV nuclear reactor designs require materials that can withstand long term operation in extreme environments of elevated temperatures, corrosive media, and fast neutron fluences (E>1MeV) with up to 100 displacements per atom (dpa). Full understanding of irradiation response is paramount to long-term, reliable service. The Mn+1AXn phases have recently shown potential for use in such extreme environments because of their unique combination of high fracture toughness values and thermal conductivities, machinability, oxidation resistance, and ion irradiation damage tolerance. Herein we report, for the first time, on the effect of neutron irradiation of up to 0.5 dpa at 70°C and 700 °C on Ti3AlC2, Ti2AlC, Ti3SiC2, and Ti2AlN. Evidence for irradiation induced dislocation loops and their effect on electrical resistivity is also presented. X-ray diffraction refinement of the resultant microstructures is provided. Based on the totality of our results, it is reasonable to assume that the MAX phases, especially Ti3AlC2, are very promising materials for high temperature nuclear applications.
5:30 AM - EE2.10
Surface Sensitive Spectroscopy Study of Ion Beam Irradiation Induced Structural Modifications in Iron Borophosphate Glasses
Amy S Gandy 1 Martin C Stennett 1 Neil C Hyatt 1
1University of Sheffield Sheffield United KingdomShow Abstract
Iron phosphate glasses are being considered as an immobilisation matrix for high level nuclear waste (HLW), including minor actinides and plutonium residues, due to their high chemical durability and ability to incorporate diverse chemical compositions. Iron borophosphate glasses are of particular interest as the addition of boron, which has a high thermal neutron absorption cross-section, increases glass thermal stability and provides criticality control. Incorporated actinides undergo α-decay, resulting in the formation of α-particles (MeV He nuclei) and energetic (~100KeV) daughter recoil nuclei. Interactions between recoil nuclei and glass atoms results in atomic displacements which form collision cascades, potentially altering glass network polymerisation and cation valance states. Such modifications can affect glass durability and long-term performance as an immobilisation matrix. In this study, heavy ion implantation (e.g. 2MeV Kr or 2MeV Au irradiation) was used as an analogue for α-recoil damage. Iron borophosphate glasses, with nominal molar composition 60P2O5 - (40 - x) Fe2O3 - xB2O3 (x = 0, 10, 20) were irradiated at room temperature, producing a damaged region extending from the surface to a depth of approximately 1µm. To probe exclusively the damaged region, surface sensitive techniques were employed. The effects of simulated α-recoil damage were investigated by probing the speciation and valence of Fe, and by examining the glass structure. In this contribution, we report on structural and chemical modifications as a consequence of heavy ion irradiation, elucidated using Reflectance Fourier-Transform Infrared (FT-IR), Mossbauer, and X-ray absorption spectroscopies.
EE1: Fuel Cladding Materials
Monday AM, December 02, 2013
Hynes, Level 3, Room 309
9:30 AM - *EE1.01
Understanding Environmental Degradation of Zr Cladding
Anton Van der Ven 1 John C. Thomas 2 Brian Puchala 2
1University of California Santa Barbara Santa Barbara USA2University of Michigan Ann Arbor USAShow Abstract
Predicting high temperature thermodynamic and kinetic properties of materials for nuclear applications from first principles remains a major challenge. Important properties of materials used in nuclear applications include their resistance to degradation and chemical corrosion. The corrosion of nuclear materials involves surface and interface reactions, electronic and ionic transport and the occurrence of a variety of phase transformations, all driven by extreme chemical and mechanical driving forces. First-principles statistical mechanical methods are now capable of predicting a wide variety of thermodynamic and kinetic properties as well as the couplings between chemistry and mechanics that determine the rate and mechanisms of degradation of structural materials. In this talk, we will describe how corrosion of Zr in aqueous environments can be modeled from first principles. The approach relies on the use of first-principles parameterized effective Hamiltonians that rigorously account for all relevant atomic and electronic excitations. A combination with Monte Carlo techniques allows the statistical mechanical prediction of finite temperature thermodynamic and kinetic properties relevant to corrosion processes in nuclear materials.
10:00 AM - EE1.02
Kinetics of Hydrogen Desorption from Zirconium Hydride and Zirconium Metal in Vacuum
Xunxiang Hu 1 2 Kurt A. Terrani 2 Brian D. Wirth 1
1University of Tennessee Knoxville Knoxville USA2Oak Ridge National Laboratory Oak Ridge USAShow Abstract
Given an optimized set of neutronic and mechanical properties, zirconium alloys play a very important role in the nuclear field, as fuel cladding and by default as a barrier against radioactive material release during used fuel storage. Zirconium hydride formed in normal operation and accident scenarios is a major concern, and in particular, hydrogen behavior during vacuum annealing of used nuclear fuel, in addition to other de-hydriding processes, is an area of significant interest.
We describe the hydrogen desorption behavior from zirconium hydride and zirconium metal in vacuum observed during coordinated experimental and modeling activities. A δ-zirconium hydride is produced in an oxygen-free tube furnace from Zircaloy-4. The resulting hydride phase and hydrogen concentration have been verified by x-ray diffraction, weight change and gas desorption. Subsequently, the kinetics of hydrogen during thermal processessing has been studied using Thermal Desorption Spectroscopy (TDS) to directly obtain the hydrogen desorption spectra of δ-zirconium hydride as a function of initial conditions under a pre-determined temperature profile. The TDS results have been analyzed and compared to a one-dimensional, two-phase moving boundary model coupled with a kinetic description of hydrogen desorption from a two-phase region of δ-zirconium hydride and α-zirconium. The model and experimental comparison demonstrates the ability to successfully reproduce the TDS experimental results, which validates the assumption of zeroth-order and second-order hydrogen desorption kinetics for δ-zirconium hydride and α-Zr, respectively.
This study provides fundamental insights into the behavior of hydrogen and zirconium hydride, in addition to demonstrating a modeling paradigm to predict the performance of the hydride fuel and the cladding failure under vacuum annealing of used nuclear fuel.
10:15 AM - EE1.03
Ductility Evaluation of As-Hydrided and Hydride Reoriented Zircaloy-4 Cladding under Simulated Dry-Storage Condition
Yong Yan 1 Kenyong Plummer 1 Holly Ray 1 Tyler Cook 1
1Oak Ridge National Laboratory Oak Ridge USAShow Abstract
Fuel cladding is the first barrier for retention of fission products and nuclear fuel. Safety analyses of dry casks containing high-burnup light water reactor (LWR) fuel require measurement of cladding mechanical properties in order to better understand fuel behavior. Pre-storage drying-transfer operations and early stage storage expose cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to normal operation in-reactor and pool storage. Under these conditions, radial hydrides could precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature. As a means of simulating this behavior, hydrided Zircaloy-4 samples were fabricated at Oak Ridge National Laboratory (ORNL) by a gas charging method to levels that encompass the range of hydrogen concentrations observed in current used fuel. For low hydrogen content samples, the hydrided platelets appear elongated and needle-like, orientated in the circumferential direction. In addition, a hydride reorientation system was developed at ORNL to simulate the effects of drying-storage temperature histories. Mechanical testing was carried out by the ring compression test (RCT) method at various temperatures to evaluate the sample&’s ductility for both as-hydrided and hydride reorientation treated specimens. As-hydrided samples with higher hydrogen concentration resulted in lower strain before fracture and reduced maximum load. The trend between temperature and ductility was very clear: increasing t