Symposium Organizers
Karl Whittle, University of Liverpool
Felix Brandt, Forschungszentrum Juelich
Philip D Edmondson, Oak Ridge National Laboratory
Blas Uberuaga, Los Alamos National Laboratory
ES08.01: Carbides and Graphite
Session Chairs
Izabela Szlufarska
Karl Whittle
Monday PM, November 27, 2017
Hynes, Level 2, Room 206
8:45 AM - ES08.01.01
In Situ Irradition of Binary Carbide Hybrids—The Effect of Composition
Tanagorn Kwamman 2 , Philip D Edmondson 3 , Karl Dawson 1 , Mark Rainforth 2 , Karl Whittle 1
2 Materials Science and Engineering, The University of Sheffield, Sheffield United Kingdom, 3 Physical Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 1 School of Engineering, University of Liverpool, Liverpool United Kingdom
Show AbstractCeramic-ceramic, or ceramic-metallice hybrids are being increasingly studied for their applications within the nuclear context, for example silicon carbide has examined for its use within nuclear reactors, both fission and fusion based, both as single or binary based systems such as SiC/SiC. Further hybrids, such as those based on M(n+1)AXn (MAX phases), e.g. Ti3SiC2 and Ti3AlC2 have been investigated with mixed results, along with SiC-ZrC based binary carbides.
A binary carbide based on TiC-SiC has been fabricated and irradiated at the IVEM-TANDEM facility, to high levels of damage, with the damage monitored in situ. The damaged samples once irradiated have been analysed using high resolution TEM imaging, electron diffraction, and transmission Kikuchi diffraction. The results show that titanium and silicon carbide both have differing responses to damage, with TiC differently to damage than SiC, which has implications for similar systems such as ZrC-SiC.
9:00 AM - *ES08.01.03
Effects of Irradiation on Evolution of Atomic Structure and Composition of Grain Boundaries in Silicon Carbide
Izabela Szlufarska 1 , Hao Jiang 1 , Xing Wang 1 , Tomonori Baba 1
1 , University of Wisconsin, Madison, Wisconsin, United States
Show AbstractWhile grain boundaries (GBs) can act as sinks of defects in irradiated materials, much less is known about the effects of radiation on the evolution of atomic structure and consequently on the sink strength of GBs. In addition, in multi-component materials, such as silicon carbide, there can be unbalanced flux of different species to the GBs, raising questions about potential compositional changes of GBs under irradiation and the effects of these changes on the continuing ability of GBs to absorb defects. Here, using a combination of density functional theory, molecular dynamics, rate theory equations, and kinetic Monte Carlo we have shown how tilt and twist GBs in SiC evolve under irradiation. In the case of tilt GBs, we have found conditions under which the mechanism for defect accommodation in GBs transitions from dislocation climb to diffusion along GBs to other sinks (e.g., triple junctions). Effects of irradiation on twist GBs have been described in terms of evolution of GB dislocation networks, and we have demonstrated how twist GBs can continue absorbing defects without saturation, in spite of the defect flux being off-stoichiometric. Defect kinetics and chemical changes taking place in GBs during irradiation have been also investigated using a combination of scanning transmission electron microscopy and electron energy loss spectroscopy. We found that GBs in CVD-SiC are intrinsically carbon-poor and the relative C composition can be as low as 45%. This carbon depletion disappeared in samples irradiated to 1 dpa at 300 C, which is likely due to the unbalanced flux of C interstitials to defect sinks and low diffusivity of defects at GBs. Interestingly, the C depletion appeared again in samples irradiated at 600 C and the depletion was found to more significant than in the non-irradiated sample. The off-stoichiometry of SiC is surprising given that it is a line compound. These results indicate that the role of GBs may change from defect clustering reservoirs at low irradiation temperatures to defect diffusion channels at high irradiation temperatures. Changes in GBs chemistry due to irradiation are expected to impact such properties of SiC as its fracture strength, corrosion resistance, as well as the ability of this material to provide a barrier to diffusion of fission products in nuclear reactor applications.
9:30 AM - ES08.01.04
Hélium Behavior in Nano-Polycrystalline Silicon Carbide
Nathalie Millard-Pinard 1 , Joffrey Baillet 1 , Stéphane Gavarini 1 2 , Vincent Garnier 3 , Christophe Peaucelle 1 , Xavier Jaurand 2 , Antony Duranti 1 , Clement Bernard 1 , Romain Rapegno 1 , Luis Escobar Sawa 1
1 , Univ Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, IPNL, F-69622, Lyon France, 2 , Univ Lyon, Université Lyon 1, CTµ, F-69622, Villeurbanne Cedex France, 3 , Univ Lyon, INSA de Lyon, MATEIS CNRS UMR5510, F-69621, Villeurbanne France
Show AbstractSilicon carbide (SiC) is a good candidate for nuclear applications, fusion reactors as well as Gen IV reactors, because of its interesting properties, such as low activation, high temperature resistance and chemical inertness.
3C-SiC samples were prepared by Spark Plasma Sintering (SPS) from a 15-nm-nanopowder. A microstructure with average grain size of about 70 nm was obtained with a densification ratio of about 95%. SiC pellets were then implanted at room temperature with 30 keV 3He++ ions at three fluencies: 5.1015at.cm-2 , 1.1017 at.cm-2 and 1.1018 at.cm-2 to simulate helium produced during fusion or fission process. In thses conditions of implantation, the projected ion range (Rp) is about 160 nm and the maximal concentration is near 0.5, 9.7 and 52 at.% for each fluence respectively. SiC Pellets were then characterized by ion beam analysis and electronic microscopy to determine compositional and morphological evolution. The composition of the material is almost not modified after implantation at low fluence but a strong oxidation occurred at the highest fluence. Helium is almost completely retained in the material up to 1.1017 at.cm-2 . TEM analysis showed a complete amorphization of the material for the two highest fluences and the formation of SiO2 phases together with remaining SiC within the implanted region. Noticeable changes in surface morphology are observed after implantation at the highest fluence with a global increase of surface roughness due to the apparition of blisters. Nanometric bubbles are formed at a fluence of 1.1017 at.cm-2 and these let place to huge gas bubbles near Rp for the highest fluence of 1.1018 at.cm-2.
9:45 AM - ES08.01.05
Hydrothermal Corrosion of Environmental Barrier Coatings on Silicon Carbide in Boiling Water Reactor (BWR) Conditions
Stephen Raiman 1 , Peter Doyle 2 1 , Caen Ang 1 , Kurt Terrani 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 2 , University of Tennessee, Knoxville, Knoxville, Tennessee, United States
Show AbstractSiCf/SiC composite materials are attractive candidates for use as advanced fuel cladding, combining excellent neutronic properties with suitable mechanical strength and resistance to oxidation in accident scenarios. However, SiCf/SiC has been found to be susceptible to aqueous dissolution in LWR coolant environments. To address this issue, environmental barrier coatings have been developed to mitigate dissolution during normal operating conditions.
For this study, Cr, CrN, TiN, ZrN, and NiCr coatings on SiC substrates were exposed for up to 400h in a constantly-refreshing autoclave at 288°C water with either 150 wppb DH or 2 wppm DO to simulate BWR-HWC and BWR-NWC. After exposure, samples were analyzed for mass change, and the environmental barrier coatings were analyzed to determine their effectiveness in mitigating corrosion. Cr and CrN coated samples performed favorably when compared to uncoated SiCf/SiC in both BWR environments. Samples coated with TiN, ZrN, and NiCr exhibited significant corrosive attack, including rapid oxidation and spallation, and were found to be less suitable as environmental barrier coatings.
This work was funded by U.S. Department of Energy’s Office of Nuclear Energy, Advanced Fuel Campaign.
10:30 AM - *ES08.01.06
Understanding Changes to Graphite Properties in Nuclear Reactor Environments
Anne Campbell 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractGraphite is the primary structural and moderating material for the gas cooled and molten salt advanced nuclear reactors. Graphite was the first “nuclear material” in that it was both the moderator and structural material for the first man-made critical pile (CP-1 in Chicago). Over the last 75 years graphite has continued to be used for multiple nuclear related applications, and as such a large repository of data is available that allows for limited prediction of the behavior of a new type of graphite. But, of this large data set, almost all the data is engineering related so very little information is available that describes the microstructural changes that are the controlling mechanisms for the various macroscopic property changes. This presentation will first discuss the current knowledge regarding the response of graphite when used in nuclear reactor environments, including the changes that occur to the physical, thermal, and mechanical properties. The primary focus of this presentation will be the traditional (i.e. optical microscopy, X-ray diffraction, and Raman microscopy, gas porosimetry) and cutting-edge techniques (i.e. elastic and inelastic neutron scattering, slice and view in SEM, X-ray tomography) that are being used to try to understand the microscopic/macroscopic relationships.
This research was performed at the Oak Ridge National Laboratory (ORNL) and sponsored by Tokai Carbon Co., Ltd. (NFE-09-02345) and IBIDEN Co., Ltd. (NFE-11-03389), with the U.S. Department of Energy. A portion of this research at ORNL’s High Flux Isotope Reactor and the Spallation Neutron Source was sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, US Department of Energy. Oak Ridge National Laboratory is managed by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 for the U.S. Department of Energy
11:00 AM - ES08.01.07
Real Time 3D X-Ray Computed Tomographic Imaging of the Mixed-Mode Fracture of Nuclear Graphite at 1000°C
Dong Liu 1 , Jon Ell 2 , Bernd Gludovatz 2 , Harold Barnard 2 , Robert Ritchie 2
1 , University of Oxford, Oxford United Kingdom, 2 , Lawrence Berkeley National Laboratory, Berkeley, California, United States
Show AbstractGilsocarbon graphite is a neutron moderator as well as a structural core component in the Advanced Gas-cooled Reactors (AGRs) in the U.K. It is a medium-grained, near-isotropic synthetic polygranular graphite composite material with microstructural features such as pores and defects ranging from the nanometre to macro-scale. There are thousands of graphite components in the core of the AGRs forming the fuel channels, control rod channels and reflectors, all of which are irreplaceable and hence life limiting. As graphite is also the material to be used in several designs of future higher-temperature Gen IV reactors, it is essential to assure the structural integrity of these graphite components. To predict potential fracture and to mitigate premature failure of the graphite components, a faithful mechanical response of graphite has to be evaluated in real time, in three-dimensions, under load, and at service-relevant temperatures. However, due to experimental limitations all the in situ 3-D tomography work up to date has been performed at ambient temperatures. These experiments can never adequately describe the mechanical behaviour and damage evolution in graphite at realistic temperatures (~650°C for AGRs; ~1000°C outlet temperature for Gen IV reactors).
This paper provides an update on the most recent advances made in the in situ high-temperature mechanical testing of Gilsocarbon graphite using synchrotron X-ray computed micro-tomography. A unique in situ ultrahigh temperature tomography rig that permits real-time investigation of damage evolution under load at temperatures up to 1700°C was adopted. In this paper, we summarise our work so far on the flexural strength, fracture toughness (resistance curves) under pure Mode I and mainly on the mixed-mode (KI and KII) conditions at both ambient and elevated temperatures (650°C and 1000°C). We find that the strength and mode I fracture toughness both increase with temperature. We have attributed this to the relaxation of residual stresses that are ‘frozen-in’ during the manufacturing process. The mixed-mode fracture toughness of graphite, which is more related to the realistic load-bearing condition in a reactor core, was found to follow the same trend. As X-ray tomography provides the unique insight to the 3-D microstructure of the material under deformation, digital volume correlation (DVC) was applied to investigate the local formation of damage and microcracking in the material at 1000°C. Different toughening mechanisms at ambient and high temperature will be discussed.
11:15 AM - ES08.01.09
FIB-SEM Tomography for Carbon Base Materials, Graphite and SiC/PyC Composites
Jose Arregui-Mena 1 , Philip D Edmondson 1 , Tyler Gerczak 1 , Anne Campbell 1
1 , Oak Ridge National Laboratory, Knoxville, Tennessee, United States
Show AbstractFIB-SEM tomography is a technique that combines the sequential milling and SEM scanning of an area of interest to produce a stack of images. The images generated by FIB-SEM tomography are normally processed and segmented to produce 3D models of the phases of a given material. Two materials used in the nuclear industry, graphite and SiC/PyC composites were analyzed with this technique.
The first case of study analyzed the pore structure of AGX graphite an electrode graphite. Graphite is a structural component and moderator of fast neutrons in nuclear reactors as well as an electrode for multiple industrial applications. Voids characteristics, connectivity and volume fraction influence most of the mechanical properties, fracture mechanics and oxidation rate of graphite. Therefore, it is important to understand the porosity structure of graphite to predict its behavior under different environmental and loading conditions. In this part of the study the authors use FIB-SEM tomography to identify the structure, connectivity, size and content of porosity of two regions in the sample.
The second case study focuses on measuring the degree of interfacial stitching and porosity at a planar PyC/SiC interface. The planar PyC/SiC interface in this study is optimized for a diffusion couple experiment, however, it is fabricated by the same conditions and methods used to fabricate tristructural-isotropic (TRISO) coated particle fuel to represent the layers in TRISO fuel. Measuring the interfacial stitching and porosity is critical to understand the interfacial reactions that occur with fission products in the system as the SiC layer is a primary release barrier to fission products not retained in the kernel of TRISO fuel.
11:30 AM - ES08.01.10
TMSR Materials Development—Carbide Dispersed Strengthening Nickel Based Alloys
Hefei Huang 1 , Jie Gao 1 , Chao Yang 1 , Massey De Los Reyes 2 , Xingtai Zhou 1
1 , Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai China, 2 , Australian Nuclear Science and Technology Organisation, Lucas Heights, New South Wales, Australia
Show AbstractThe development of high-temperature irradiation-resistant nickel-based alloys has been receiving much attention due to their potential applications in molten salt reactors (MSRs). Silicon carbide nanoparticle-reinforced nickel-based composites (Ni-SiCNP), with milling time ranged from 8 to 48 h, were prepared using mechanical alloying and spark plasma sintering. In addition, unreinforced pure nickel samples were also prepared for comparative purposes. The microstructure of the Ni–SiCNP composites was characterized by TEM and their mechanical properties were investigated by tensile measurements. The TEM results showed well-dispersed SiCNP particles, either within the matrix, between twins or along grain boundaries (GB), as well as the presence of stacking faults and twin structures, characteristics of materials with low stacking fault energy. The tensile test results indicated that the addition of SiCNP can effectively strengthen the nickel. Furthermore, the helium diffusion behavior of such composites and pure nickel under 3 MeV helium ion irradiation at 600 °C with ion fluence up to 3×1020 ions/m2 has also been studied. The TEM results indicated that the presence of dispersed SiCNP in nickel can inhibit the growth of helium bubbles, thereby mitigate the helium embrittlement and swelling of nickel-based alloys. The theoretical calculation results using the density functional theory (DFT) showed that the helium atoms prefer to diffuse to the interface between SiCNP and nickel matrix, and thus avoid the grain boundary segregation and also the growth of helium bubbles. This study confirmed the feasibility of dispersing carbides in nickel-based alloys to improve the irradiation-resistant performance of materials.
ES08.02: Nuclear Fuel I
Session Chairs
Christopher Stanek
Michael Tonks
Monday PM, November 27, 2017
Hynes, Level 2, Room 206
1:30 PM - ES08.02.01
Track of Fission-Fragment Energy Deposition in UO2 Using Electron and Phonon Multiscale Analyses
Woong Kee Kim 1 , Corey Melnick 2 , Ji Hoon Shim 1 , Massoud Kaviany 2
1 , Pohang University of Science and Technology, Pohang Korea (the Republic of), 2 , University of Michigan–Ann Arbor, Ann Arbor, Michigan, United States
Show AbstractEnergetic charged fission fragments are decelerated by the UO2 electronic net and in turn this energy is transferred from the conduction electrons to the lattice causing local drastic distortions which are in turn relaxed by heat diffusion leaving an atomic disordered track radius. While the electronic stoppage has been relatively well understood, the electron-phonon coupling in UO2 has not be resolved party due to the rather metallic behavior of the drastically compressed and distorted atomic region. Here we use ab initio, molecular dynamic and mesoscale (using transient two-temperature electron-phonon coupling model) treatments to explore these metallic behavior and properties and predict the tracking radius marked by atomic distortion caused by fission-fragment (swift heavy ion) energy dissipation through collision cascade. Good agreement is found with the available measured tracking radius.
1:45 PM - ES08.02.02
Off-Stoichiometric Cluster Dynamics in Irradiated Oxides
Sarah Khalil 1 , Todd Allen 2 , Anter El-Azab 3
1 , University of Alexandria, Alexandria Egypt, 2 , University of Wisconsin–Madison, Madison, Wisconsin, United States, 3 School of Materials Engineering and School of Nuclear Engineering, Purdue University, West Lafayette, Indiana, United States
Show AbstractA cluster dynamics model describing the formation of vacancy and interstitial clusters in irradiated oxides has been developed. The model, which tracks the composition of the oxide matrix and the defect clusters, was applied to the early stage formation of voids and dislocation loops in UO2, and the effects of irradiation temperature and dose rate on the evolution of their densities and composition was investigated. The results show that Frenkel defects dominate the nucleation process in irradiated UO2. The results also show that oxygen vacancies drive vacancy clustering while the migration energy of uranium vacancies is a rate-limiting factor for the nucleation and growth of voids. In a stoichiometric UO2 under irradiation, off-stoichiometric vacancy clusters exist with a higher concentration of hyperstoichiometric clusters. Similarly, off-stoichiometric interstitial clusters form with a higher concentration of hyperstoichiometric clusters. The UO2 matrix was found to be hyper-stoichiometric due to the accumulation of uranium vacancies.
2:00 PM - *ES08.02.03
Structural Features of Aliovalent Substitutions in UO2
Gianguido Baldinozzi 1 2 , Luis Casillas 3 , David Simeone 2 1 , Lionel Desgranges 5 , Henry Fischer 6 , Maulik Patel 4 , Kurt Sickafus 3
1 SPMS, CNRS, Gif-sur-Yvette France, 2 DMN, SRMA, CEA, Gif-sur-Yvette France, 3 MSE, University of Tennessee, Knoxville, Knoxville, Tennessee, United States, 5 DEC SESC, CEA, Saint Paul les Durance France, 6 , Institut Laue-Langevin, Grenoble France, 4 Mechanical Materials and Aerospace, University of Liverpool, Liverpool United Kingdom
Show AbstractInnovative fuels were and are developed with a number of technical objectives, including: more efficient utilization of fissile and fertile materials; enhanced proliferation resistance through passive control of nuclear materials using new fuel types and configurations; greater reliance on passive safety features; and technology advances to mitigate the volume and radiotoxicity of high level and long lived wastes. Those technical innovations often involve chemical changes that need to be characterized and modelled by fundamental approaches providing a reliable ground for redefining the regulatory requirements, industrial codes, and standards. A better understanding of the fundamental phenomena underpinning the physical, structural, and chemical properties of the mixed oxides of U, and more specifically in oxidizing condition, is also highly desirable for the goal of a safe and sustainable management of the spent reactor fuels.
In this talk we will try to address and review some of the structural changes occurring in UO2 induced by aliovalent substitutions on the metal sublattice, with particular emphasis on the lanthanides.
2:30 PM - ES08.02.04
Synthesis and Characterization of d-UO2 Nanoparticles for Nuclear Fuel Microanalysis
Samuel Briggs 1 , Bonnie Klamm 1 , Ryan Hess 1 , Khalid Hattar 1
1 , Sandia National Laboratories, Albuquerque, New Mexico, United States
Show AbstractUranium dioxide (UO2) fuel in the form of ceramic pellets has a long history of use as the preferred fuel form for commercially-operated thermal, light water reactors (LWRs). In this history, fabrication of these fuels has been dominated by the standardized processing route from uranium ores to the bulk ceramic fuel form currently employed. As such, fabrication and potential applications of different forms of UO2 prepared via novel synthesis routes has not been extensively explored. This work details the process by which depleted uranium dioxide (d-UO2) nanoparticles have been synthesized by polymer assisted deposition (PAD) and subsequent annealing, and explores the effect of annealing in neutral, oxidizing, and reducing atmospheres on the resulting nanoparticle phase and particle morphology. These nanoparticles have a variety of potential applications especially if tailored stoichiometries and microstructures can be achieved, including small-scale LWR fuel property assessment, study of interaction with fuel cladding or other material systems, and nuclear forensic analysis applications.
In the present study, d-U(IV)O2 nanoparticles were synthesized using an aqueous solution consisting of 0.1M uranyl nitrate (UO2(NO3)2), 0.125M polyethylenimine (PEI) and 0.125M ethylenediaminetetraacetic acid (EDTA). This solution was drop-casted on silicon nitride transmission electron microscopy (TEM) support films and annealed at 1000 °C for one hour in either Ar (neutral), air (oxidizing), or Ar-H (reducing) atmospheres. The phase of the resulting nanoparticles was investigated by both powder X-ray diffraction (PXRD), selected-area electron diffraction (SAED), and precession electron diffraction (PED). Standard diffraction contrast-based TEM imaging techniques were employed to investigate nanoparticle morphology. In all cases, some form of UO2±x phase was formed under sintering in all atmospheres, though the structure of the nanoparticles formed varied significantly between different sintering environments. The radiation tolerance of these particles was probed using TEM imaging with in-situ ion irradiation, as the stability of these particles under irradiation is essential if they are to be used to model fuel behavior in nuclear reactor environments.
2:45 PM - ES08.02.05
Quantifying the Impact of High Thermal Conductivity Phases on the Effective Thermal Conductivity of UO2 Fuel Pellets
Michael Tonks 1 , Floyd Hilty 1
1 Department of Materials Science and Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractUranium Dioxide has a low thermal conductivity that gets even lower during reactor operation, due to the generation of defects and fission products. However, metallic fission product precipitates can raise the effective thermal conductivity. In addition, additives such as beryllium oxide and silicon carbide are being added to fuel pellets to raise their effective thermal conductivity and improve the accident tolerance of the fuel. However, the amount that these high thermal conductivity phases raise the effective thermal conductivity of the fuel depends on the volume fraction, the topology, and the interfaces between the phase and the fuel. In this work we use experimental data and 3D simulations to inform the development of a mechanistic equation that predicts the increase in the thermal conductivity of the fuel, considering the volume fraction, the connectivity, and the interface of the secondary phase.
ES08.03: Fusion Materials
Session Chairs
Monday PM, November 27, 2017
Hynes, Level 2, Room 206
3:30 PM - *ES08.03.01
Interstitial-Mediated Dislocation Climb
Steve Fitzgerald 1
1 Department of Applied Mathematics, University of Leeds, Leeds United Kingdom
Show AbstractDislocations can climb out of their glide plane by absorbing (or emitting) point defects (vacancies and self-interstitial atoms). In contrast with conservative glide motion, climb relies on the point defects’ thermal diffusion and hence operates on much longer timescales, leading to some forms of creep. Whilst equilibrium point defect concentrations allow dislocations to climb to relieve non-glide stresses, point defect supersaturations also lead to osmotic forces, driving dislocation motion even in the absence of external stresses. Self-interstitial atoms typically have significantly higher formation energies than vacancies, so their contribution to climb is usually ignored [see e.g. 1-3]. However, under irradiation conditions, both types of defect are athermally created in equal numbers. In this work, we use simple thermodynamic arguments to show that the contribution of interstitials cannot be neglected in irradiated materials, and that the osmotic force they induce is many orders of magnitude larger than that caused by vacancies. This explains why the prismatic dislocation loops observed by in situ transmission electron microscope irradiations are often of interstitial rather than vacancy character. We also investigate, via discrete dislocation dynamics simulations, the effect on dislocation-obstacle interactions and its relevance to radiation creep.
References
[1] D Raabe, Phil. Mag. 77 p751 (1998)
[2] D Mordehai et al. Phil. Mag. 88 p899 (2008)
[3] B Bako et al, Phil. Mag. 91 p3173 (2011)
4:00 PM - ES08.03.02
Deformation Mechanism of Small-Volume Copper Containing High Density of Helium Bubbles
Weizhong Han 1
1 , Xi'an Jiaotong University, Xi'an China
Show AbstractThe workability and ductility of metals usually degrade with exposure to irradiation, hence the phrase “radiation damage”. Here, we found that Helium (He) radiation can actually enhance the room-temperature deformability of submicron-sized copper. In particular, Cu single crystals with diameter of 100 nm to 300 nm and containing numerous pressurized sub-10 nm He bubbles, become stronger, more stable in plastic flow and ductile in tension, compared to fully dense samples of the same dimensions that tend to display plastic instability (strain bursts). The sub-10 nm He bubbles are seen to be dislocation sources as well as shearable obstacles, which promote dislocation storage and reduce dislocation mean free path, thus contributing to more homogeneous and stable plasticity. Failure happens abruptly only after significant bubble coalescence. Furthermore, we discover that the helium bubble not only can coalesce with adjacent bubbles, but also can split into several nanoscale bubbles under tension. Alignment of the splittings along a slip line can create a bubble-free-channel, which appears softer, promotes shear localization, and accelerates the failure in shearing-off mode. Detailed analyses unveil that the unexpected bubble fragmentation is mediated by the combination of dislocation cutting and internal surface diffusion, which is an alternative micro-damage mechanism of helium irradiated copper besides the bubble coalescence. These results shed light on plasticity and damage developments in metals. Ref. Ding MS et al. PRL 117 (2016) 515501 and Nano Lett.16 (2016)4118.
4:15 PM - ES08.03.03
Modeling of Surface Morphological Evolution of Helium-Ion-Irradiated Tungsten
Dwaipayan Dasgupta 1 , Dimitrios Maroudas 2 , Brian Wirth 1
1 Department of Nuclear Engineering, University of Tennessee, Knoxville, Knoxville, Tennessee, United States, 2 Department of Chemical Engineering, University of Massachusetts Amherst, Amherst, Massachusetts, United States
Show AbstractNuclear fusion is widely recognized as one of the grand challenges for engineering and applied science in the 21st century. The structural and thermomechanical response of the plasma-facing materials (PFMs) that are exposed to the harsh plasma environment of the nuclear reactor and damaged by the high fluxes of helium (He) atoms produced by nuclear fusion is one of the major limitations toward realizing nuclear fusion. Due to its low hydrogen solubility, low sputtering yield, high melting point, and high thermal conductivity, tungsten (W) is considered as a suitable PFM candidate for divertor and first-wall systems, capable of tolerating the extreme reactor conditions. Nevertheless, experiments have shown that helium from linear and tokamak plasma devices is responsible for the formation of a nanostructure with a fuzz-like morphology on the W surface after a few hours of plasma exposure. Fuzz formation can potentially affect the reactor performance leading to an increased nucleation of He bubbles, retention of hydrogen isotopes, and production of high-atomic-number dust. Over the past decade, numerous atomistic simulation studies have aimed at obtaining a fundamental understanding of the dynamics of fuzz formation; however, their results were limited to spatial and temporal scales lower by orders of magnitude than the experimentally relevant ones. Here, we focus on developing an atomistically-informed, continuous-domain model capable of describing the surface morphological evolution of helium-ion-irradiated tungsten and predicting the initial stage of fuzz formation, focusing on W{110} surfaces. Based on this model, a systematic protocol of self-consistent dynamical simulations of the dynamics of the irradiated tungsten surface morphology is conducted to benchmark the simulation results against experimental studies available in the literature and to identify the critical range of conditions for nanotendril formation on the surface, a precursor to fuzz-like surface growth. We examine a broad range of operating conditions, including surface temperatures from 1000 to 2300 K, He ion energies from ~10 eV to ~1 keV, and He fluxes over several orders of magnitude from 1016 to 1022 m-2 s-1. We also present the results of a sensitivity analysis of the key model parameters, such as He concentration and He nanobubble size. Future extensions of our model, driven by comparisons of the model predictions with carefully designed experiments and aimed at establishing stronger links between multiscale models and experimental data and observations, also will be discussed.
4:30 PM - ES08.03.04
Multiscale Irradiation Effects of Tungsten Based Materials
Osman Atwani 1 , Erika Esquivel 1 , Mert Efe 2 , Eda Aydogan 1 , Enrique Martinez 1 , Blas Uberuaga 1 , S. Maloy 1
1 , Los Alamos National laboratory, Los Alamos, New Mexico, United States, 2 Metallurgical and Materials Engineering, Middle East Technical University, Ankara Turkey
Show Abstract
Development of advanced structural materials for nuclear power requires overcoming several material design, synthesis and testing challenges. For fusion, tungsten (W) is considered to be the best candidate as a plasma material interface (PMI). However, W, as a PMI, will be exposed to a high energy neutron flux and the intense plasma which can trigger detrimental changes to the microstructure and the mechanical properties. Addressing these issues via designing irradiation resistant materials (e.g. nc materials and alloys) and investigating their performance is crucial. Here we present multiscale phenomena in irradiated tungsten materials as a step to generate a “figure of merit” correlating materials morphological and mechanical response with irradiation parameters. For small scale studies, low energy helium irradiations and high energy heavy ion irradiation on several tungsten grades (nanocrystalline, ultrafine, coarse grained, and alloys) are performed. Defect densities, sizes and the overall swelling as a function of grain size is presented based on transmission electron microscopy images. The effect of grain boundary density on limiting the irradiation damage is quantified. The grain boundary sink efficiency is discussed based on denuded zone formation, and the denuded zone/sink efficiency concept is revised. At the large scale, nanoindentation and tests before and after irradiation (eg. nanocrystalline tungsten samples with helium bubble-loaded grain boundaries) are discussed. Plasma irradiations on the same samples and the resulting fuzz formation are illustrated and compared among the different grades. The small scale and the large scale studies are then correlated. Conclusions regarding the advantages and possible disadvantages of using pure nanocrystalline tungsten materials are presented.
4:45 PM - ES08.03.05
Thermal Conductivity of Tungsten—Effects of Plasma-Related Structural Defects from Molecular-Dynamics Simulations
Lin Hu 1 , Brian Wirth 2 , Dimitrios Maroudas 1
1 Department of Chemical Engineering, University of Massachusetts Amherst, Amherst, Massachusetts, United States, 2 Department of Nuclear Engineering, University of Tennessee, Knoxville, Knoxville, Tennessee, United States
Show AbstractThe high thermal conductivity of tungsten is one of the main reasons for its consideration as a plasma-facing material (PFM) in nuclear fusion reactors. However, the exposure of PFMs to large fluxes of low-energy helium (He) irradiation from the plasma causes significant nanostructural changes in the PFM’s near-surface region. Structural defects such as He nanobubbles due to plasma exposure are expected to have significant impact on tungsten’s thermal conductivity in the region near its plasma-exposed surface. In this presentation, we report on the effects of plasma-related structural defects on the lattice (or phonon) thermal conductivity of tungsten based on an atomic-scale computational analysis. Using non-equilibrium molecular-dynamics (MD) simulations, we have conducted a systematic study of thermal transport and computed the lattice thermal conductivities of tungsten single crystals containing nanoscale-size pores or voids and He nanobubbles as a function of void/bubble size and gas pressure in the He bubbles. The plasma-related defects in our simulations are representative of He nanobubbles that form in large-spatial-scale MD simulations of near-surface structural evolution of He-implanted tungsten. The calculated lattice thermal conductivities of these defective tungsten single crystals are compared with those of perfect tungsten crystals with heat fluxes along the [100], [110], [111], and [211] crystallographic directions. The presence of nanoscale voids in the crystal causes a significant reduction in its lattice thermal conductivity, which decreases with increasing void size. Filling the voids with He to form He nanobubbles and increasing the bubble pressure leads to further reduction, down to 20%, of perfect tungsten’s lattice thermal conductivity, with weak anisotropy in heat conduction over the range of void size and He nanobubble pressure examined. In addition, we have analyzed the pressure and atomic displacement fields in the crystalline region that surrounds the He nanobubbles to further understand phonon transport and the origin of phonon scattering mechanisms that mediate heat conduction in plasma-exposed tungsten. The analysis shows that the significant reduction of tungsten’s lattice thermal conductivity in the region that contains He nanobubbles is due to phonon scattering from these defects, as well as the deformation of the lattice around the nanobubbles and the formation of lattice imperfections in these regions at higher bubble pressure.
Symposium Organizers
Karl Whittle, University of Liverpool
Felix Brandt, Forschungszentrum Juelich
Philip D Edmondson, Oak Ridge National Laboratory
Blas Uberuaga, Los Alamos National Laboratory
ES08.04: Oxides
Session Chairs
Gianguido Baldinozzi
Blas Uberuaga
Tuesday AM, November 28, 2017
Hynes, Level 2, Room 206
8:30 AM - ES08.04.01
Structural Features in A2O3:BO2 Family of Compounds (A: Sc/Ln and B: Zr/Hf)
Maulik Patel 1 , Karl Whittle 1 , Jeffery Aguiar 2 , Sven Vogel 3 , Gianguido Baldinozzi 4 5 , Kurt Sickafus 6
1 , University of Liverpool, Liverpool United Kingdom, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 3 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 4 , SPMS, Gif-sur-Yvette France, 5 , DMN, SRMA, CEA,, Gif-sur-Yvette France, 6 , University of Tennessee, Knoxville, Knoxville, Tennessee, United States
Show AbstractComplex oxides with structure derived from the fluorite (CaF2-type) like, Perovskites (ABO3), Spinels (AB2O4), Bixbyite (A2O3) and Pyrochlores (A2B2O7), have a variety of applications, including thermal barrier coatings, electrolytes in solid oxide fuels and the accommodation of radioactive nuclides coming from spent nuclear fuel for transmutation or waste disposal. Of these, the structural modifications in the δ-(A4B3O12), γ-(A2B5O13) and β-(A2B7O17) family of compounds (A: Sc/Ln and B: Zr/Hf)) with a rhombohedral structure (R-3) need to be better defined. Thus, the presentation will discuss some unique crystallographic features in these fluorite derivative oxides. The discussion will include inferences drawn from the analysis of selected area electron diffraction, X-ray diffraction and neutron diffraction data to present structural ordering differences between δ-, γ- and β- family of compounds. An attempt will also be made to present a correct description for a β-(phase) and the influence of non-stoichiometry.
8:45 AM - ES08.04.02
Determining the Feasibility of Using Impedance Spectroscopy to Probe Radiation Damaged Surfaces in Ceramics
Richard Veazey 1 , Amy Gandy 1 , Julian Dean 1 , Derek Sinclair 1
1 , The University of Sheffield, Sheffield United Kingdom
Show AbstractFor some time, the affect of structure and composition on a materials response to radiation damage has been of particular interest to the nuclear industry. To study the effects of radiation damage, Transmission Electron Microscopy (TEM) and Glancing Angle X-ray Diffraction (GAXRD) have been used extensively. However, electrical properties of a material may give a better understanding of a materials response to radiation damage. Impedance Spectroscopy is a powerful technique that is commonly used to characterise electroceramics. It measures a samples response over a wide frequency range (Hz – MHz), and allows different electrical regions to be identified. If damage and amorphisation affects the electrical properties of the material, this region will “relax out” at a different frequency to the bulk, allowing for quantifiable analysis of this region. Here, we use a combination of finite element modelling (FEM) and experiments to determine the feasibility of this new approach.
Using an in-house developed FEM package, ElCer1, allows us to simulate this process. We use this to create a model that incorporates a surface layer, with different electrical properties (conductivity and permittivity) to that of the bulk, mimicking the damaged region. Different electrode geometries were used, from Macro-contacts (conventional), measuring the materials overall response, to Micro-contacts, locally probing the surface of the material. The simulations showed that Macro-contacts are capable of measuring the surface layer, if it has significant differences in its electrical properties. For less significant differences, Micro-contacts can obtain promising results. Using the software, current density plots are also obtained, for various contact size and separations, allowing visualisation of the current flow through the material, and thus allow us to better understand how different components contribute to the impedance response.
In addition to the modelling, heavy ion implantation was used to induce radiation damage in polycrystalline and single crystal materials of composition NaBiTi2O6, SrTiO3 and Y2Ti2O7. Samples were subsequently studied by GAXRD and TEM to confirm the damage. We will present results from Impedance Spectroscopy measurements of the damaged samples. These will be compared to the FEM results to obtain a greater understanding of the effects of damage on a materials impedance response.
References:
1. Dean, J. S., Harding, J. H. & Sinclair, D. C. Simulation of Impedance Spectra for a Full Three-Dimensional Ceramic Microstructure Using a Finite Element Model. J. Am. Ceram. Soc. 97, 885–891 (2014).
9:00 AM - *ES08.04.03
Nuclear Materials under Extreme Conditions—Local Defect Structure and Disorder
Maik Lang 1
1 , University of Tennessee, Knoxville, Tennessee, United States
Show AbstractActinide oxides and fluorite-derivative pyrochlores are attractive candidates for various nuclear-related applications owing to their resilient nature under various extreme conditions. We have investigated local defect structures and disorder in these materials induced under intense ion irradiation by the utilization of state-of-the-art instruments at large user facilities. Relativistic heavy ion beams (GSI Helmholtz Center, Darmstadt Germany) were used to induce high energy densities in materials, which drive the local atomic structure far from equilibrium. This triggers complex structural modifications, which require analysis by advanced characterization techniques in order to obtain fundamental understanding of damage mechanisms, defect behavior, phase transformations, and nanoscale structural alterations. Neutron total scattering experiments have been performed on the irradiated materials at the Nanoscale Ordered Materials Diffractometer (NOMAD) beamline at the Spallation Neutron Source at Oak Ridge National Laboratory. Neutrons scatter strongly from low-Z elements, permitting a detailed analysis of both cation and anion defect behavior. Pair distribution function (PDF) analysis elucidates the local defect structure, including changes in site occupation, coordination, and bond distance. For example, using complementary neutron total scattering and calorimetry we have gained new insight into the nature of disorder in Dy2Ti2O7 over a range of length-scales. This experimental effort was supported by DFT calculations confirming the complex disordering mechanism in fluorite-derivative oxide materials. The damage recovery at high temperatures also appears to be more complex than previously thought. We have further examined how local structural modifications in defective materials affect physical properties by revealing a significant increase of ionic conductivity after ion irradiation using broadband dielectric spectroscopy. This experimental approach has also been expanded to actinide compounds in order to probe the defect structure induced by irradiation (ThO2) or by deviation from ideal stoichiometry (UO2+x). PDF analysis from neutron total scattering shows that the defect structure is governed by complex oxygen defect clustering in these compounds.
9:30 AM - ES08.04.04
Nanoparticle Formation in Irradiated and Annealed Ceria
Weilin Jiang 1 , Michele Conroy 1 , Karen Kruska 1 , Nicole Overman 1 , Timothy Droubay 1 , Jonathan Gigax 2 , Lin Shao 2 , Ram Devanathan 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States, 2 , Texas A & M University, College Station, Texas, United States
Show AbstractWe present the results of an integrated experimental and simulation study of the mechanism of metallic nanoparticle formation in an irradiated ceramic. The goal is to understand the formation of precipitates of metallic fission products in nuclear fuel under the combined effects of irradiation and temperature. Of particular interest is the hexagonally close packed epsilon phase that typically consists of five metals, Ru, Mo, Tc, Pd and Rh. Our novel approach to examine nanoparticle formation involves doping metals (Mo, Ru, Rh, Pd, and Re) into a nuclear fuel surrogate, CeO2; ion implantation and furnace annealing to create defects and voids; molecular dynamics simulation of the atomic level defect production process; and characterization of the stages in nanoparticle formation using electron microscopy. Our results show that Pd particles of about 3 nm in size formed near dislocation edges as a result of 90 keV He+ ion irradiation at 673 K. Subsequent thermal annealing at 1073 K led to formation of larger nanoparticles containing Mo and Pd at grain boundaries. Grain boundaries in CeO2 appear to act as both fast diffusion pathways and effective sinks for Mo and Pd to precipitate. Further annealing at 1373 K produced 70 nm sized particles that contained all the five metallic dopants.
9:45 AM - ES08.04.05
Corrosion of Energy System Materials in Supercritical Carbon Dioxide (sCO2)
Benjamin Adam 2 1 , Lucas Teeter 2 , Mark Anderson 3 , Julie Tucker 2 , Bruce Pint 4 , Gordon Holcomb 5 , Changheui Jang 6
2 MIME, Oregon State University, Corvallis, Oregon, United States, 1 MME, Portland State University, Portland, Oregon, United States, 3 Engineering Physics, University of Wisconsin–Madison, Madison, Wisconsin, United States, 4 Corrosion Science and Technology group, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 5 , National Energy Technology Laboratory, Albany, Oregon, United States, 6 Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon Korea (the Republic of)
Show AbstractSupercritical CO2 in the Brayton cycle is among the most promising heat conversion processes for Gen-IV nuclear power plants, Advanced Ultra-supercritical fossil power plants and concentrating solar power plants. Since increasing heat source outlet temperature generally yield an increase in overall plant efficiency, the in-service conditions for sCO2 systems are expected to be between 300 [°C] and up to 750 [°C] and up to 20 [MPa] pressure. These conditions are expected to lead to material loss and degradation through oxidation and carburization phenomena, affecting life and performance of heat exchanger components. The need for reliable data and sound understanding of the corrosion behaviour of materials to be used in this environment is identified as a critical factor on the way to establish safe material requirements. A Round Robin-type study has been initiated, where selected alloys are tested, analysed, and the recorded results then shared between 6 partnering institutions, ensuring unbiased and verifiable datasets. In this study, five promising candidate materials, Gr 91, HR-120, Inconel 625, Inconel 740H and SS316, were exposed to sCO2 and their corrosion behaviour and structure-property relationships were characterised. Exposure was performed in an autoclave with a testing capacity of 1 [L], operating at temperatures of 550 and 700 [°C] and a pressure of 20 [MPa], at times up to 1500 hours in 500 hrs increments.
ES08.05: Metals
Session Chairs
Tuesday PM, November 28, 2017
Hynes, Level 2, Room 206
10:30 AM - *ES08.05.01
Investigating Vanadium Alloys as Potential Future Fusion Materials Using Small Punch Testing
Bradley Wynne 1 2 , Mark Richardson 2 , Mike Gorley 2 , Elizabeth Surrey 2
1 Materials Science and Engineering, The University of Sheffield, Sheffield United Kingdom, 2 , United Kingdom Atomic Energy Authority, Abingdon, Oxfordshire, United Kingdom
Show AbstractA significant impediment to the design and development of a commercially viable tokamak-based fusion reactor is the choice of materials currently available that are low activation and can withstand the ultra-extreme environments experienced in the first wall/blanket. The blanket has multiple roles and requirements: a) convert the kinetic energy of neutrons into heat, b) extract the heat to generate power, c) produce tritium to continue the fusion reaction, and d) maintain structural integrity for at least 5 years. Currently, there are no materials capable of satisfying all of these requirements and thus there is a significant technological barrier for incorporation of fusion into our energy supply systems. Vanadium alloys are a promising future fusion material; V-Cr-Ti compositions in particular have been shown to have high creep strength for long-term operation up to temperatures of 700°C, with good levels of strength and irradiation tolerance and a relatively low ductility loss because of radiation-induced defects. Here, we present on-going work on assessing the non-irradiated mechanical properties of a variety of vanadium alloys from commercially pure through to extended V-Cr-Ti compositions for the production of oxide dispersion strengthened systems. These alloys are manufactured using a number of techniques including traditional wrought methods and advanced powder based methods such as spark plasma sintering and additive layer manufacturing. All assessments are being undertaken using the small punch technique. Small punch testing is one of a number of small specimen testing techniques currently being assessed across the sector for their capability to extract reliable “full-scale” mechanical properties with high enough confidence levels that engineers could use them for design purposes. This is with the goal of extracting reliable mechanical properties from test samples that have undergone irradiation campaigns. To this end, the study of the suite of vanadium alloys and manufacturing techniques has highlighted a number of issues including the differentiation between assessment of a ductile and brittle materials, the sensitivity to manufacturing defects and variability, and the impact of temperature on mechanical properties. Finally, the presentation will end with a case study on how small scale testing can highlight a number of issues with the quality of material being tested, leading to a rapid refinement of process before a larger and more expensive material component would be manufactured and tested.
11:00 AM - ES08.05.02
Structural Integrity Characterisation of Nuclear Materials via Nano Additive Manufacturing
Mahmoud Mostafavi 1 , Mike Taverne 1
1 , University of Bristol, Bristol United Kingdom
Show AbstractThe structural integrity of nuclear fission and fusion power plant components is the focus for this research. The state of the art is using micro scale specimens milled with a focussed ion beam (FIB). Because of their very low volume such specimens can be lab tested, even when irradiated to low or medium level of activity. This offers a possibility of testing multiple specimens to investigate stochastic effects, e.g. effects of irradiation on the shift of the ductile to brittle transition. However, FIB milled specimens suffer from gallium contamination, to the degree that the validity of fracture data obtained on such specimens is questionable. We propose to use nano-additive manufacturing as an alternative to FIB for making micro scale fracture specimens. A combination of two-photon polymerisation, thermal evaporation and electron beam induced deposition is used to manufacture micro-scale fracture specimens with fully reproducible initial crack topologies. Such specimens will be free from gallium contamination and thus better suited for the study of irradiation effects on structural integrity. In addition, micro-scale geometries with reproducible sharp cracks can be embedded in large components to investigate effects of localised sharp microcracks on structural integrity. In this study we will focus on nano-additive manufacturing of micro scale specimens of tignsgten, fracturing them in un-irradiated form, extracting statistical distributions of the fracture toughness and strain energy release rate and on using CAFÉ multi-scale modelling with measured microscale properties to predict macro scale structural integrity.
11:15 AM - ES08.05.03
Microstructure and Mechanical Properties of Laser Beam Surface Melted Inconel 718
Sumit Sharma 1 , Ashish Nath 1 , Koushik Biswas 1 , J. Majumdar 1
1 , IIT Kharagpur, Kharagpur India
Show AbstractIn the present study, Inconel 718 has been subjected to laser surface melting using a 2 kW continuous wave (CW) Yb:fibre laser under a varied process parameters. There is a significant refinement of microstructure, the morphology of which varied with process parameters. The effect of laser parameters on the grain size and its distribution, morphology and grain orientation have been studied in detail. The surface modified layer consists of γ-Ni primary dendrites and fine precipitates of γ’ and γ’’ in the interdendritic regions. The mass fraction of individual phase as determined by X-ray diffraction study has been found to vary with process parameters. There is an improvement in hardness value. The kinetics and mechanism of wear have been evaluated in detail.
Keywords: Laser Beam Melting; Inconel 718 super alloys; XRD, microstructures, Wear
11:30 AM - ES08.05.04
Microstructure Evolution of Zirconium Alloy in the Process of Deuterium Absorption
Cheng Zhang 1 , Xiping Song 1
1 State Key Laboratory for Advanced Metals and Materials, University of Science and Technology, Beijing China
Show AbstractAs one of the most impressive materials, zirconium alloys have received more and more attentions currently in the nuclear industry, not only because of their combination of low neutron absorption cross-section, high corrosion resistance, and excellent mechanical properties, which have been used as nuclear fuel claddings in the light water and heavy water reactors, but also because of their excellent deuterium absorption properties, which will be used as a potential deuterium storage material in the fusion reactor, as it is known that the deuterium is the basic fuel in the fusion reaction. However, cracking has occurred in zirconium alloys after deuterium absorption, which heavily damages its deuterium storage properties.
In our previous study, it was found that the cracking behavior is highly related to the deuteriding microstructure. But the information about the microstructure evolution of zirconium alloy during deuterium absorption is very limited, which in turn limits the understanding of cracking of zirconium alloys as deuterium storage materials.In this paper, the effects of deuteriding pressure, time and temperature on the deuteride distribution, phase types, and deuterium absorption kinetic in zirconium alloy were investigated. The results presented that the deuterium absorption pressure played an important role on the formation of deuterides. With the increase of pressure from 1bar to 3bar, the deuterides morphology evolved from intragranular needle-like δ deuterides to intergranular mass δ deuterides, and eventually formed a deuteride network around grain boundary. At a higher pressure of 3bar, the surface deuteride layer formed first, and then the inner deuteride network was gradually developed. The kinetic of deuterium absorption at 1173K followed the chemical reaction mechanism at lower deuteriding pressures, and nucleation and growth mechanism at higher deuteriding pressures. With the increasing of deuteriding temperatures from 973K to 1173K, the deuteride content decreased. The TEM results showed that the ε deuterides nucleated and grew at the interface of δ deuterides, and small bands with different crystal orientation were found within the ε deuterides. An orientation relationship of [011]δ//[011]ε, {111}δ//{111}ε between δ and ε deuterides was also determined by TEM analysis. The γ deuterides were found with twins and tweed structure. Based on the results above, a schematic diagram of microstructure evolution with deuterium absorption in zirconium alloy was proposed.
11:45 AM - ES08.05.05
The Role of Grain Boundaries and Second Phase Particles in Predicting Nucleation Sites for Hydride Precipitation in Zirconium Alloys
Said El Chamaa 1 , Catrin Davies 1 , Mark Wenman 1
1 , Imperial College London, London United Kingdom
Show AbstractZirconium alloys are used in water-cooled nuclear reactors and are susceptible to a time dependent degradation mechanism know as delayed hydride cracking (DHC). Corrosion of zirconium alloys in hot aqueous environments generates hydrogen that subsequently diffuses through the metallic structure towards stress concentrating features such as notches or cracks. Hydrogen resides in solid solution below the solubility limit but at low temperatures, hydrogen precipitates as brittle hydrides.
This work focuses upon microstructural effects on hydride precipitation in zirconium alloys. The cathodic hydrogen charging technique was employed to introduce 0.007 wt% hydrogen, followed by a homogenisation heat-treatment and post-hoc hydride and SPP analysis for two alloys, namely zircaloy-4 (Zr-4) and near pure zirconium (zircadyne, Zr-702). Microstructures with different textures, grain size and second phase particle (SPP) distribution were hydrided and analysed in-order to study the potency of hydride precipitation at different microstructural nucleation sites. SPP number density was quantified to be 1.87 × 10-2 and 3.44 × 10-2 SPP/µm2 for Zr-4 and Zr-702 respectively. The number density of macro-hydride precipitation was also quantified to be 375 and 338 hydrides/mm2 for Zr-4 small (15 µm) and large grained samples (300 µm) respectively. This was slightly different for zircadyne where the hydride number density was calculated to be 383 and 303 hydrides/mm2 for small (20 µm) and large grained (300 µm) samples respectively. Therefore, around 66 % of hydrides precipitate at grain boundaries in small grained Zr-4 compared to 89% for small grained zircadyne. Considering the difference in grain size between these alloys was small this is attributed to the difference in SPP content, which is 1.8x higher in Zr-4 leading to more intra-granular precipitation and showing that SPPs do actively change hydride nucleation sites. As grain size was increased from 15-20 to 300 µm, for both alloys, the percentage of grain boundary precipitation reduces in both Zr-4 (from 66 to 11%) and in zircadyne (from 89 to 26%) highlighting the potency of the grain boundaries as hydride nucleation sites. However, again in Zr-4, with the higher SPP density, grain boundary precipitation was much less at 11% compared to 26% in zircadyne again because of the role of SPPs acting as intragranular sites in the Zr-4 alloy.
These preliminary results, along with hydride area fraction and size distribution, will be utilised to parameterise models capable of predicting hydride evolution with respect to microstructure currently under development. This model will account for hydride nucleation sites as a function of polycrystallinity and anisotropy as well as the hierarchy of microstructural features that act as heterogeneous nucleation sites such as grain boundaries, SPPs and dislocation networks.
ES08.06: Radiation Damage in Metals I
Session Chairs
Xianming Bai
Laurent Capolungo
Tuesday PM, November 28, 2017
Hynes, Level 2, Room 206
1:30 PM - *ES08.06.01
Rate Theory Model for Irradiation Evolution of Nanoclusters
Matthew Swenson 3 2 , Janelle Wharry 1
3 Department of Mechanical Engineering, University of Idaho, Moscow, Idaho, United States, 2 Micron School of Materials Science and Engineering, Boise State University, Boise, Idaho, United States, 1 School of Nuclear Engineering, Purdue University, West Lafayette, Indiana, United States
Show AbstractThe objective of this talk is to demonstrate a versatile, experimentally validated predictive model for irradiation evolution of nanoclusters in advanced nuclear structural alloys. Nanoclusters, defined as finely dispersed precipitates having diameters ≤20 nm, play a critical role in the irradiation tolerance and mechanical integrity of structural materials for advanced nuclear fission reactors. Examples include fine G-phase, Cu-rich, and Cr-rich α’ phase precipitates recently reported in ferritic/martensitic (F/M) alloys, all of which cause undesirable embrittlement; and Y-Ti oxide nanoclusters included in oxide dispersion strengthened (ODS) alloys to enhance irradiation tolerance and provide high temperature strength. Numerous reports in the archival literature have studied the irradiation evolution of nanoclusters in F/M and ODS alloys. But these studies have used a wide variety of experimental parameters, including initial alloy/composition, irradiation temperature, dose, dose rate, particle type, and analysis technique. So while these studies have presented a large volume of data on the effects of irradiation on nanoclusters, their lack of systematic design has precluded any notable trends from surfacing.
In this work, we advance a predictive model that can accurately forecast the irradiation evolution of all aforementioned nanocluster types over a wide variety of experimental conditions, enabling us to explain the diverse experimental reports from the archival literature. The model is based upon the Nelson, Hudson, and Mazey formulation (Journal of Nuclear Materials, 1972), in which the change in nanocluster radius with time (i.e. fluence) is a function of recoil dissolution, disordering dissolution, and growth attributed to irradiation-enhanced diffusion. Recent developments in high-spatial resolution atom probe tomography (APT) enable us to precisely measure necessary model input parameters.
Model results are validated against experiments from a wide variety of materials (F/M alloys HT9 and HCM12A, and two varieties of ODS), irradiation temperatures 300-600°C, and irradiating particle types (neutron, 2 MeV proton, and 2-9 MeV Fe ion) that deliver correspondingly variable dose rates. We find nanoclusters converge on a steady-state size at a fluence 3-10 displacements per atom (dpa); the steady-state size and fluence varies with alloy composition and irradiation parameters. We also find that higher dose rate irradiations require a negative temperature shift in order to produce nanocluster morphologies consistent with lower dose rate irradiations. This work is the first versatile model that can meaningfully predict irradiation evolution of such a wide variety of nanocluster compositions, alloy types, and irradiation conditions. As such, this work represents a tremendous advancement in our mechanistic understanding of irradiation effects on nanoclusters.
2:00 PM - ES08.06.02
Investigation of Helium Bubble Formation within Palladium
Kathryn Yates 1 , Na Ni 1 , David Wheeler 2 , William Lee 1 , John Knowles 2 , Alexander Morris 3 , Simon Pimblott 3
1 , Imperial College London, London United Kingdom, 2 , AWE, Reading United Kingdom, 3 , University of Manchester, Manchester United Kingdom
Show AbstractUnderstanding the formation and mobility of helium bubbles within face-centred cubic (fcc) metals is required to appreciate the complex changes that take place in plutonium (Pu) alloys as they age. The self-irradiation of Pu results in lattice damage, compositional changes, and helium accumulation at a rate of 41 appm/year due to a-decay [1,2]. These phenomena are thought to affect the mechanical properties of Pu [3] and may also influence the corrosion behaviour.
Palladium is a stable and readily available metal with an fcc structure and by subjecting samples of palladium to controlled ion implantations, aspects of self-irradiation induced ageing can be observed. This study investigates the formation of helium bubbles within He implanted single crystals of palladium by using electron microscopy and ion beam analysis to determine bubble locations, sizes, densities and distributions. Helium implantations to a dose of 3.5 x 10 17 ions/cm2 were performed using the Dalton Cumbrian Facility (DCF) ion accelerator [4]. The damage profiles within the irradiated crystals under these conditions were predicted using the software package Stopping and Range of Ions in Matter (SRIM) [5]. Preliminary results have revealed that the irradiation generated helium bubbles over a range of different sizes (1.35 – 4.05 nm) with an average diameter of 1.70 nm and the ion implantation produced unusual effects upon the microstructure.
2:15 PM - ES08.06.03
Helium Nanochannel Formation in Metal Nano-Layers
Di Chen 1 , Nan Li 3 , Dina Yuryev 2 , John K. Baldwin 3 , Michael Demkowicz 4 , Yongqiang Wang 1
1 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 3 Center for Integrated Nanotechnologies, Materials Physics and Applications Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Department of Materials Science and Engineering, Massachusetts Institute of Technology, Boston, Massachusetts, United States, 4 Department of Materials Science and Engineering, Texas A&M University, College Station, Texas, United States
Show AbstractWe explore He precipitation within a nano-layer of Cu sandwiched between two V layers. Using transmission electron microscopy (TEM), we show that confinement within the nano-scale layer causes He precipitates to depart from their classical growth trajectories: rather than expanding continuously while remaining equiaxed, He precipitates spontaneously coalescence into elongated, He-filled channels. To explore potential applications of He nanochannels for plasma facing materials in fusion energy applications, we deposit W/Cu nano-composites where the Cu component intersects the sample surface, providing a potential path for He outgassing. We explore the underlying physics of He precipitation and nanochannel formation using phase field modeling. Our findings suggest a novel method of manipulating He aggregation, thus providing a new concept for He damage-resistant materials for nuclear energy.
2:30 PM - ES08.06.04
Effect of Residual Gas on Damage Accumulation in Reactor Materials
Stanislav Golubov 1 , Alexander Barashev 2
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 2 , University of Tennessee, Knoxville, Tennessee, United States
Show AbstractA cluster dynamics model of microstructure evolution in reactor materials is used to elucidate the materials and irradiation parameters that control cavity nucleation under conditions of the light water and fast breeder reactors. The model takes into consideration the interaction between radiation vacancies and helium atoms produced in nuclear transmutation reactions and residual gas atoms present in the material prior to irradiation. This interaction enhances stability of small vacancy aggregates, thus increasing void nucleation rate. The model accounts for realistic production of the primary damage in cascades of atomic displacements, in particular, formation of clusters of self-interstitial atoms and their one-dimensional migration. The calculation results are compared with available experimental data for austenitic stainless steels. A significant effect of residual gas atoms, as compared to those produced by irradiation, is demonstrated. It is argued that most experimental data may only be rationalized if the effect of residual gas on void nucleation is taken into consideration.
2:45 PM - ES08.06.05
Helium Bubble Coarsening and Helium Release in Hastelloy N Alloy
Jie Gao 1 , Hefei Huang 1 , Yan Li 1
1 , Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai China
Show AbstractHastelloy N alloy was implanted with 30 keV, 5 × 1016 ions/cm2 helium ions at room temperature, and subsequent annealed at 600 C for 1 hour and further annealed at 850 C for 5 hours in vacuum. The depth profiles of helium concentration and helium bubbles in helium-implanted Hastelloy N alloy were investigated using respectively elastic recoil detection analysis (ERDA) and transmission electron microscope (TEM). After helium ion irradiation at room temperature and subsequent annealing at 600 C (1 h), the TEM micrograph indicates the presence of helium bubbles with size of 2 nm in the depth range of 0–300 nm. As for the sample further annealed at 850 C (5 h), on one hand, the “Ripening Zone” (0-64 nm) and ”Coalescence Zone” (64-300 nm) with huge differences in size and separation of helium bubbles, caused by different coarsening rates, are observed. In addition, the diffusion of helium and molybdenum elements to surface occurred, and it was also observed that bubbles in molybdenum-enriched region were much larger in size than those in deeper region. Moreover, it is worth noting that plenty of nano-holes can be observed on the surface of helium-implanted sample after high temperature annealing by scanning electron microscope (SEM). This observation provides the evidence for the occurrence of helium release, which can be also inferred from the results of ERDA and TEM analysis.
3:30 PM - ES08.06.06
In Situ TEM Study of Microstructure Evolution in Concentrated Solid Solution Alloys under Irradiation
Shi Shi 1 , Shuai Wang 1 , Rigen Mo 1 , Ke Jin 2 , Hongbin Bei 2 , Kazuhiro Yasuda 3 , Syo Matsumura 3 , Kenji Higashida 3 , Ian Robertson 1
1 , University of Wisconsin-Madison, Madison, Wisconsin, United States, 2 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 3 , Kyushu University, Fukuoka Japan
Show AbstractIn comparison to conventional alloys, high entropy alloys potentially provide a fundamentally new way to enhance the irradiation damage tolerance by means of increasing compositional complexity. In this talk, we will indicate how the alloy complexity increases tolerance to irradiation damage. The relationship is not simply dependent on the number of elements but rather is influenced by the specific element introduced. Specifically, the defect evolution in a series of Ni-containing concentrated solid solution alloys was determined by conducting electron (1250 kV, 600 kV) and ion (1 MeV Kr++) irradiations in a transmission electron microscope. Defects produced by ion irradiation include perfect and Frank dislocation loops, and stacking fault tetrahedra. The ratio of perfect to Frank loops decreased as does the defect density in the order of Ni, NiCoCr, NiCoFeCr. Under electron irradiation, only Frank loops were found in Ni, Frank and perfect interstitial loops were found in NiFe, NiCoFeCr, and large faulted loops were found in NiCoFeCrMn, NiCoFeCrPd. The loop growth behavior was influenced by the alloy composition, and the growth rate was fastest in NiFe, and slowest in NiCoCr. In addition, the five-element alloy showed evidence for elemental segregation to the loops. These results will be explained in terms of modifications of the energy dissipation processes, defect migration and formation energy, which are influenced by the compositional complexity of the alloys.
3:45 PM - ES08.06.07
Radiation Resistance of Thin Films of FeCrNiMn, a High-Entropy Alloy
Stephen Donnelly 1 , Matheus Tunes 1 , Vladimir Vishnyakov 1
1 , University of Huddersfield, Huddersfield United Kingdom
Show AbstractWith the development of sustainable and improved nuclear power plants (so-called Generation IV) that can operate for several decades without refuelling and with low radioactive waste production, a major challenge is the design of so-called accident-tolerant fuel systems that can sustain structural integrity over long-term exposure to both particle irradiation and corrosive environments.
High-entropy alloys (HEAs) have emerged as potential candidates for structural components of nuclear reactors. HEAs are a class of multicomponent alloy formed with two or more elements in an equiatomic composition. The high-entropy of mixing for these elements is responsible for lowering the Gibbs free energy and for forming a solid solution with enhanced phase stability. HEAs have potential applications in nuclear technology as this enhanced phase stability may lead to greater radiation tolerance than comparable “low-entropy” systems; however, research needs to be done to identify HEA systems best suited to specific nuclear environments.
We have prepared 1.5 µm thick layers of equiatomic FeCrNiMn on a Zircaloy-4 substrate, by sputter-deposition from elemental targets, with the aim of conducting tests on the mechanical properties and chemical stability of this system following irradiation to high levels of radiation damage. For comparison purposes, we are also carrying out similar experiments with AISI 348 austenitic stainless steel which contains the same elements but in a non-equiatomic composition.
We report on a preliminary set of experiments aimed at elucidating the defect morphology of the HEA film under irradiation to high damage levels and gas injection. Cross-sectional FIB lift-out specimens, approximately 50 nm in thickness, were prepared so as to include the Zircaloy-substrate. These were then irradiated with 30 keV Xe ions at fluences up to 2 x 1017 ions/cm2 (190 dpa*) at 673K at the MIAMI Facility (Microscopes and Ion Accelerators for Materials Investigations) at the University of Huddersfield enabling a direct comparison of defect development in the Zircaloy and in the HEA to be made. In addition, similar experiments were carried out on the non-HEA 348 stainless steel.
Xe bubbles were observed to form within the HEA but only at a fluence of 6.6 x 1016 ions/cm2 – a fluence 5 times higher than that at which bubbles were observed to form in the steel and the Zircaloy. In addition, the bubbles remained small (mean diameter 1.6.nm) and were not observed to grow under continued irradiation, unlike in the Zircaloy and the steel where bubbles continued to grow to much larger sizes. No other signs of radiation damage were seen in the HEA, even at high dpa levels in contrast to the Zircaloy and steel in which so-called “black-spot” damage was clearly observed.
We will present the experimental data together with a tentative explanation for the resistance to radiation damage and specifically to the growth of Xe bubbles.
*displacements per atom
4:00 PM - ES08.06.08
Investigation of Chemical Effects on Defect Evolution in Single-Phase Concentrated Solid Solution Alloys
Xing Wang 1 , Brian Sneed 1 , Ke Jin 1 , Hongbin Bei 1 , Jonathan Poplawsky 1 , William Weber 1 2 , Yanwen Zhang 1 2 , Karren More 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 2 , University of Tennessee, Knoxville, Oak Ridge, Tennessee, United States
Show AbstractA novel class of materials, single-phase concentrated solid-solution alloys (SP-CSAs), has attracted increasing interest recently as a result of their outstanding properties, such as high strength and ductility, exceptional radiation resistance, and corrosion resistance at high temperature. Unlike conventional alloys, which are based on one principle element, SP-CSAs are composed of two to five principle elements in equal or near-equal molar ratios, and these elements form random solutions in simple lattice structures like fcc. The random arrangement of multiple elemental species in an ordered structure creates extreme chemical complexity in the materials as each atom experiences unique lattice distortions and chemical environment. It is expected that the chemical complexity can have significant effects on the generation, migration, and evolution of defects, which can further change the material properties under extreme conditions such as irradiation. In this study, the chemical effects on the defect evolution in Ni-based SP-CSAs are investigated by combining irradiation of 200 keV He ions and post-analysis using scanning transmission electron microscopy and atom probe tomography. The growth rate of He bubbles is decreased with increasing Fe concentration in SP-CSAs, but not changed in SP-CSAs containing Cr. Chemical segregation analysis indicates that the fast diffusion of vacancies via Cr lattice sites may lead to the accelerated bubble growth. Our results suggest that chemical complexity can have competing effects on the material radiation-resistance by modifying the defect production and migration.
4:15 PM - ES08.06.09
Analyzing the Effect of Milling Intensity on the Nano-Precipitate Evolution in ODS FeCrAl Alloys Using Atom Probe Tomography
Caleb Massey 1 , Philip D Edmondson 2 , Sebastien Dryepondt 2 , Kurt Terrani 2 , Steven Zinkle 1 2
1 , University of Tennessee, Knoxville, Tennessee, United States, 2 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractDuring an accident scenario, the extreme mechanical and thermal stresses coupled with the oxidative steam environment makes finding suitable accident tolerant fuel cladding materials for current nuclear reactors a challenging endeavor. To meet this challenge, oxide dispersion strengthened (ODS) FeCrAl alloys are currently being investigated as a potential materials solution due to their combination of high temperature oxidation resistance and strength, along with anticipated good radiation resistance. The ability of these alloys to mitigate radiation damage and to maintain high temperature tensile strength is dependent on the distribution of nucleated nano-precipitates within the metal matrix, which is a strong function of the processing methodology utilized for alloy fabrication. Recent consolidated heats of ODS FeCrAl alloys mechanically alloyed in two different high energy attritor mills followed by identical extrusion conditions have shown differences in their grain size distributions and their resultant mechanical properties. It was thus of interest to determine the extent to which the chosen ball mill processing conditions and the resulting mechanical alloying intensity affected the number density of nano-precipitates in a consolidated ODS FeCrAl alloy. Two ODS ferritic FeCrAl specimens with nominal composition (wt.%) Fe-10Cr-6.1Al-0.3Zr+0.3Y2O3 (106ZY) were consolidated via hot extrusion at 1000C following high energy ball milling. 106ZY alloy ZY10C1 was mechanically alloyed in a smaller Zoz Simoloyer CM01 attritor while comparative alloy ZY10C8 was produced using a larger Zoz Simoloyer CM08 model. To account for the difference in the size of the attritor mills, and thus the equivalent energy deposition between the two units, the rotational speeds used during mechanical alloying were varied while keeping the total milling time constant at 40h. The CM01 was operated at 400rpm/900rpm while the larger CM08 was operated at 350rpm/600rpm. Two milling speeds were used for each unit to minimize the formation of inhomogenous mechanical mixing zones during the mechanical alloying process. Miniature SS-3 tensile specimens were machined from the as-extruded material for high temperature tensile property evaluations at temperatures up to 800C. Atom probe tomography (APT) specimens were prepared using the standard FIB lift-out technique using a Quanta 3D FIB-SEM. A Cameca LEAP 4000X Si and a LEAP 4000X HR, both in laser mode, were used to analyze the ZY10C1 and ZY10C8 specimens, respectively. Using insights from electron microscopy coupled with the nano-precipitate characteristics provided by this APT analysis, microstructure-property-processing relationships are summarized and recommendations are provided in the context of an optimized processing route for ODS FeCrAl alloys as an Accident Tolerant Fuel Cladding. This research was funded by the U.S. Department of Energy’s Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.
4:30 PM - ES08.06.10
Molecular Dynamics Simulations of Concentration-Dependent Defect Production and Diffusion in Fe-Cr Alloys
Yaxuan Zhang 1 , Daniel Schwen 2 , Xian-Ming Bai 1
1 Materials Science and Engineering, Virginia Polytechnic Institute and State University, Blacksburg, Virginia, United States, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractFe-Cr based alloys are promising structural and cladding materials for next-generation fission and fusion reactors. The cascade-induced defect production and the subsequent defect diffusion play important roles in the long-term microstructural evolution in these alloys. In this work molecular dynamics simulations are conducted to study these processes in pure Fe and Fe-Cr alloys of different Cr concentration, using a concentration-dependent interatomic potential which correctly captures the change of sign in the heat of mixing. It is found that although the Cr concentration in alloys does not affect the total number of produced Frenkel pairs, the fraction of Cr interstitials is much higher than the concentration of Cr in the alloys and the fraction is sensitive to the Cr concentration. Most of Cr interstitials form Fe-Cr mixed <110> dumbbells. It is also found that Cr solute slows down the Fe interstitial diffusion as the Cr concentration increases but it has a weaker effect on vacancy diffusion. Finally, the defect formation energies, binding energies and heat of mixing are calculated to interpret these simulation results.
4:45 PM - ES08.06.11
Carbide Precipitation Reactions under the Influence of Ar Bubbles in Ion Irradiated Solubilized AISI 316L Alloys
Ítalo Oyarzabal 1 , Mariana Timm 1 , Willian Pasini 2 , Franciele Oliveira 1 , Francine Tatsch 1 , Lívio Amaral 1 , Clarice Kunioshi 3 , Paulo Fichtner 2
1 Institute of Physics, Federal University of Rio Grande do Sul (UFRGS), Porto Alegre Brazil, 2 Metallurgy Department, Federal University of Rio Grande do Sul, Porto Alegre Brazil, 3 , Centro de Pesquisa da Marinha em São Paulo, São Paulo Brazil
Show AbstractIon irradiation is widely used to get more insight into the microstructure modifications correlated to degradation effects observed in materials exposed to a nuclear environment. A deep understanding of the ion irradiation effects is challenging because it strongly depends on the complexity of the material composition, on its initial microstructure configuration as well as on the irradiation conditions (e.g. ion mass, energy, flux and target temperature). In this contribution, we report on heavy ion irradiation effects in a solubilized AISI 316L alloy (regarded as a model case material for a complex alloy system) to investigate the growth of argon bubbles and their influence on the development of irradiation induced precipitate reactions.
Mechanically polished 200 μm thick AISI 316L stainless steel foils were thermally treated at 1100°C for 2 hours in high vacuum to relax the stress from the cold work and solubilize all carbon content present in the matrix. A set of these samples was implanted with Ar ions to produce a 0,25 at. % concentration-depth plateau extending from the near surface to a depth of approximately 250 nm, and then annealed at 550°C during 2 hours to form Ar-vacancy clusters small bubbles. Distinct sets of samples (including control ones without Ar) were then irradiated at temperatures from 450 to 550 °C with Au ions accelerated at 5 MeV or with Ag ions accelerated at 2,5 MeV up to fluences about 20 and 40 dpa at the depth region containing the Ar plateau. These samples were investigated by transmission electron microscopy using plan-view specimens prepared by ion milling. The results obtained demonstrate the formation of M23C6 and MC precipitates only in the samples containing inert gas bubbles, whereas the concomitant irradiation of control samples (i.e. without Ar) results on the formation of large cavities and extended defects. The evolution of the precipitate system is discussed considering the diffusion controlled supply of C atoms from the matrix and the irradiation induced vacancy supply. The total volume of the cavity system from the control samples and the total volume of the bubble system from the samples implanted with Ar are evaluated. The volume comparison suggests that the Ar bubbles prevents a fast agglomeration and annihilation of the irradiation induced vacancies, thus favoring the nucleation of precipitate phases presenting a positive volume misfit.
ES08.07: Poster Session
Session Chairs
Felix Brandt
Philip D Edmondson
Blas Uberuaga
Karl Whittle
Wednesday AM, November 29, 2017
Hynes, Level 1, Hall B
8:00 PM - ES08.07.01
Thermal Heat Transfer behavior of UAlx-Al Composites Dependent on Microstructural Morphology
Eui-Hyun Kong 1 , Soo-Hyun Joo 2 , Jae-Yong Oh 1 , Young-Wook Tahk 1 , Hyun-Jung Kim 1 , Hyun-Woo Jeon 1 , Jeong-Sik Yim 1
1 Nuclear Fuel Design Group, Korea Atomic Energy Research Institute (KAERI), Daejeon Korea (the Republic of), 2 Institute for Materials Research, Tohoku University, Sendai Japan
Show AbstractTc-99m for medical application can be obtained through radiative decay of Mo-99. Mo-99 had been produced through nuclear fission of U-235 in U-Al alloy with highly enriched uranium (HEU). According to the nuclear non-proliferation policy, low enriched uranium (LEU) fules have been recently researched by many scientists and engineers. Early LEU fuels were composed of crushed UAl3 and UAl4 particles in Al matrix. Usually crushed particles show nonuniform dispersion in matrix and poor reproducibility. Korea Atomic Energy Research Institute (KAERI) has developed atomized UAlx particles for the production of Mo-99. Atomized particles were uniformly dispersed in Al and remarkably reproduced. Unlike normal materials, it is difficult to measure thermal properites of nuclear fuel materials due to high radioactivity of them. Thus, we predicted thermal heat transfer behavior UAlx-Al composites dependent on microstructural morphology with finite element method (FEM). Based on heat transfer analysis results, fission molybdenum target meat using atomized particles for the Kijang research reactor (KJRR) showed lower maximum centerline temperature. We expect the atomized particle provides the consevative design of nuclear fuels from a safety point of view.
8:00 PM - ES08.07.02
Thermal Jamming of Ions in the Superionic State of UO2
Dillon Sanders 1 , Jacob Eapen 1
1 , North Carolina State University, Raleigh, North Carolina, United States
Show AbstractSolid state superionic materials exhibit a significant increase in ionic conductivity within a certain temperature range, referred to as the superionic regime. This increase is caused by mobile ions in the system that diffuse rapidly within a superlattice of relatively stationary ions. Recent studies of superionic conductors have uncovered the presence of dynamical heterogeneity among the mobile ions, a phenomenon in which spatially-distinct regions of mobile ions exhibit different dynamical behavior than those in nearby regions, as well as quasi one-dimensional string-like hopping among the mobile ions. An analogy may be made to glass-forming liquids, in which string-like cooperative motion and dynamical heterogeneity accompany the increasing frustration of glass-forming liquids under super-cooling.
In this work, using atomistic simulations in conjunction with the Two-Phase Thermodynamic (2PT) Method that partitions a system into a hard sphere (gas) component and a harmonic oscillator (solid) component, we quantify the effective “thermal” packing fraction of UO2 ions in the superionic state. UO2 has a fluorite structure in which the uranium ions are arranged in a face-centered cubic lattice with the oxygen ions arranged in a simple cubic lattice. Recent investigations have shown that UO2 becomes superionic at temperatures above 2000 K; the anion disorder continues to increase which culminates in the Bredig or λ transition at ~2600 K. Our analysis with the 2PT methodology shows that at temperatures near 2000 K, the oxygen ions are in a jammed state with a thermal packing fraction close to the value for a random close-packed configuration (~0.64). With increasing temperature, the thermal packing fraction for oxygen ions reduces – signifying increasing disorder – and attains a value corresponding to a cubic lattice (~0.52), which is consistent with the sub-lattice structure of the oxygen ions, at temperatures above 2600 K. We also observe that the thermal packing fraction of the uranium ions maintains a value close to that for a face-centered cubic lattice (~0.74) throughout the superionic regime. Our work provides an alternate direction for investigating the high temperature properties of UO2 and other actinide fuels. We plan to extend our analysis to probe the effects of fission products on the jamming characteristics of ions in fluorite actinide fuels (such as UO2, ThO2 and PuO2) at superionic temperatures.
8:00 PM - ES08.07.03
Synthesis and Characteristics of Nano-Polycrystalline Synroc Ceramic Forms by a Solution Combustion Synthesis
ChoongHwan Jung 1 , Young-Min Han 1
1 , Korea Atomic Energy Research Institute, Daejeon Korea (the Republic of)
Show AbstractSynroc (Synthetic Rock) consists of four main titanate phases : peroveskite (CaTiO3), zirconolite (CaZrTi2O7), hollandite (BaAl2Ti6O16) and rutile (TiO2). Nano-polycrystalline synroc powders were made by a synthesis combustion process. The combustion process, an externally initiated reaction is self-sustained owing to the exothermic reaction. A significant volume of gas is evolved during the combustion reaction and leads to loosely agglomerated powders. This exothermic reaction provides necessary heat to further carry the reaction in forward direction to produce nanocrystalline powders as the final product. Glycine is used as a fuel, being oxidized by nitrate ions. It is inexpensive, has high energy efficiency, fast heating rates, short reaction times and high compositional homogeneity. In this study, combustion synthesis of nano-sized Synroc powder is introduced. The fabrication of Synroc powder result of observation XRD were prepared for polycrystalline structures. The characterization of the synthesized powders is conducted by using XRD, SEM/EDS and TEM. And the short and long-term leaching tests of Synroc constituents under the groundwater conditions at 80~120C were investigated, and the leaching characteristics of Synroc prepared by SCS process were compared with that of borosilicate glass and Synroc prepared by traditional oxide mixing method.
8:00 PM - ES08.07.04
18F Labeled Magnetic Nanoparticles for Bimodal (MRI/PET) Cellular Imaging
Markus Schütz 1 , Isabel Gessner 1 , Stefan Schonauer 1 , Bernd Neumaier 2 , Sanjay Mathur 1
1 Institute of Inorganic Chemistry, University of Cologne, Cologne, North Rhine-Westphalia, Germany, 2 Institute of Radiochemistry and Experimental Molecular Imaging, University Clinic Cologne, Cologne, North Rhine-Westphalia, Germany
Show AbstractThe design and the rapid development of molecular imaging technology have shown the possibility of preparing multimodality imaging probes. These probes can circumvent the limitations of the single mode of imaging. This combination of two different imaging methods can provide more accurate physiological and pathological information about disease diagnosis. Multiple imaging methods are expected to provide improved imaging capabilities and hence result in a faster and better understanding of metabolic mechanisms. The progress on medical technology has led to a variety of imaging techniques to diagnose the human body. Since the introduction of PET/CT scanners, multimodal imaging has received increasing attention. The positron emission tomography (PET) offers quantitative real time 3D-visualization of physiological and pathological processes in vivo by means of probes labeled with PET-nuclides.
In this work we report the synthesis of radioactively labeled iron oxide (Fe3O4) nanoparticles (NPs) as bimodal imaging agents. Magnetite NPs were synthesized via a solvothermal approach by in situ reduction. Moreover, the surface of NPs was modified by catechol-ligands such as dopamine. The crystallinity of NPs was evaluated by X-ray diffraction analysis. The quantitative confirmation of different functionalities was determined by Fourier-transform infrared spectroscopy and ultraviolet-visible microscopy. We used transmission electron microscopy to show size and shape of the NPs. Further surface functionalization was subsequently accomplished using the green and efficient copper-catalyzed azide-alkyne cycloaddition reaction and the carbodiimide coupling reaction between an amino group a carboxyl group by using N,N′-dicyclohexylcarbodiimide. The “Click-reaction” was used to covalently attach 18F as positron emitting source on the particle surface. Decay corrected radioactive yields of 29.1 % were obtained and particle stability in the culture medium was proven. Furthermore, in order to specifically target cancer cells in the body, nanoparticles were decorated with an over layer of folic acid molecules.
The combination of the properties of iron oxide NPs, which can be used for magnetic resonance imaging (MRI) and radioactive 18F labeling for positron emission tomography (PET), respectively, makes this approach an excellent strategy for enhanced cancer imaging.
8:00 PM - ES08.07.05
Residual Stress and Oxidation Behavior of Cr Alloy Coatings on Zircaloy-4
Jung-Hwan Park 1 , Yang-Il Jung 1 , Dong-Jun Park 1 , Byoung-Kwon Choi 1 , Hyun-Gil Kim 1 , Jae-Ho Yang 1
1 , KAERI, Daejeon Korea (the Republic of)
Show AbstractEver since the Fukushima accident, the major issue of nuclear researchers has been the improvement of oxidation resistance under a beyond-design accident. Therefore, accident tolerant fuel (ATF) has been widely studied, and coating technology for a nuclear fuel cladding surface has been considered to improve the high-temperature steam oxidation resistance of zirconium-based alloys. For the development of ATF claddings, several coating materials have been suggested, including Mo, TiN, TiAlN, FeCrAl and Cr. These coatings are generally deposited by physical vapor deposition (PVD) methods. The microstructure of PVD coatings is characterized by a columnar microstructure and local surface defects. The microstructure and defects affect the oxidation of protective films. The appearance of defects in coatings is closely related to residual stress in films. Therefore, the study of the residual stress of these protective coatings and relationship of this parameter with the deposition conditions and oxidation behavior is very interesting. In this study, Cr and Cr-Al coatings were deposited on a Zircaloy-4 tube by arc ion plating to investigate the relationship of the residual stress and corrosion resistance. The residual stress was measured by iso-inclination method with an x-ray diffractometer at room temperature. A high-temperature steam oxidation test at 1473 K was performed to evaluate the oxidation behavior of Cr and Cr-Al coatings.
8:00 PM - ES08.07.06
Effect of Radiation on Embrittlement and Matrix Cu Content of a RPV Weld with Different PWHT Conditions
Mikhail Sokolov 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractThe influence of temperature and cooling rate on the embrittlement and the copper level in the matrix has been investigated on a weld fabricated from the same weld wire used for HSSI Weld 73W. This weld has a relatively high bulk copper content, 0.32% wt. The heat treatment consisted of heating the material to the desired temperature, holding at the post-weld heat treatment (PWHT) temperature, and then cooling down to room temperature. Except for special cases, all PWHTs were performed with a heating and cooling rate of 15oF/h (8oC/h) to simulate the heating/cooling rate of a real vessel. In two special cases, material was heated with 15oF/h (8oC/h) rate but water quenched after holding at the PWHT temperature. The highest PWHT temperature was 650oC/24h, while the other PWHTs were 610oC/24h, typical PWHT of reactor pressure vessels (RPV), and 580oC/100h. Charpy impact properties were measured using sub-size 3x4 mm specimens and matrix Cu content was measured by atom probe tomography before and after irradiation. Small-angle X-ray scattering (SAXS) was used to measure size and volume fraction of copper-rich precipitates in the irradiated specimens. Charpy specimens were irradiated in the Ford Reactor at 288oC to 0.8x1019 neutron/cm2 (E>1MeV). It was found that the higher PWHT temperature resulted in higher Charpy upper-shelf energy (USE) with little effect on the ductile-to-brittle transition temperature (DBTT). The lower PWHT temperature and slower cooling rate were found to be beneficial in reducing the matrix Cu content. The matrix Cu content after irradiation was approximately the same for all three welds measured regardless of their different matrix Cu contents in the unirradiated condition. Consequently, the weld with the lowest PWHT temperature exhibited the lowest shift of DBTT and drop in USE. SAXS results reveal smallest volume fraction of copper-rich precipitates in this weld compared to others.
8:00 PM - ES08.07.07
Synthesis and Characterization of CeO2-Based Simulated Fuel Containing CsI
Yuki Takamatsu 1 , Ken Kurosaki 1 2 , Hiroto Ishii 1 , Yuji Ohishi 1 , Hiroaki Muta 1 , Shinsuke Yamanaka 1 3 , Kunihisa Nakajima 4 , Eriko Suzuki 4 , Shuhei Miwa 4 , Masahiko Osaka 4
1 , Osaka University, Suita, Osaka, Japan, 2 , JST PRESTO, Kawaguchi, Saitama, Japan, 3 , University of Fukui, Tsuruga, Fukui, Japan, 4 , Japan Atomic Energy Agency, Naka, Ibaraki, Japan
Show AbstractFollowing the accident at the Fukushima Daiichi Nuclear Power Plant in 2011, radioactive nuclides were released into the environment and caused serious radioactive contamination. In particular, caesium (Cs), which is a volatile fission product (FP), was a major substance of radioactive contamination because of its relatively large release amount and strong radiation toxicity. However, its release behavior from nuclear fuels during an accident has not been clarified. Understanding such behavior can contribute to improving the accuracy of the source term evaluation. Simulated nuclear fuels containing non-radioactive FPs, named SIMFUELS, can be used in laboratory experiments to understand such behavior. In the present study, we prepared cerium dioxide (CeO2) based simulated fuels containing Cs by using Spark Plasma Sintering (SPS), which allows us to make bulk samples rapidly with low temperature compared to conventional sintering methods. Here, CeO2 was used as a simulated material of UO2 due to the similar chemical and physical properties with those of UO2. We selected caesium iodine (CsI) as FP species, which has been assumed to be one of the chemical forms of Cs in irradiated fuels. We characterized thus obtained simulated fuels by X-ray diffraction (XRD) and scanning electron microscope/energy dispersive X-ray spectrometry (SEM/EDX). It was confirmed that CsI existed as small precipitates almost uniformly throughout the fuels. Further details of the synthesis method and results of the characterizations will be presented.
8:00 PM - ES08.07.08
Grain Size Dependence on the Thermal Conductivity of UO2
Yuki Takamatsu 1 , Ken Kurosaki 1 2 , Yuji Ohishi 1 , Hiroaki Muta 1 , Shinsuke Yamanaka 1 3
1 , Osaka University, Suita Japan, 2 , JST PRESTO, Kawaguchi, Saitama, Japan, 3 , University of Fukui, Tsuruga, Fukui, Japan
Show AbstractFuel pellets with large grain size have been developed for the purpose of suppressing the fission product (FP) gas release during irradiation. The released FP gas increases the fuel rod internal pressure and deteriorates the heat transfer at the pellet-cladding gap, thus the FP gas release rate should be decreased. On the other hand, the thermal conductivity generally increases as the grain size increases. Although the accurate evaluation of the thermal conductivity of UO2 pellet is important for understanding the fuel performance, the influence of the grain size on the thermal conductivity of UO2 has not been completely evaluated. In the present study, therefore, we prepared multiple UO2 pellets with different grain sizes and measured their thermal conductivity to quantitatively evaluate the grain size dependence on the thermal conductivity of UO2. The grain size was controlled by adjusting the conditions of ball milling and sintering (spark plasma sintering/pressureless sintering). Further details of the synthesis methods and the thermal conductivity will be presented.
8:00 PM - ES08.07.09
Nanoceramic Coatings—Outperforming Barriers against Tritium Permeation in Future Generation Fusion Systems
Matteo Vanazzi 1 , Daniele Iadicicco 2 , Francisco Garcia Ferre 1 , Marco Utili 3 , Marco Beghi 2 , Fabio Di Fonzo 1
1 , Istituto Italiano di Tecnologia - IIT, Milan Italy, 2 , Politecnico di Milano, Milan Italy, 3 , ENEA, Bologna Italy
Show AbstractConcerning the design of future fusion reactors, the most relevant concepts will take the tritium deuterium reaction as a reference. Thus, the availability of tritium to fuel the fusion reactors assumes a relevant importance. The breeding process starting from the Pb-16Li eutectic represents one of the focus point of technological R&D activities worldwide, as well as the tritium confinement. Indeed, the tritium permeation inhibition through the Breeder Blanket moduli is mandatory to achieve tritium balance in the reactor chain. Once tritium is produced, it must be appropriately confined to preclude losses by permeation and so the contamination of the outer parts of the reactor, due to its radiotoxicity. In order to minimize such losses, an adequate permeation barrier is required. Here, we report on optimized nanoceramic coatings, specifically alumina (Al2O3) films grown by Pulsed Laser Deposition (PLD) technique. PLD-grown alumina coatings appear suitable towards this task due to their chemical inertia, high density and amorphous structure. To investigate the performance of such a material, permeation tests are performed using custom-made facilities. Firstly, hydrogen is used in order to simulate the effects of tritium and several tests are performed at different conditions in terms of background temperature (from 350 to 650 °C). Fixed a standard film thickness and a hydrogen partial pressure of 100mBar, various coatings morphology are investigated. Results collected in this way indicate an excellent behavior, with a permeation reduction factor (PRF) up to 105 (PRF values required in the actual DEMO design move from 100 to 1000).
These observations are confirmed through further permeation tests under electron irradiation. In this second experimental campaign, heated samples (450 °C) are exposed to deuterium fluxes in the presence of a 2 MeV electron accelerator. PRF values remain almost constant even after irradiation and thermal cycling. Finally, to evaluate the chemical compatibility of Al2O3 films in liquid eutectic Pb-16Li, samples are subjected to short-time corrosion tests (1000 hours). Post-test analyses are accomplished by Scanning Electron Microscopy (SEM), Energy Dispersive X-ray (EDX) and Auger Electron Spectroscopy (AES). No corrosive attack on the steel substrate is detected. To conclude, collecting data confirm the effectiveness of this coating as a protective barrier against corrosion and tritium permeation phenomena, making PLD-grown Al2O3 a suitable candidate in future fusion nuclear reactors.
8:00 PM - ES08.07.10
Chemical and Electrochemical Dissolution Behavior of Barium Hollandites in Oxidizing Conditions
Priyatham Tumurugoti 1 , Xialoei Guo 2 , Gerald Frankel 2 , Jie Lian 1
1 , Rensselaer Polytechnic Institute, TROY, New York, United States, 2 , The Ohio State University, Columbus, Ohio, United States
Show AbstractIn the multi-barrier approach for nuclear waste form disposal, corrosion products of an exterior stainless steel canister may adversely affect the chemical durability of the interior waste form material. This study reports the electrochemical dissolution behavior of hollandites in oxidizing/acidic environments (Fe3+, CrO42- etc.) that simulate the induced repository conditions upon penetration of stainless canisters by corrosion. Barium hollandites with compositions, Ba1.15M3+2.3Ti5.7O16 (M = Al, Ti and Cr), and with controlled microstructures, are prepared by solid-state sintering followed by consolidation by spark plasma sintering. Dynamic polarization experiments coupled with dissolution experiments are used to identify the overall dissolution behavior and thereby evaluate the chemical durability. Interfacial chemical reactions are identified by surface characterization and leachant analysis techniques. Tunnel cations, Ba/Cs, generally are accommodated in the sites surrounded by octahedral framework hosting trivalent cation M3+. Hence oxidation of M3+ is a crucial factor, and the effects of electrochemical dissolution behavior of the barium hollandite waste forms will be analyzed. The nature and kinetics of the formation of a passivation layer for each composition, their dependence on microstructure, and the tunnel-ion release behavior before and after passivation will be discussed.
8:00 PM - ES08.07.11
(Ga55In45)2S300 Nanocrystallites as Novel Materials for Nonlinear Optical Detection of the Gamma Radiation
Iwan Kityk 1 , V. Halyan 2 , Katarzyna Ozga 1 , M. Piasecki 3
1 , Institute of Optoelectronics and Measuring Systems, Faculty of Electrical Engineering, Czestochowa University of Technology, Czestochowa Poland, 2 , Department of Experimental Physics, Eastern European National University, Lutsk Ukraine, 3 , Institute of Physics, J. Dlugosz University Czestochowa, Czestochowa Poland
Show AbstractWe have shown a drastic increase o sensitivity to the gamma radiation during use of (Ga55In45)2S300 nanocrystals with respect to bulk materials. The effects is maximal for the nanocrystals varying within the 30…50 nm. Initial single crystals (Ga55In45)2S300 and (Ga54.59In44.66Er0.75)2S300 possessing space group P61 were grown and explored with respect to laser induced nonlinear optical effects like second harmonic generation and third harmonic generation. The sensitivity of the nanocrystals is higher at least 20 % with respect to the bulk-like crystals. It was tested in the powder states as well as embedded into the different polymer matrices like PMMA, PVK, PVA etc. Structural parameters were evaluated using the Rietveld methods following the X-ray diffraction data. The chemical composition of the synthesized samples was controlled by EDS analysis. A principal role of Er3+-doping on the radiation sensitivity and related NLO is shown. The nanocrystals were fabricated by acoustical and mechanical millings have enhanced their sensitivity to gamma radiation at least 25 %.
The average energy of the incident g-rays was equal to about 1.25 MeV. Absorbed dose was controlled using a VDEG2-34 SP-1 device for the detection and measurement of γ-rays. The energy range of the γ-radiation detection was varied within the 0.05–3 MeV with the doses 420…5040 Gy.
Contrary to the compounds which utilize traditional photoluminescence here is proposed some multi-functional materials which use both photoinduced nonlinear optics and fluorescence. We use not a traditional nonlinear optics, however photoinduced nonlinear optics which is more sensitive and precise. That means - it may be used at least two physically independent methods. The principal physical origin of the enhanced sensitivity is caused by detection of the higher excited states, and formation of the photoinduced bicolour generated space ratings. The beams corresponding to the fundamental and doubled frequency (due to second harmonic generation) of Er:glass 20 ns laser with frequency repetition 10Hz and wavelength 1540 nm were optimal for s-p light polarizations. We have found that for the titled crystal the optimal angle was varied within the degree and the process of the treatment and was durated up to 2…3 min. The proposed materials may be proposed like high sensitive express sensors of the gamma irradiation.
8:00 PM - ES08.07.12
Irradiation-Induced Damage in Concrete—The Enthalpy Landscape Viewpoint
Mathieu Bauchy 1
1 , University of California, Los Angeles, Los Angeles, California, United States
Show AbstractConcrete is one of the primary structural elements in nuclear power plants. However, the harsh conditions experienced in nuclear power plants (high service temperature, vibrations, radiations) can compromise its integrity. In particular, aggregates can exhibit amorphization and swelling upon irradiation, which enhances the risk of cracking and alkali-silica reaction. Here, based on atomistic simulations, we show that the topography of the enthalpy landscape strongly governs irradiation-induced damage. At low deposited energy, irradiated minerals are found to explore forbidden regions of the enthalpy landscape, that is, inaccessible by heating and cooling. In contrast, it is shown that damage saturates when the system reaches the enthalpy landscape of an allowable liquid. At this stage, a sudden decrease of the height of the energy barriers enhances relaxation, thereby preventing any further accumulation of defects. The effect of the atomic topology on the enthalpy landscape and, hence, the resistance to irradiation is also discussed.
8:00 PM - ES08.07.13
Mathematical Modelling for The FLUOREX Nuclear Fuel Recycling Process
Artur Braun 1 , Shunji Homma 2
1 , Empa. Swiss Federal Laboratories for Materials Science and Technology, Duebendorf Switzerland, 2 Division of Materials Science, Saitama University, Saitama Japan
Show AbstractWe present a mathematical gas-solid reaction which is developed for use in the FLUOREX Nuclear Fuel Recycling Process. Specifically, it is representing the fluorination of uranium dioxide (UO2), which consists of a two-step reaction: the formation of a solid intermediate of uranyl fluoride (UO2F2) on the core of unreacted UO2 and the consumption of UO2F2. The model is an extension of the unreacted shrinking core model with a shrinking spherical particle and takes into account particle expansion resulting from the density difference between UO2 and UO2F2. This model successfully represents the initial expansion of the particle by the formation of the low-density UO2F2 intermediate. The accuracy of this model is higher than that of the original model, which does not allow particle expansion [Homma 2008].
[Homma 2008] Homma S, Uoi Y, Braun A, Koga J, Matsumoto S: Reaction model for fluorination of uranium dioxide using improved unreacted shrinking core model for expanding spherical particles. Journal of Nuclear Science and Technology 2008, 45:823-827.
8:00 PM - ES08.07.14
Evolution of Entropic and Vibrational Properties During Amorphization of SiC Following Radiation
William Lowe 1 , Jacob Eapen 1
1 Nuclear Engineering, North Carolina State University, Raleigh, North Carolina, United States
Show AbstractSilicon carbide (SiC) is a ceramic material with potential applications in high-temperature and extreme radiation environments such as those found in nuclear power reactors. SiC is known to amorphize under irradiation, resulting in deleterious changes to mechanical properties like volumetric swelling, reduced hardness, and reduced elastic modulus. From a design perspective, it is imperative to understand the underlying physics associated with amorphization in response to radiation. We expect that vibrational and entropic signatures will be evident during the accumulation of defects and the subsequent crystalline-to-amorphous transition. To investigate these processes in SiC, we compute and monitor the velocity autocorrelation function of the atoms after repeated radiation knocks using atomistic simulations. We then compute the entropic changes and phonon modes from the velocity autocorrelation.
Interatomic forces in SiC are described using a hybrid Tersoff potential that incorporates a bond order term in the attractive two-body force. To account for large repulsive forces during the cascade, the Tersoff potential is smoothly stitched with a Ziegler-Biersack-Littmark (ZBL) electrostatic screening function. Progressively disordered states are generated via successive recoil events. After each cascade, we compute the velocity autocorrelation of Si and C atoms, separately. We then extract the vibrational and configurational entropies using the recent two-phase method for computing free energy. Using the same velocity autocorrelation function, we also generate the phonon dispersion curves to elucidate the phonon softening and dynamic lattice instability following radiation. We thus quantify the changes to the structure and the attendant entropic increase along with the degeneration of phonon modes with increasing disorder. Interestingly we observe that Si atoms are driven to a dynamically arrested state upon radiation. We then show that this dynamic instability is the primary driver for the amorphization process in SiC under irradiation.
Symposium Organizers
Karl Whittle, University of Liverpool
Felix Brandt, Forschungszentrum Juelich
Philip D Edmondson, Oak Ridge National Laboratory
Blas Uberuaga, Los Alamos National Laboratory
ES08.08: Nuclear Fuel II
Session Chairs
Felix Brandt
Sarah Finkeldei
Wednesday AM, November 29, 2017
Hynes, Level 2, Room 206
8:30 AM - ES08.08.01
A Phase Field Model for High Burn-up Structure Formation and Evolution
Karim Ahmed 1 , Yongfeng Zhang 2 , Daniel Schwen 2 , Cody Permann 2 , Xianming Bai 3
1 Department of Nuclear Engineering, Texas A&M University, College Station, Texas, United States, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 3 , Virginia Polytechnic Institute and State University, Blacksberg, Virginia, United States
Show AbstractWe present a quantitative phase field model for investigating High Burn-up Structure (HBS) formation and evolution. The model takes into consideration the evolution of dislocation density, diffusion of gas atoms, and the migration of grain boundaries. As such, the model captures both the nucleation of sub-grains and bubbles. The model has been applied to study HBS formation in UO2. The model predicts a threshold dislocation density for nucleation of the sub-grains and a threshold gas concentration for bubble formation. The predicted threshold values are in good agreement with reported data in literature. The effect of initial grain size, dislocation density and dose rate on the kinetics of HBS formation and evolution were thoroughly investigated.
8:45 AM - ES08.08.02
Solid State Neutron Detection Employing Hydrothermally Grown Uranium and Thorium Dioxide
Christina Dugan 1 , James Petrosky 1 , John McClory 1 , Martin Kimani 2 3 , J. Matthew Mann 2 , Karl Rickert 4 , George Peterson 5 , Edward Cazalas 4 , Alyssa Mock 6 , Peter Dowben 6
1 Engineering Physics, Air Force Institute of Technology, Wright Patterson AFB, Ohio, United States, 2 Sensors Directorate, Air Force Research Laboratory (AFRL), Wright Patterson AFB, Ohio, United States, 3 , KBRWyle Aerospace Group, Ridgecrest, California, United States, 4 , Oak Ridge Institute for Science and Education, Oak Ridge, Tennessee, United States, 5 Mechanical & Materials Engineering, University of Nebraska-Lincoln, Lincoln, Nebraska, United States, 6 Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, Nebraska, United States
Show AbstractHydrothermally grown single-crystal UO2 crystals are being developed for potential use in solid-state neutron detectors. There is, however, a lack of fundamental characterization regarding these materials, particularly with regard to their electronic and physical characteristics. This work provides fundamental measurements of some of the bulk material properties in single crystal UO2. The excellent quality of the UO2 crystal’s electronic structure was confirmed with Hall measurements resulting in a mobility of 2.22 cm2/V.s, carrier concentrations of 4.75 x 1016 cm-3, and conductivity of 1.69 x 10-2 Ω -1 cm-1 as compared to literature values of 0.1 – 0.01 cm2/V.s, 1 x 1016 cm-3, and 6.66 x 10-4 Ω-1 cm-1 respectively. The quality of UO2 crystals physical structure was confirmed by Photoemission Spectroscopy and ellipsometry measurements. The occupied states in the vicinity of the valence band maximum and the conduction band minimum ascertain the placement of the band gap along with intrinsic growth defects. Measurements of the UO2 crystals provide a complex energy dependent dielectric constant in agreement with published literature values when applying Gauss-Lorentz and Tauc Lorentz oscillator models. Characterization of the electronic and physical structure of single crystal actinide oxides is an important step toward the goal of creating a compact, portable actinide neutron detector which provides a larger signal to noise ratio than prevailing technology.
9:00 AM - *ES08.08.03
UO2-Based Model Systems—A Bottom-Up Approach for Understanding the Matrix Corrosion of Spent Nuclear Fuel
Sarah Finkeldei 1 , Angela Baena 1 , Felix Brandt 1 , Guido Deissmann 1 , Martina Klinkenberg 1 , Raul Palomares 2 , Maik Lang 2 , Annika Maier 3 , Mats Jonsson 3 , Rodney Ewing 4 , Dirk Bosbach 1
1 , Forschungszentrum Juelich, Juelich Germany, 2 , UT Knoxville, Knoxville, Tennessee, United States, 3 , KTH, Stockholm Sweden, 4 , Stanford University, Stanford, California, United States
Show AbstractDemonstrating the long-term safety of a deep geological repository for spent nuclear fuel (SNF) requires a sound understanding of radionuclide release mechanisms from SNF. The concentration and distribution of fission products in the UO2 matrix, such as lanthanides and metallic epsilon-particles, influence the electrochemical properties of the SNF, which control its long-term matrix corrosion. Due to its chemical and structural complexity and the intense radiation field, studies on SNF cannot unravel all concurrent corrosion mechanisms. Here, a bottom-up approach of simplified UO2-based model systems is presented, which enables single-effect studies on processes affecting SNF matrix corrosion.
A toolbox for the fabrication of tailor-made ceramic UO2-based model systems (UO2, UO2+Nd, UO2+Pd) has been developed applying wet-chemical approaches such as coprecipitation, sol-gel and ion exchange reactions on weak acid resins. Nd was included to mimic the presence of lanthanide fission products (FP). The structural and microstructural changes of the model systems as a consequence of Nd-doping were studied in-depth by advanced methods. SEM studies showed an increase in grain size with increasing Nd concentration. Total neutron scattering with pair distribution function analysis revealed structural changes in particular in the oxygen sublattice. Different levels of local distortions of the fluorite structure were observed for Nd-doped ceramics in comparison to undoped UO2 reference samples. Results of Raman spectroscopy are in-line with these findings. To determine the reactivity of the various model materials oxidative dissolution experiments are carried out. The results of these experiments provide the link between the reactivity of the UO2-based model systems towards oxidative dissolution and their structural and microstructural properties.
9:30 AM - ES08.08.04
Atomic Scale Insights on the Microstructure Evolution of Urania under Irradiation
Alain Chartier 1 , Claire Onofri 2 , Laurent Van Brutzel 1 , Catherine Sabathier 2
1 , CEA-Saclay, Gif-Sur-Yvette France, 2 , CEA, Saint Paul lez-Durance France
Show AbstractUrania is commonly used as a fuel in nuclear industry. Urania is heavily irradiated during its in-reactor stay, and faces drastic microstructural modifications, including few percent swelling and increase of dislocation density. Dislocations loops nucleate first [1] and transform with increasing fluence into lines. However, the early stages of their nucleation are hardly attainable experimentally. One commonly infers that their nucleation is related to the aggregation of point defects or defects clusters into dislocations.
In the present paper [2], we clarify the first steps of the effect of irradiation on urania by means of molecular dynamics simulations using empirical potentials. The irradiation dose is simulated by continuous accumulation of Frenkel pairs at 600°C, skipping the cpu-expensive displacement cascades.
Starting from a defectless urania, we observe the nucleation and growth of dislocations under Frenkel pairs accumulation. Detailed analysis shows a four stages evolution : (i) an increase of point defects (ii) then the nucleation of Frank loops 1/3 <111> from the aggregation of point defects, (ii) the transformation of Frank loops into perfect loops 1/2 <110> (iv) and finally their stabilization as lines. Our simulations also show a swelling up to 3.2% during the first stage in which point defects are present. This swelling suddenly decreases to 1.5% in the second stage, as soon as dislocations nucleate.
Both stage (iv) and swelling agree with experimental data [1,3] and therefore strengthen the four stages scenario of the microstructure evolution of urania under irradiation.
[1] C. Onofri, C. Sabathier, H. Palancher, G. Carlot, S. Miro, Y. Serruys, L. Desgranges, and M. Legros, Nucl. Instrum. Meth. Phys. Res. B374 (2016) 51.
[2] A. Chartier, C. Onofri, L. Van Brutzel, C. Sabathier, O. Dorosh, and J. Jagielski, Appl. Phys. Lett. 109 (2016) 181902.
[3] J. Spino, and D. Papaionnou, J. Nucl. Mater. 281 (2000) 146.
9:45 AM - ES08.08.05
Exploring the Pellet-Cladding Interaction with Atomistic Simulation
Adam Plowman 1 , Conor Gillen 1 , Alistair Garner 1 , Philipp Frankel 1 , Chris Race 1
1 , University of Manchester, Manchester United Kingdom
Show AbstractFailure of fuel rods in light water reactors via the Pellet-Cladding Interaction (PCI) imposes stringent limits on operational parameters. PCI is thought to be due to stress corrosion cracking (SCC) under the action of aggressive species, such as iodine, released from the fuel pellets during fission. However, an incomplete understanding of the underlying mechanisms of PCI has resulted in conservative restrictions on reactor operation. Improved mechanistic understanding would allow these limits to be relaxed and enable reactors to respond to energy demand changes more quickly, without compromising safety. More nimble power manoeuvring is important when nuclear power is required to work alongside significant fluctuating sources of renewable energy.
In tandem with new experimental observations, we are using atomistic simulation to improve our mechanistic understanding of PCI. We present a systematic study of zirconium grain boundary properties, including cleavage energies, undertaken using density functional theory. We have further studied the thermodynamics of impurities (including iodine) in these boundaries and their effect on grain boundary cohesion. We compare our results with new experimental data on iodine-induced SC cracks in commercial Zr alloys.
10:30 AM - *ES08.08.06
Thorium Oxide-Based Fuels—Historic Notes, Ongoing Research and Future Trends
Marc Verwerft 1
1 Institute for Materials Science, Belgian Nuclear Research Centre (SCK-CEN), Boeretang Belgium
Show AbstractThorium has recently been presented in the popular press as “the green future of nuclear” and it might seem as if thorium based fuel cycles are a recent breakthrough that remained overlooked for decades. Alas, this is not the case: research on thorium as possible alternative resource for nuclear energy applications has a long history and it is hardly possible to account of all the accomplishments achieved over more than five decades of research on thorium fuel cycles. In this presentation, a brief summary will be given of the historic Light and Heavy Water Reactor thorium fuel research. Most of the early work was performed with the aim to achieve a thorium-uranium breeding cycle. Around fifteen years ago, several European research projects were launched to investigate the use of thorium-plutonium fuels in Light Water Reactors with the goal to reduce plutonium inventories. An overview will be given of the ongoing research in this more recent domain with an emphasis on materials research and solid state aspects. Some thoughts on the possible implementation of a thorium-plutonium based fuel cycle will be discussed.
11:00 AM - ES08.08.07
Thermodynamics of Advanced Uranium Silicide and Silicide-Nitride Fuel and Behavior with Ferritic Cladding
Emily Moore 1 , Mallikharjuna Bogala 1 , Vancho Kocevski 1 , Tashiema Wilson 1 , Theodore Besmann 1 , Jacob McMurray 2 , Elizabeth Sooby Wood 3 , Andy Nelson 3 , Simon Middleburgh 4 , Peng Xu 4
1 , University of South Carolina, Columbia, South Carolina, United States, 2 , Oak Ridge National Laboratory, Oakridge , Tennessee, United States, 3 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 4 , Westinghouse Electric, Västerås Sweden
Show AbstractNew candidates for Advanced Technology Fuels (ATF) that seek to improve performance, lower cost and increase tolerance to accident scenarios include uranium nitride and silicide compositions. Particular interest is given to U3Si2 and to a lesser degree U3Si5 for their increased thermal conductivity and U-density compared to current oxide systems. The proposed UN- U3Si2 composite is designed to surround grains of UN with U3Si2. A potential new cladding candidate consists of a ferritic FeCrAl(Y) alloy. The presented research focuses on the thermochemical behavior and compatibility of the U-Si, U-Si-N fuels and the ferritic alloy cladding. CALPHAD (CALculation of PHAse Diagrams) type models are being developed and are supported by experimental efforts and first principal calculations. Phase equilibria in important compositional regions of the U-Si system and U-Si-N ternary are being explored. Computational efforts to investigate the U-Fe-Si ternary phase formation from fuel-cladding contact are also presented.
This research is being performed using funding received from the DOE Office of Nuclear Energy's Nuclear Energy University Programs.
11:15 AM - ES08.08.08
Thermal and Mechanical Properties of U3Si2
Afiqa Mohamad 1
1 , Osaka University, Osakafu Japan
Show AbstractRecently, development of accident tolerant fuel (ATF) has become one of the primary focus after the accident in Fukushima. The potential fuel for enhanced ATF should have higher uranium density and higher thermal conductivity (λ) than those of the current nuclear fuel of UO2. Several candidates have been explored as alternative nuclear fuel materials for ATF. Among of them, U3Si2 is receiving more attention recently.
U3Si2 is currently acknowledged as one of the alternative fuels which can be used in light water reactors due to the higher l and higher uranium density compared with those of UO2. The density of U3Si2 is 11.3 g-U/cm3, larger than the 9.7 g-U/cm3 of UO2[1]. The λ of U3Si2 and UO2 at 1000 K are 12.4~17.6 W/mK and 3.6 W/mK, respectively. In addition, U3Si2 possesses good properties such as thermal stability up to melting temperature and good corrosion oxidation resistance [2]. For designing U3Si2-based ATF, it is important to accurately understand the properties of U3Si2 such as thermal and mechanical properties. For example, there are discrepancies of thermal conductivity in the previous reports[1]. In addition, there are no experimental data regarding the mechanical properties of U3Si2 such as elastic constant and hardness yet. Therefore, in the present study we aim to gain single phase U3Si2 and report about thermal as well as mechanical properties of U3Si2.
The nominal composition of U3Si2 was prepared from natural uranium and Si with 2 wt% of excess Si by arc melting in Ar atmosphere. The U3Si2 bulk sample was then synthesized by spark plasma sintering at a pressure of 75 MPa and a temperature of 1123 K followed by annealing at 1000 K for 72h. X-ray diffraction measurement showed that almost single phase of tetragonal U3Si2 was obtained. The average linear thermal expansion coefficient was evaluated from high temperature XRD and dilatometer methods. The obtained values were 16.9±0.2×10-6 K-1 over the temperature range from 300 K to 1200 K and 15.8±1.3×10-6 K-1 over the temperature range from 300 K to 1023 K, respectively. λ was calculated from the heat capacity (measured by DSC), measured density and thermal diffusivity. The λ increased with increasing temperature and the value range is between 6.9~18.9 W/mK. At 1000 K, λ obtained from the present study agrees well with the values reported by White et al [1], but higher than that from Shimizu et al [3]. The Vickers indentation and fracture toughness test were performed using a Vickers hardness tester at applied load of 9.8 N. The Vickers hardness and fracture toughness of U3Si2 were 7.51±0.41 GPa and 2.90±0.33 MPam1/2, respectively. The elastic constant of U3Si2 will be discussed in the presentation.
[1] J.T. White et al., J. Nucl. Materials, 464, 2015, 275-280.
[2] G.L. Hofman et al., Internal Report, ANL, Argonne, IL, 1996.
[3] Shimizu et al., Tech. Rep NAA-SR-10621, 1965.
11:30 AM - ES08.08.09
Simulations of Effective Thermal Conductivity in U3Si2 Using Phase-Field Microstructures
Linyun Liang 1 , Zhi-Gang Mei 1 , Yinbin Miao 1 , Abdellatif Yacout 1
1 , Argonne National Laboratory, Lemont, Illinois, United States
Show AbstractU3Si2 has attracted much attention as a potential nuclear fuel in more demanding reactor applications such as those of commercial light water reactors (LWR) due to its high uranium density, improvement in thermophysical properties, chemical and irradiation stability. Thermal conductivity being one of the most important property is essential to predicting the most important fuel performance. The limited availability of thermophysical properties for U3Si2 suggests that it possesses a thermal conductivity vastly superior to that of UO2. However, thermal conductivity is a microstructure sensitive property, and is directly affected by thermal- and irradiation-driven processes such as grain growth, gas bubbles, and precipitation. These processes can significantly change the thermal conductivity in both time and space during irradiation. In this work, we studied the effective thermal conductivity of single crystal and polycrystalline U3Si2 containing intragranular gas bubbles and intergranular gas bubbles, respectively. The microstructure evolution was captured by phase-field modeling approach. The effective thermal conductivities of the microstructure were analyzed with respect to average grain size and pore fraction. The correlation of effective thermal conductivity and pore fraction was derived for U3Si2. The effects of grain boundary thermal conductivity and alignment of intergranular gas bubbles on the effective thermal conductivities were investigated. The simulation provides an improved understanding of the thermophysical properties of U3Si2 for current and proposed fuel designs.
ES08.09/TC04.07: Joint Session: Advanced Nuclear Modeling
Session Chairs
Wednesday PM, November 29, 2017
Hynes, Level 2, Room 206
1:30 PM - *ES08.09.01 /TC04.07.01
Highlights of Advanced Nuclear Fuel Research within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program
Christopher Stanek 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThe US Department of Energy – Office of Nuclear Energy program Nuclear Energy Advanced Modeling and Simulation (NEAMS) is developing a mechanistic computational toolset for nuclear fuel design and/or analysis. Multiscale materials modeling of advanced fuels is an important element of this approach in order to allow the transition from empirical to more mechanistic models. By design, atomic and mesoscale models are necessarily connected to the development of an advanced fuel performance code. In this talk, several highlights of advanced nuclear fuel research within this approach will be provided, including insights in to doped-uranium dioxide and uranium silicide-based fuels.
2:00 PM - ES08.09.02/TC04.07.02
Temperature Accelerated Rate Matrix Construction in the ParSplice Framework
Thomas Swinburne 1 , Danny Perez 1
1 , T-1 Group, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractAtomistic simulations provide essential information to higher order simulation schemes by discovering new system states and evaluating the rate of interstate transitions. However, as the majority of interstate transitions are very rare on typical simulation timescales accelerated techniques are required in order to explore nontrivial regions of state space. Here, we combine the recently developed ParSplice simulation framework with the temperature accelerated dynamics method to construct low temperature rate matrices, optimizing the use of massively-parallel computational resources through an uncertainty quantification scheme. The key concepts will be presented and applications relevant to nuclear materials science will be discussed.
2:15 PM - ES08.09.03 /TC04.07.03
Understanding the Amorphization Resistance of Complex Oxides via Machine Learning
Ghanshyam Pilania 1 , Karl Whittle 2 , Chao Jiang 3 , Robin Grimes 4 , Christopher Stanek 1 , Kurt Sickafus 5 , Blas Uberuaga 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , University of Liverpool, Liverpool United Kingdom, 3 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 4 , Imperial College London, London United Kingdom, 5 , University of Tennessee, Knoxville, Knoxville, Tennessee, United States
Show AbstractThe response of complex oxides to irradiation is dictated by a number of factors that are challenging to connect to experimental observables. For example, the critical amorphization temperature Tc, the temperature at which a compound can no longer be amorphized, is an inherently kinetic property that depends on the behavior of multiple defect types in a chemically disordered system. Developing a predictive capability based on first principles for such properties is daunting. Here, we use machine learning to relate Tc to fundamental properties of pyrochlores (A2B2O7). We use basic properties of the elemental constituents as well as DFT-computed phase energetics as features in a kernel-based ridge regression learning framework. We identify the energy of amorphization as a critical feature. This framework enables design maps that estimate Tc as a function of pyrochlore chemistry. This work highlights the utility of machine learning to provide fundamental insight into inherently complex materials problems.
2:30 PM - ES08.09/TC04.07
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3:30 PM - *ES08.09.04 /TC04.07.04
Increasing the Power of Accelerated Molecular Dynamics Methods and Plans to Exploit the Coming Exascale
Arthur Voter 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractMany important materials processes take place on time scales that far exceed the roughly one microsecond accessible to molecular dynamics simulation. Typically, this long-time evolution is characterized by a succession of thermally activated infrequent events involving defects in the material. In the accelerated molecular dynamics (AMD) methodology, known characteristics of infrequent-event systems are exploited to make reactive events take place more frequently, in a dynamically correct way. For certain processes, this approach has been remarkably successful, offering a view of complex dynamical evolution on time scales of microseconds, milliseconds, and sometimes beyond. We have recently made advances in all three of the basic AMD methods (hyperdynamics, parallel replica dynamics, and temperature accelerated dynamics (TAD)), exploiting both algorithmic advances and novel parallelization approaches. I will describe these advances, present some examples of our latest results, and discuss what should be possible when exascale computing arrives in roughly four years.
4:00 PM - ES08.09.05 /TC04.07.05
Modeling Point Defects in Alloys with DFT, Cluster Expansions and KMC
Normand Modine 1 , Alan Wright 1 , Stephen Lee 1 , Stephen Foiles 1 , Corbett Battaile 1 , John Thomas 2 , Anton Van der Ven 2
1 , Sandia National Laboratories, Albuquerque, New Mexico, United States, 2 Materials Department, University of California, Santa Barbara, California, United States
Show AbstractModeling defects in alloys is a challenging problem because defect properties are sensitive to the occupations of nearby atomic sites and thus vary with location in the alloy. This leads each defect species to have an entire distribution of formation energies, defect levels, activation energies for diffusion, etc. in an alloy. Furthermore, defects can form, diffuse, and annihilate by slow, activated processes with time scales of seconds or even years. Unless the defects are in equilibrium, the distributions of defect properties will change over these same time scales. Density Functional Theory (DFT) allows the accurate determination of ground state and transition state energies for a defect in a particular local environment in the alloy but requires thousands of processing hours for each such calculation. Kinetic Monte-Carlo (KMC) can be used to model the relevant slow, activated processes and the changing distribution of defect properties but requires energy evaluations for millions or billions of local environments. We have used the Cluster Expansion (CE) formalism to “glue” together these seemingly incompatible methods in order to model defect diffusion in alloys. In the CE approach, the occupation of each alloy site is represented by an Ising-like variable, and products of these variables are used to expand quantities of interest. Once a CE is fit to a training set of DFT energies, it allows very rapid evaluation of the energy for an arbitrary configuration, while maintaining the accuracy of the underlying DFT calculations. These energy evaluations are then used to drive our KMC simulations. We will demonstrate the application of our DFT/MC/KMC approach to model thermal and carrier-induced diffusion of radiation-induced point defects in III-V alloys and show that trapping in energetically favorable regions of the alloy leads to a diffusion rate the slows dramatically with time. We will also discuss application of our approach to doping and the formation of compensating defects in semiconductors.
Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology and Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA-0003525.
4:15 PM - ES08.09.06 /TC04.07.06
Electronic Stopping Power in a Non-Uniform Electron Gas
Magdalena Caro 1 , Artur Tamm 2 , Alfredo Correa 2 , Alfredo Caro 3
1 , Department of Mechanical Engineering, Virginia Polytechnic Institute and State University, Falls Church, Virginia, United States, 2 , Lawrence Livermore National Laboratory, Livermore, California, United States, 3 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractTheoretical predictions of energy losses for projectiles traveling through solid targets are usually based on results for a uniform electron gas, jellium. In those models, dissipation is a function of the electron gas density.
In this work, we use Time Dependent Density Functional Theory, TD-DFT, to calculate the energy dissipated by energetic projectiles on non-uniform electron density targets, namely the case of binary collisions, and the case of a projectile traveling along a high symmetry direction in an fcc crystal. In particular, we study the case of a Ni projectile traveling in a Ni target along channeling trajectories. We relate the instantaneous dissipation, β, to the local electron density, ρ, experienced by the projectile, and find that β is a multivalued function of the host electronic density, represented by loops in the β-ρ plane, in contrast to all published results describing dissipation in jellium.
We conclude that real inhomogeneous electron gases have a significantly different effect on dissipation, and that jellium results represent an average approximation for the actual dissipation.
This work was supported as part of the Energy Dissipation to Defect Evolution (EDDE), an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Basic Energy Sciences (Award Number 2014ORNL1026).The authors acknowledge computing support from the Lawrence Livermore National Laboratory Institutional Computing Grand Challenge program.
4:30 PM - ES08.09.07 /TC04.07.07
Modeling Radiation Damage Using SRIM, MD and AKMC
Steven Kenny 1 , Mark Wootton 1
1 , Loughborough University, Loughborough United Kingdom
Show AbstractA model of an experiment which uses proton based radiation damage to simulate material damage in a nuclear power plant has been created using a combination of SRIM, molecular dynamics and adaptive kinetic Monte Carlo. Through the use of this model multiple radiation damage events in an iron chrome material have been simulated with collision cascade energies and rates chosen from a distribution consistent with bombardment by 3 MeV protons. These results will be compared and contrasted with a model where the material is simulated purely using molecular dynamics, where although the collision cascade energies can be matched the rate of bombardment is many orders of magnitude too high. The results show that a significant amount of annealing of defects take place in the material when realistic timescales are modelled between cascade events. This would be expected to lead to significant changes in the damage evolution in the material.
4:45 PM - ES08.09.08 /TC04.07.08
Accelerated Quantum Molecular Dynamics
Enrique Martinez 1 , Christian Negre 1 , Danny Perez 1 , Marc Cawkwell 1 , Arthur Voter 1 , Anders Niklasson 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThe accurate study of the long-term evolution of rare events is extraordinarily challenging as computations are arduous and quantum-based molecular dynamics simulation times are limited to, at most, hundreds of ps. Here, the Extended Lagrangian Born-Oppenheimer molecular dynamics formalism is used in conjunction with Parallel Replica Dynamics to obtain an accurate tool to describe the long-term dynamics of reactive benzene. Langevin dynamics has been employed at different temperatures to calculate the first reaction times in a periodic benzene sample at different pressures. Our coupled engine run for times on the order of the ns (two to three orders of magnitude longer than traditional techniques) and is capable of detecting reactions characterized by rates significantly lower than we could study before.
Symposium Organizers
Karl Whittle, University of Liverpool
Felix Brandt, Forschungszentrum Juelich
Philip D Edmondson, Oak Ridge National Laboratory
Blas Uberuaga, Los Alamos National Laboratory
ES08.10: Reactor Materials
Session Chairs
Thursday AM, November 30, 2017
Hynes, Level 2, Room 206
8:30 AM - ES08.10.01
The Defect Clusters Evolution Mechanisms During the Cavity Swelling in Irradiated RAFM Steels
Mingjie Zheng 1 , Jiawei Fu 1 , Shenyang Hu 2 , Xiaodong Mao 1 , Shaojun Liu 1
1 Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, CAS, Hefei, Anhui, China, 2 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractThe Reduced Activation Ferritic/Martensitic (RAFM) steel has been selected as the primary candidate structural material for the fusion DEMO and first fusion power plant. One reason for such selection is due to the good radiation resistance of RAFM steel, however, the radiation-induced cavity swelling in RAFM steel may become unacceptable once the radiation dose is high enough (>50 dpa). The key issue is to understand the underlying mechanisms for the onset of the cavity swelling, including the mechanisms for the nucleation and growth of defect clusters. In this work, the effects of the nucleation and growth of the defect clusters on the cavity swelling in RAFM steels are studied. A model for the evolution of the radiation-induced voids, bubbles and cavities was developed based on the cluster dynamics at the mesoscale. We found that the nucleation of these defect clusters is related to the incubation period of the cavity swelling in RAFM steels. The existence of helium atoms stabilizes the nucleation of small voids and enhances the nucleation and growth of cavities, which results in the onset of cavity swelling. As radiation dose increases, the size of cavities grows faster than their number densities and leads to the increasing swelling. The Cr content effects were studied through the dependence of cluster mobility on the Cr content. The slow down of cluster mobility due to the existance of Cr delay the growth of cavities and hence put off the cavity swelling. This research reveals the mechanism of the cavity swelling and will have potential applications in updating the RAFM steels with higher radiation-induced swelling resistance.
8:45 AM - *ES08.10.02
Cluster Dynamic Modeling of Mn-Ni-Si Precipitates in Low-Cu RPV Steels
Huibin Ke 1 , Peter Wells 2 , Philip D Edmondson 3 , Nathan Almirall 2 , Leland Barnard 4 , G. Robert Odette 2 , Dane Morgan 1
1 Computational Materials Group, Department of Materials Science and Engineering, University of Wisconsin–Madison, Madison, Wisconsin, United States, 2 , University of California, Santa Barbara, Santa Barbara, California, United States, 3 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 4 , Elysium Industries, Boston, Massachusetts, United States
Show AbstractFormation of large volume fractions of Mn-Ni-Si precipitates (MNSPs) may cause significant irradiation embrittlement of some reactor pressure vessel (RPV) steels under life-extension conditions. Atom probe tomography (APT) shows these precipitates can also form even in Cu-free alloys. To better understand these precipitates, a semi-empirical cluster dynamics model has been developed to study the evolution of MNSPs in low-Cu RPV steels.
The model is based on CALPHAD thermodynamics and radiation enhanced diffusion kinetics. The thermodynamics dictates the compositional and temperature dependence of the free energy reductions that drive precipitation. The model treats both homogeneous and heterogeneous nucleation, where the latter occurs on cascade damage, like dislocation loops. The model has only four adjustable parameters that were fit to a large body of high quality experimental atom probe tomography (APT) data. The model predictions are in semi-quantitative agreement with systematic Mn, Ni and Si composition variations in alloys characterized by APT, including a sensitivity to local tip-to-tip variations even in the same steel. The model predicts that heterogeneous nucleation plays a critical role in MNSP formation in lower alloy Ni contents. Single variable assessments of compositional effects show that Ni plays a dominant role, while even small variations in irradiation temperature can have a large effect on the MNSP evolution. For purposes of illustration, the effect of MNSPs on transition temperature shifts are presented based on well-established microstructure-property and property-property models.
The model will serve as a foundation for future modeling of Cu-Mn-Ni-Si precipitation in Cu bearing RPV steels. Further, this physics-based model can guide reduced-order model, but still physically based, correlation equations fit to the actual surveillance and test reactor databases. Finally, the model can provide basis for embrittlement prediction interpolation and extrapolation.
9:15 AM - *ES08.10.03
Evolution of Irradiation-Induced Damage in Low Alloy Steels—Role of Advanced Microstructural Analysis in Developing Mechanistic Understanding of Irradiation Embrittlement
Grace Burke 1
1 Materials Performance Center, University of Manchester, Manchester United Kingdom
Show AbstractUnderstanding the nanoscale features and compositional changes induced by neutron irradiation is essential for the development of mechanistic models that can be used to aid in the development of improved and optimized alloys. The hardening and reduction in toughness of low alloys steels and welds used in light water reactors is associated with the nanoscale features produced by neutron irradiation. The physical changes in the microstructure associated with the degradation in properties can be characterised using advanced analytical techniques, particularly very high spatial resolution analytical electron microscopy. The successful application of such nanoscale analysis techniques is providing essential data concerning irradiation-induced solute clustering and segregation that previously was only attainable using atom probe microanalysis. Numerous applications and the implications on our understanding of irradiation damage in low alloy steels and welds will be discussed.
9:45 AM - ES08.10.04
Understanding Lithium-Accelerated Zircaloy Corrosion—Coupled Chemical and Mechanical Effects
Jing Yang 1 , Mostafa Youssef 1 , Bilge Yildiz 1 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 AbstractLithium incorporation in zirconium oxide is relevant to a well-known cladding corrosion phenomenon. In commercial nuclear reactors, LiOH is added into coolant water in contact with the cladding layer in order the balance the pH value of boric acid, which is added as neutron absorber. It was found in earlier work that when zirconium alloy is immersed in aqueous solution containing high concentration of lithium ion, the oxide growth rate is significantly increased. This phenomenon, called lithium-accelerated corrosion, is detrimental because lithium ions tend to concentrate at certain regions at the oxide film surface and cause localized corrosion. Decoupling the complicated chemical and mechanical effects imparted by Li incorporation is difficult experimentally. In this work, we performed first-principles and thermodynamic calculations on the effect of lithium incorporation in zirconium oxide. The results show that chemically, lithium presence in the ZrO2 leads to faster oxygen diffusion, and thus can accelerate corrosion rate. This is because lithium exists in ZrO2 as a positively charged interstitial defect, and so it raises the concentration of negatively charged oxygen interstitials and zirconium vacancies. Mechanically, lithium affects the compressive stress at which the protective tetragonal phase of ZrO2 is stabilized. Li interstitials shrinks the volume of the oxide matrix, releases the compressive stress that is needed for stabilizing tetragonal phase ZrO2 and promotes tetragonal-to-monoclinic phase transformation. The increase of zirconium vacancy concentration also leads to an increase in the phase transformation critical stress and further makes the T-to-M phase transformation favorable. These findings help gain mechanistic understanding of lithium-accelerated corrosion of zirconium alloy as well as show the importance of considering intrinsic defects response to extrinsic ion insertion.
ES08.11: Nanoscale Effects I
Session Chairs
Thursday PM, November 30, 2017
Hynes, Level 2, Room 206
10:30 AM - *ES08.11.01
Towards Microstructure Sensitive Radiation Damage Predictions—A Hybrid Discrete Dislocation Dynamics and Cluster Dynamics Framework
Laurent Capolungo 1
1 , Los Alamos National Laboratory, Atlanta, Georgia, United States
Show AbstractUpon subjecting a metallic material system to irradiation or to ion implantation, its microstructure will significantly change as a result of the instantaneous collision between the high-energy particles and the material considered. Over long periods of time as damage accumulates or simply as a result of thermal fluctuations, the defects induced by irradiation (e.g. Frenkel pairs, vacancy clusters, self-interstitial atom clusters) can migrate and/or interact with other defects. Microstructure evolutions during irradiation is thus the result of the collective interactions between a wide spectrum of defects which migrate and interact over time scales that can differ by orders of magnitudes. The work to be presented aims at accounting for the effect of internal stresses –due to deformation history- on the kinetics of radiation damage. To this end, a spatially resolved stochastic cluster dynamics method is concomittently used with discrete dislocation dynamics in order to predict the effect of dislocation densities on both rate and spatial distribution of radiation induced defects. Thw work to be presented will introduce a numerically based homogenization schemes allowing for the baton pass of information between the discrete model of plasticity induced defects and the continuous represenation of radiation induced defects.
11:00 AM - ES08.11.02
In Situ Studies on the Irradiation Response of Nanoporous Metals
Jin Li 1 2 , Cuncai Fan 1 , Youxing Chen 3 , Haiyan Wang 1 , Xinghang Zhang 1
1 Materials Engineering, Purdue University, West Lafayette, Indiana, United States, 2 MSEN, Texas A&M University, College Station, Texas, United States, 3 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractHigh-energy particle radiation induces severe microstructural damage in metallic materials. Nanoporous (NP) materials have great potentials to alleviate irradiation-induced damage due to their giant surface-to-volume ratio. Here we show, by in situ irradiation studies on NP Au, both defects and nanopores evolve with irradiation condition. It is observed in room temperature studies that nanopores shrink due to the absorption of irradiation-induced defects, and the shrinkage rate is pore-size-dependent. At elevated irradiation temperatures, the behavior of defects and nanopores develops differently. For instance, higher temperatures result in lower defect density and reduced shrinkage rate of nanopores. Moreover, NP Au exhibits significantly enhanced swelling resistance compared to coarse-grained Au. Potential mechanisms for irradiation resistance of NP metals are discussed. This study sheds light on the design of radiation-tolerant NP metallic materials.
11:15 AM - ES08.11.03
Grain Size Saturation in an Irradiated Thermally Stabilized Nanocrystalline Alloy
Prince Singh 1 , Di Chen 2 , Lin Shao 2 , Yoosuf Picard 1 , Maarten De Boer 1
1 , Carnegie Mellon University, Pittsburgh, Pennsylvania, United States, 2 , Texas A&M University, College Station, Texas, United States
Show AbstractNanocrystalline (nc) metals are of great interest in nuclear materials applications because grain boundaries may act as effective recombination sites for point defects. Consequently, they may be able to sustain high irradiation doses with minimal damage. Here we investigate nc-NiW, a thermally stabilized nc-metal with an initial grain diameter of 6 nm. We find that when subject to low doses of Ni+ self-ion irradiation, grain growth is not distinguishable from that in nc-Ni. However, once the grains grow to an average diameter of 32 nm at 10 displacements per atom (DPA), grain growth saturates up to 100 DPA. Such saturation is not predicted by previous thermal spike models. From 0 to 10 DPA, the microstructure evolves from a fiber to a (111) biaxial texture, while a high fraction of low-energy grain boundaries develops. Up to 100 DPA, the microstructure does not evolve further. The result is not consistent with experiments in which ion channeling induced texture. On the other hand, grain growth from such a small initial grain size will significantly increase residual tension. Yet, strain energy minimization would favor soft grains with (100) texture, which is again not observed. A tentative model that considers a competition between the increase in film strain energy in (111) grains and the lowering of their grain boundary energy as grain size increases is in good agreement with the results.
11:30 AM - *ES08.11.04
Atomistic to Mesoscale Modeling of Precipitation Hardening in FeCu Alloys
Xianming Bai 1 , Yaxuan Zhang 1 , Huibin Ke 2 , Yongfeng Zhang 3 , Benjamin Spencer 3
1 , Virginia Tech, Blacksburg, Virginia, United States, 2 , University of Wisconsin–Madison, Madison, Wisconsin, United States, 3 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractReactor pressure vessels are made of Fe-based low-alloy ferritic steels that contain many types of minor alloy elements and impurities such as Cu. Since Cu has a very low solubility in Fe, the radiation and high temperature environment in reactors can cause the precipitation of nanometer size Cu-rich clusters. These clusters become obstacles for dislocation gliding, which in turn cause the hardening and embrittlement of the steels. In this work, firstly mesoscale cluster dynamics modeling is used to model the radiation enhanced precipitation of Cu clusters in dilute FeCu alloys. The calculated Cu precipitation kinetics information is directly coupled with a dispersed barrier hardening model to predict the radiation hardening in the alloys. Good agreement between modeling and experiments is obtained. The implementation of the cluster dynamics models in the MOOSE-based Grizzly software will also be discussed. Next some fundamental assumptions in the mesoscale models are verified with atomistic scale molecular dynamics simulations. Specifically, cascade simulations are conducted to study the production of Cu interstitials in the alloys and their subsequent evolution to verify the assumption of vacancy mediated Cu diffusion in the alloys. The diffusion of matrix point defects at different Cu concentration is also conducted to verify the assumption of solute drag effects on matrix defect diffusion. These atomistic results may provide valuable insight for developing more robust mesoscale models in the future.
ES08.12: Radiation Damage in Metals II
Session Chairs
Thursday PM, November 30, 2017
Hynes, Level 2, Room 206
1:30 PM - *ES08.12.01
The Stored Energy Fingerprints of Radiation Damage
Penghui Cao 1 , Charles Hirst 1 , Rachel Connick 1 , Logan Abel 1 , Sean Lowder 1 , Ki-Jana Carter 1 , Kangpyo So 1 , R. Scott Kemp 1 , Michael Short 1 , Mikhail Merezhko 2 , Diana Merezhko 2 , Oleg Rofman 2 , Kira Tsay 2 , Oleg Maksimkin 2 , Brian Turner 3 , Kevin Menard 3 , Frank Garner 4
1 , Massachusetts Institute of Technology, Cambridge, Massachusetts, United States, 2 , Institute of Nuclear Physics, Almaty Kazakhstan, 3 , Mettler-Toledo, Polaris, Ohio, United States, 4 , Nuclear Effects Consulting, Richland, Washington, United States
Show AbstractThe current unit of radiation damage, the displacements per atom (DPA), is a calculated exposure parameter that does not directly yield the defect populations responsible for irradiation-induced material properties. Were an 'a posteriori' measure of radiation damage to exist, it would help to answer numerous, lingering questions about the nature and effects of irradiation. We propose the use of stored energy fingerprints as this new, more descriptive unit of radiation damage. They can be measured after irradiation, and they are hypothesized to yield information about the resulting defect populations. A combination of time-accelerated molecular dynamics (MD) simulations and nanoscale differential scanning calorimetry (nanoDSC) measurements is presented, which together paints a more measurable picture of the multiscale nature of radiation damage. We focus on two specific examples: single radiation damage cascades in single crystal metals, and irradiation-induced martensitic (reverse Bain) transformations in metastable austenitic stainless steels. First, simulating and measuring low-dose irradiation in single crystal metals provides insights into the actual defects created in the early stages of radiation damage. Second, monitoring irradiation-induced martensitic transformations in the absence of deformation transforms metastable austenitic stainless steels into built-in dosimeters, as confirmed by both magnetic and calorimetric measurements. Potential applications range from settling the question of neutron/ion irradiation equivalency, to quantitatively understanding dose rate effects, to verification of historical uranium enrichment.
2:00 PM - ES08.12.02
Low Temperature Diffusivity of Self-Interstitial Defects in Tungsten
Thomas Swinburne 1 2 , Leo Ma 2 , Sergei Dudarev 2
1 , T-1 Group, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , CCFE, UKAEA, Abingdon United Kingdom
Show AbstractThe low temperature diffusivity of nanoscale crystal defects, where quantum mechanical fluctuations are known to play a crucial role, are essential to interpret observations of irradiated microstructures conducted at cryogenic temperatures. Using density functional theory calculations, quantum heat bath molecular dynamics and open quantum systems theory, we evaluate the low temperature diffusivity of self interstitial atom clusters in tungsten valid down to temperatures of 1K. Due to an exceptionally low defect migration barrier, our results show that interstitial defects exhibit very high diffusivity of order 103μm2/s over the entire range of temperatures investigated.
2:15 PM - ES08.12.03
Transmutation-Induced Precipitation in Neutron-Irradiated Tungsten
Xunxiang Hu 1 , Chad Parish 1 , Kun Wang 1 , Yutai Katoh 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractAs the leading plasma facing material in fusion reactors, tungsten is confronted with extremely hostile environment, characterized by high temperature, and high fluxes of heat and particles (i.e., D, T, He, and neutrons). One of the primary concerns is the generation of transmutation elements (i.e., Re, Os) and the subsequent radiation-induced segregation and precipitation, and the resulting mechanical property degradation induced by the 14 MeV-peak neutron irradiation. In this study, we have used advanced electron microscope methods to explore the response of tungsten to high dose neutron irradiation in the High Flux Isotope Reactor, focusing on the detailed characterization of irradiation-induced W-Re-Os precipitates in polycrystalline tungsten through TEM, X-ray mapping in STEM, multivariate statistical analysis data-mining of the X-ray data and transmission Kikuchi diffraction. The association of voids and precipitates, the chemical compositions, crystal structures and phases of precipitates along the grain boundary and within the grains were identified. The results showed that the intragranular precipitates are laves-phase while the precipitates along the grain boundaries appear to be Os2ReW HCP. The kinetics process of transmutant elements and radiation defects were briefly discussed to reveal the formation process of the observed precipitates. In addition, we also investigated the hardening contribution of W-Re-Os precipitates. A dispersed barrier hardening model informed by the available microstructure data was used to predict the hardness. The results indicated that the formation of intermetallic second phase precipitate dominant the radiation-induced strengthening with a relatively modest dose (>0.6 dpa). The hardening strength factor of the transmutation-induced precipitates was also determined to be 0.6. Moreover, the strength of the neutron irradiated tungsten increased up to the modest dose level (~1 dpa) and then began to decrease as the irradiation dose continued to increase. This behavior imposed more challeges to understanding the mechncial property degradation arising from the production of precipitation in neutron-irradiated tungsten. A robust undestanding of transmutation-induced precipitation in tungsten is expected to aid the research and development of W-based plasma facing materials in fusion reactors.
2:30 PM - ES08.12.04
Nano-Oxide-Dispersed Ferritic Steel for Fusion Energy System
Luke Hsiung 1 , David Hoelzer 2 , Michael Fluss 3
1 , Lawrence Livermore National Laboratory, Livermore, California, United States, 2 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 3 Nuclear Engineering, University of California, Berkeley, Berkeley, California, United States
Show AbstractA critical challenge in designing fusion power reactors is to develop high-performance, low-activation structural materials for first wall and blanket components, which will be exposed to an intense high-energy (14 MeV) neutron flux and helium (He) and hydrogen (H) transmutation gases. The intense neutron flux will generate large numbers of point defects which give rise to cavity formation. The process of cavity formation can be further promoted in the presence of helium and hydrogen gases in stimulating the formation of large helium bubbles and voids which lead to void swelling. The formation of bubbles and voids at grain boundaries can potentially cause a premature failure and reduce the lifetime of the first wall/blanket components of a fusion reactor. We now have experimental data resulting from (Fe + He +H) simultaneous ion-beam experiments that show how nanoparticles dispersed in a steel matrix can mitigate the accumulation of radiation damage that deleteriously change the dimensions of materials through swelling.
ES08.13: Fuel and Cladding I
Session Chairs
Thursday PM, November 30, 2017
Hynes, Level 2, Room 206
3:15 PM - *ES08.13.01
Design and Installation of an Analysis Capability for Reactor Irradiated Materials
S. Morgan 1
1 , National Nuclear Laboratory, Sellafield United Kingdom
Show AbstractThe challenges of working with reactor irradiated materials are relatively well known and understood. The nuclear industry as a whole is able to make provision for the examination of these materials when required, and can utilise a range of experimental and measurement methods to do so.
Post irradiation examination (PIE), whether on a large multi-fuel element scale or a micro electron microscopy scale, plays a key part in understanding current reactor designs and in development of future designs. The initial stages of PIE are often undertaken in a heavily shielded facility, many of which are now ageing and in need of some form of refurbishment or replacement.
This paper will discuss the challenges inherent in integrating a new materials analysis capability in an older, heavily shielded facility. The potential for using information from existing irradiated material stocks to assist in the design and development of future materials will be discussed with the aid of a case study.
3:45 PM - ES08.13.02
In Situ Ion Irradiation of Multilayer (TiN, TiAlN) Ceramic Coating for Accident Tolerant Zr-Alloy Fuel Claddings
Jing Hu 1 , Douglas Wolfe 2 , Arthur Motta 2 , Meimei Li 1 , Mark Kirk 1
1 , Argonne National Laboratory, Lemont, Illinois, United States, 2 , The Pennsylvania State University, University Park, Pennsylvania, United States
Show AbstractAccident tolerant fuel (ATF) claddings are being developed to help mitigate the consequences of loss of coolant accidents in a nuclear reactor. In this study, multilayer (TiN, TiAlN) ceramic coatings with compositions of Ti20Al80N and Ti50Al50N deposited on ZIRLO® were subjected to ion irradiation in situ. Preliminary in situ ion irradiation of samples with 1 MeV Kr ions at 300 °C up to 20 dpa on the coatings showed good irradiation resistance of the coatings with the integrity of the coatings remaining after irradiation. No irradiation defects above 1 nm during irradiation and obvious change in existing coating columnar grain structure and size were observed. A limited amount of porosity at the interface did not to grow under irradiation, and the coating showed good adhesion to the Zr matrix. Detailed analysis of the ion irradiated samples with different layer architecture and composition provided a better understanding of the irradiation performance and clearer guidance for future neutron irradiations.
4:00 PM - ES08.13.03
Assessment of Irradiation Damage and Chemical Interactions in Neutron Irradiated U-10Zr fuel and HT9 Cladding with High-Energy X-Rays
Jonova Thomas 1 , Sri Nori 1 , Alejandro Figueroa 1 , Ran Ren 2 , Peter Kenesei 3 , Jon Almer 3 , Jason Harp 4 , Maria Okuniewski 1
1 Department of Materials Engineering, Purdue University, West Lafayette, Indiana, United States, 2 Department of Civil Engineering, Purdue University, West Lafayette, Indiana, United States, 3 Advanced Photon Source, Argonne National Laboratory, Argonne, Illinois, United States, 4 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractHigh energy X-rays generated from synchrotron sources are beneficial for the characterization of nuclear materials and fuels that have undergone drastic changes in reactors when exposed to extreme environmental conditions during irradiation. These X-rays have the capability of penetrating through high-Z materials. The Advanced Photon Source (APS) at Argonne National Laboratory (ANL) is among one of the few facilities that can provide these unique characterization techniques on bulk neutron irradiated materials utilizing high energy X-ray synchrotron tomography, far-field diffraction, and near-field diffraction. The effects of neutron irradiation on three-dimensional microstructural phase and porosity evolution in uranium – 10 weight % zirconium (U-10Zr) fuels and HT-9 cladding irradiated will be discussed. Fuel cladding chemical interaction (FCCI) in this fuel system will also be discussed.
4:15 PM - ES08.13.04
Dopants Selection to Immobilize Lanthanide Fission Products in Uranium-Based Metallic Fuels
Rabi Khanal 1 , Nathan Jerred 1 2 , Indrajit Charit 1 , Michael Benson 2 , Robert Mariani 2 , Samrat Choudhury 1
1 Chemical and Materials Engineering, University of Idaho, Moscow, Idaho, United States, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractLanthanide elements (e.g., Nd, Ce, Pr, & La) produced as fission products in the uranium-based metallic fuel can chemically interact with the steel cladding, known as fuel-cladding-chemical interaction (FCCI). The FCCI leads to thinning and weakening of cladding wall and eventual rupture of the cladding if the fuel is allowed to proceed to higher burnup. One possible way to reduce the FCCI is to add dopants to the metallic fuel. Detrimental lanthanides may be arrested within the metallic fuel matrix by forming intermetallic compounds with the dopants. In this work, we will present an integrated theoretical and experimental approach to screen dopants for effective capture of lanthanides within the uranium matrix. Based on ab-initio calculated electronic structures and binding energies between the dopant and the lanthanide embedded within the U-matrix, it is concluded that arsenic and selenium can serve as potential dopants to effectively arrest the lanthanides within the uranium matrix. Similarly, our calculations also confirm previous experimental observations that Pd as a dopant can effectively immobilize Nd within the uranium matrix. Finally, we will present experimentally observed microstructure and phase stability of As or Se doped U-metallic fuel in presence of lanthanide like Nd to assess the effectiveness of these dopants in immobilizing Nd within the uranium matrix. This work is being supported by the U.S. DOE's Nuclear Energy University Programs (NEUP) under contract DE-NE0008557.
4:30 PM - ES08.13.05
Thermophysical Properties of Metallic U and Alloys—A First Principles Study
Jianguo Yu 1 , Yongfeng Zhang 1 , Jason Hales 1
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractA fundamental understanding of the phonon and thermophysical properties of metal U and alloys is essential to predict fuel performance and design new fuel types with enhanced accident tolerance. In spite of being important, it is very challenge and non-trivial to use first principles calculations to predict phonon transport and interactions and resulting thermodynamic behaviors in U and alloys, stemming from the interplay of chemical bonding, strong electron correlation and relativistic effects of 5f valence electrons. In this work, we present results of a density-functional theory study of the phonon and thermophysical properties from metallic uranium (i.e., α, β, and γ-U) to U3Si2, and compare with results of recent experimental studies. Phonon properties will include the phonon density of states and phonon dispersion curves. Thermophysical properties will cover thermal expansion coefficient, free energy, entropy and specific heat capacity. We will also discuss the physical situations in which the effective direct Coulomb interactions are needed, not needed, and where they are actually not sufficient.
4:45 PM - ES08.13.06
Thermal Conductivity of Uranium
Fei Lin 1 , Eric Tea 1 , Manuel Umanzor 2 , Ryan Jacobs 3 , Shuxiang Zhou 3 , Dane Morgan 3 , Celine Hin 2 1
1 ME, Virginia Polytechnic Institute and State University, Blacksburg, Virginia, United States, 2 MSE, Virginia Polytechnic Institute and State University, Blacksburg, Virginia, United States, 3 MSE, University of Wisconsin–Madison, Madison, Wisconsin, United States
Show AbstractUranium and its alloys, such as U-Zr systems, are being investigated for their use as metallic fuels for fast reactors. The safety and efficiency of these fuels depend notably on their swelling characteristics and thermal conductivity. However, modeling fundamental properties of actinide materials is challenging due to the sensitivity on the partial occupation of the 5f-orbitals. This has been illustrated by recent debates on the validity of different implementations of Density Functional Theory (all electron versus pseudopotential) and different exchange and correlation approximations (LDA, GGA, DFT+U) for Uranium. Moreover, the validation of modelling is not straightforward due to the scarcity of transport experiments and the difficulty in obtaining a Uranium single crystal. This particularly impacts thermodynamic, mechanical and thermal properties. In this study, we report the development of a Projector Augmented Wave (PAW) pseudopotential for Uranium used for the calculation of cohesive energies, volume per atom, bulk modulus, magnetic moments, and thermal conductivity of uranium and its alloys.
Symposium Organizers
Karl Whittle, University of Liverpool
Felix Brandt, Forschungszentrum Juelich
Philip D Edmondson, Oak Ridge National Laboratory
Blas Uberuaga, Los Alamos National Laboratory
ES08.14: Fuel and Cladding II
Session Chairs
Philip D Edmondson
Karl Whittle
Friday AM, December 01, 2017
Hynes, Level 2, Room 206
8:30 AM - ES08.14.01
Simulation-Based Benchmarking of X-Ray Diffraction Line Profile Analysis of Irradiation-Induced Dislocation Loops
Rory Hulse 1 , Tamas Ungar 1 2 , Michael Preuss 1 , Chris Race 1
1 School of Materials, University of Manchester, Manchester United Kingdom, 2 Department of Materials Physics, Eötvös University Budapest, Budapest Hungary
Show AbstractIrradiation-induced growth (IIG) of Zr-alloy nuclear fuel cladding can limit the service life of nuclear fuel. It involves a macroscopic shape change driven by the formation and growth of populations of dislocation loops.
Efforts to create Zr-alloys that are resistant to IIG thus rely on accurate determination of the size distribution and number density of dislocation loops in irradiated candidate alloys. X-ray diffraction (XRD) can, in principle, provide this information via an analysis of changes to the diffraction peak shapes. Such methods are well developed in the study of plastic deformation, but the different character of the defects formed under irradiation complicates the analysis. An improved understanding of the effect of dislocation loops on the diffraction peak shapes is therefore required.
We have constructed atomistic models of controlled defect populations in Zr and generated theoretical XRD profiles. We have compared these with experimental profiles and analyzed changes in lineshape in terms of contributions from the strain fields of individual defects. In particular, we are able to explain the appearance of features in the experimental peaks which are peculiar to irradiated material.
8:45 AM - ES08.14.02
Comparison of Microstructures and Microchemistry in Neutron and Ion Irradiated Zr-1.0Nb-0.1Fe Cladding Alloys
Jing Hu 1 , Sergio Lozano-Perez 2 , Chris Grovenor 2 , Meimei Li 1 , Mark Kirk 1
1 , Argonne National Laboratory, Lemont, Illinois, United States, 2 , University of Oxford, Oxford United Kingdom
Show AbstractCladding alloys are the first safety barrier in a nuclear reactor, which requires good irradiation resistance at reactor operating and accidental temperatures. A detailed study using a range of advanced microscopy techniques has been done on recrystallized Zr-1.0Nb-0.1Fe alloys under neutron and ion irradiation conditions. Neutron irradiated samples were received after 540 days and 5 dpa exposure in Vogtle reactor. In situ ion irradiation experiment on the autoclave oxidized sample was carried out using 1 MeV Kr ionsat 320°C up to 5 dpa at the Intermediate Voltage Electron Microscope-Tandem facility at the Argonne National Laboratory. Neutron irradiation seems to have little effect on promoting fast oxidation or dissolution of β-Nb precipitates, but encourages dissolution of Fe from Zr-Fe-Nb precipitates. Radiation introduced precipitates in the metal matrix were observed after neutron irradiation, but no effect on precipitates chemistry was observed under ion irradiation. In situ ion irradiation did not introduce additional visible defects in the oxide, the oxide grains remained well-aligned columnar structure, while oxide grown under neutron irradiation showed a more complicated oxide grain structure, with shorter and less organized columnar grains. Dislocation loops were the main defects produced during neutron and ion irradiation, both
and loop grow under increasing ion irradiation dose. -loops were observed at the beginning of the irradiation and continued to grow, while -loop shows up at around 3 dpa and continues grow longer and denser through irradiation. Size distribution, interaction with alloy elements and second phase particles, defect motility are compared in detail between neutron and ion irradiated samples.9:00 AM - ES08.14.03
Microstructural Stability of Irradiated FeCrAl Alloys for Fuel Cladding Applications
Samuel Briggs 1 , Dalong Zhang 2 , Kevin Field 2 , Kenneth Littrell 2 , Sebastien Dryepondt 2 , Philip D Edmondson 2 , Khalid Hattar 1
1 , Sandia National Laboratories, Albuquerque, New Mexico, United States, 2 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractFe-Cr-Al-based alloys continue to be investigated as a potential alternative to Zr-based fuel cladding materials for enhanced accident tolerance in light water reactors (LWR) due primarily to their increased high-temperature oxidation and corrosion resistance. In these applications, precipitates are expected to play a key role in alloy performance, both through the formation of Cr-rich α′-phase precipitates and through the inclusion of oxide nanoclusters in oxide dispersion-strengthened (ODS) variants to increase mechanical strength and creep resistance. However, questions remain regarding the radiation tolerance and microstructural stability of the precipitates in this system, particularly regarding whether the extent of α′ precipitation could become application-limiting and whether the oxide nanoclusters will remain stable under irradiation. This work seeks to explore the susceptibility of 2nd generation bulk Fe-Cr-Al alloys and alloy weldments to α′ precipitation following neutron irradiation at various temperatures, as well as ODS nanocluster stability in both neutron and ion irradiation environments.
Three primary alloy compositions are considered: wrought Fe-13Cr-5Al-2Mo and Fe-13Cr-7Al-2Mo, and a powder metallurgy Fe-12Cr-5Al+Y-O ODS Fe-Cr-Al alloy variant. These materials, including select specimen weldments, have all been neutron irradiated in the High Flux Isotope Reactor (HFIR) to a nominal damage dose of approximately 2 dpa at nominal capsule average irradiation temperatures of 195, 363, and 559 °C. Small-angle neutron scattering (SANS) data was collected from bulk irradiated materials, while specimens for both atom probe tomography (APT) and transmission electron microscopy (TEM) analysis have been prepared from both bulk regions and fusion/heat-affected zones of welded regions using FIB techniques. Resulting α′ precipitate microstructures were quantified by fitting analytical models to SANS scattering intensity curves and by using standard APT cluster analysis techniques, whereas ODS nanocluster and dislocation loop microstructures were quantified using both TEM and scanning transmission electron microscopy (STEM) diffraction contrast-based imaging techniques. Following the TEM characterization of the neutron-irradiated materials, additional ion irradiation and TEM characterization was carried out in-situ at the Sandia In-situ Ion Irradiation TEM (I3TEM) facility to assess the effect of dose rate on ODS nanocluster stability. The results of this investigation will be compared to the existing literature and employed to further refine Fe-Cr-Al-based alloy design for LWR cladding applications in addition to providing additional insight regarding the fundamental challenges faced by ferritic steels and ODS alloys for deployment in nuclear environments.
9:15 AM - ES08.14.04
Advanced Composite Development for Molten Salt Nuclear Reactors
Samuel McAlpine 1 , Michael Short 1
1 , Massachusetts Institute of Technology, Cambridge, Massachusetts, United States
Show AbstractMolten salt reactors are a class of advanced nuclear reactors which could potentially improve the safety, resource utilization, and power cycle efficiency of nuclear energy. Most molten salt reactor designs use fluoride molten salts, which are known to be highly corrosive to most engineering alloys. This constitutes a significant barrier to the commercial development of molten salt reactors. In this work, we present the design, manufacture, and characterization of a novel metallic multilayer composite (MMLC) for fluoride molten salt reactor applications. This composite design has the potential to enhance the development of molten salt reactors by providing a strong, tough, radiation damage and corrosion resistant material. We expect that this composite has the potential to improve the economics of molten salt reactors as well.
The MMLC design consists of a structural layer of Incoloy 800H, an Fe-Ni-Cr alloy which has good high temperature mechanical properties and is relatively resilient to radiation damage. Crucially, Incoloy 800H is already approved for nuclear applications under the ASME boiler and pressure vessel code, which means that the regulatory basis for use of this MMLC in advanced reactor designs is relatively well-established. Overlaid on top of the Incoloy 800H is a layer of Ni-201 (commercially pure Ni) which provides a highly corrosion-resistant layer which protects the Incoloy 800H from the aggressive molten fluoride environment.
In this work, we will discuss the fabrication of laboratory-scale samples of the MMLC design using the hot isostatic pressing (HIP) method of diffusion bonding. Microstructural characterization was performed on these laboratory scale samples using scanning electron microscopy (SEM) and energy-dispersive x-ray spectroscopy (EDX). Results indicate that a high-quality metallurgical bond was achieved between the two layers. No intermetallic precipitates were observed in the interfacial region, but some titanium and aluminum carbides did form. Titanium and aluminum are found as minor alloying additions in Incoloy 800H and as impurities in Ni-201.
After the successful initial fabrication of these laboratory-scale samples, we produced a tube of the MMLC via the weld overlay technique in order to demonstrate successful production of the MMLC using a more industrially-scalable technique. We will conduct microstructural analysis of the industrial-scale composite using both optical and electron microscopy, as well as measure the relevant concentration profiles of the various alloy constituents across the interface using EDX. Furthermore, corrosion tests will be performed in high temperature FLiNaK (LiF-NaF-KF eutectic molten salt) with the underlying Incoloy 800H exposed to the salt. The purpose of these experiments is to investigate the acceleration of corrosion of the underlying Incoloy 800H by galvanic corrosion should a breach in the protective Ni-201 layer occur.
9:30 AM - ES08.14.06
Quantification of Irradiation Damage by Surface Acoustic Waves via MD and FEM Simulation
Peijun Yu 1
1 , City University of Hong Kong, Hong Kong Hong Kong
Show AbstractIn the past decades, with the rapid evolution and development of laser and ultrasonic technology, laser induced acoustic wave, as a detection method of near surface defects of materials based on the application of interdisciplinary subjects, has the abilities to generate multiple patterns of waves such as surface acoustic waves, longitudinal waves, latitudinal waves, etc. in materials synchronously. And the ultrasonic signals received experimentally are typically with high resolution and the characteristics of the signals depend on the properties and defects of materials sensitively.
Laser-induced surface acoustic wave propagation along the surface and near-surface of materials have great advantages in detection and characterization of defects on and near the surface of materials based on the features of studied wave signals in different aspects. In out study, two features are studied to build relationship with the properties and defects of objective materials. First is the laser-induced surface acoustic wave dispersion due to various material properties such as Young’s modulus, Poisson’s ratio, thermal conductivity, thermal expansion, specific heat capacity, etc. Second is the interaction between the acoustic waves and the defects on or near the surface of materials during propagation. With the analysis of the scattered or transmitted propagation wave signals in the frequency domain, we can perform quantitative detection of defects.
The joint of material properties determination from the MD (molecular dynamics) simulation of materials with point defects and finite element analysis (FEA) of the corresponding laser-induced surface acoustic waves (LSAW) is sufficiently applicable to our research object. MD simulation has the advantage in determining the property changes in irradiated (treated as point defects in this work) such as stiffness, thermal conduction, etc. while finite element method (FEM) can solve the thermo-elastic problems much faster.
ES08.15: Nanoscale Effects II
Session Chairs
Philip D Edmondson
Karl Whittle
Friday PM, December 01, 2017
Hynes, Level 2, Room 206
10:15 AM - ES08.15.01
Low Energy Deuterium Ion Interaction with Sn Films for Fusion Applications
Oluseyi Fasoranti 1 , Bruce Koel 2
1 Department of Chemistry, Princeton University, Princeton, New Jersey, United States, 2 Department of Chemical & Biological Engineering, Princeton University, Princeton, New Jersey, United States
Show AbstractSn is under consideration as a liquid-metal plasma-facing component (PFC) for high power load applications in the divertor region of fusion reactors due to potential abilities for self-recovery and heat-flux management. Furthermore, liquid Sn has lower vapor pressure and thus lower evaporative flux and higher operating temperature limits in tokamaks. Improved fundamental understanding of deuterium ion-Sn interactions that occur in Sn films on W substrates will be helpful for further evaluating the compatibility of this system for use in fusion reactors. We report on surface science experiments under UHV conditions to examine plasma-material interactions of Sn such as the thermal stability and deuterium ion uptake by Sn films on polycrystalline W substrates using surface diagnostic tools such as AES, XPS, LEIS, and TPD. Our results show that multilayer Sn films start to evaporate near 1170 K, but the Sn monolayer on W is not fully removed until 1800 K. Clustering or diffusion of Sn films was observed above 310 K. Deuterium uptake on Sn films at 310-750 K from irradiation using 700 eV D2+ ions showed lower uptake on liquid Sn films compared to solid Sn films. No deuterium uptake was seen at 750 K. Oxidation of solid and liquid Sn films by O2 was studied using XPS, with more extensive oxidation observed at higher temperatures. TPD shows Sn loss from SnO2 films at below the Sn multilayer desorption temperature. The interaction of Sn films with impurity oxygen from the W interface was also investigated, and LEIS was used to monitor the diffusion of this oxygen to the surface of the Sn film. Irradiation of these oxidized Sn films by 700 eV D2+ caused reduction of the film to metallic Sn. The reduction process is enhanced at higher temperatures.
10:30 AM - ES08.15.02
Pressure Determination in Single Helium Bubbles from Experiments and Simulations
Marie-Laure David 1 , Anne-Magali Seydoux-Guillaume 2 , Frédéric Pailloux 1 , Marie-France Beaufort 1 , Laurent Pizzagalli 1
1 , Institut Pprime, CNRS, Poitiers University, Poitiers France, 2 , LMV - CNRS - St-Etienne University, St-Etienne France
Show AbstractA by-product of fission and fusion reactions, helium interacts with confining materials in reactors or in waste storage. Since it is an inert element, it usually segregates and form bubbles, the latter being usually detrimental to mechanical properties. There have been a large number of studies dedicated to the understanding of the mechanisms of formation and evolution of these bubbles, and several models have been proposed. However, a key property is the internal pressure inside these bubbles, which remains difficult to determine.
One possible remedy is spatially-resolved Electron Energy-Loss Spectroscopy (EELS) in the Transmission Electron Microscope (TEM), which can probe single nanometer-sized helium bubbles and allow for a quantitative determination of their properties and especially the contained helium density. The potential of this technique is demonstrated for materials of interest, here silicon carbide and an uranium-thorium radioactive mineral. In the former case, the experiments are completed by molecular dynamics simulations. These investigations revealed that 3 to 6 nm in diameter bubbles, synthesized by high fluence helium implantation, exhibit surprisingly high helium densities, with corresponding pressures typically one order of magnitude higher than in metals. In the second case, helium density measurements suggest that 5 to 68 nm in diameter bubbles significantly influence swelling and fracture in the mineral.
10:45 AM - ES08.15.03
Using Space-Time Correlations to Identify Transient Defects
William Lowe 1 , Jacob Eapen 1
1 Nuclear Engineering, North Carolina State University, Raleigh, North Carolina, United States
Show AbstractThere currently exist a variety of methods for identifying defected, or disordered, atoms in a material from atomistic simulation techniques. In radiation damage applications, it is often useful to track the production, migration and recombination of point defects during the ballistic, thermal spike and ‘cooling down’ phases of a displacement cascade. The most widely-used geometric methods for recognizing defected atoms – Lindemann spheres, nearest-neighbor spheres, and Wigner-Seitz cells – make use of the periodicity of the unperturbed crystal lattice. Nonetheless, these methods must be tailored for specific applications, and even then the results can be ambiguous. For instance, using geometric methods is not physically appropriate in highly-disordered regions such as the core of a displacement cascade where translational order has been lost. In covalently-bonded systems, damage states are often quite complex, as are the recovery mechanisms. Transient point defects migrate to low-energy configurations and have been observed in simulations to hop back and forth between native sites while the material recovers from a high-energy recoil event.
We propose a statistical-mechanical methodology for characterizing the structure and dynamics of transient defects following radiation damage. The van Hove self-correlation function, Gs(r,t), or more accurately 4πr2Gs(r,t), signifies the probability of an atom being displaced by a distance r during at time t, if initially at the origin. From another perspective, Gs(r,t) describes the time-resolved motion of atoms, which is remarkably useful when analyzing the chaotic nature of displacement cascades. In particular, we utilize this time-correlation function to identify transient defects in displacement cascades in silicon carbide (SiC) from atomistic simulations. To effectively simulate a cascade event in this covalently-bonded material, we employ a hybrid Tersoff potential smoothly stitched with a Ziegler-Biersack-Littmark (ZBL) electrostatic screening function.
Following a 5keV carbon recoil in SiC, we compute Gs(r,t) separately for the carbon and silicon sub-lattices. The tail of the Gs curve for carbon atoms shows distinct peaks, indicating dynamic hopping of carbon atoms from one native site to another as the material recovers from the radiation knock. In addition, we observe a ‘shoulder’ at short distances for the carbon sub-lattice, likely representing the formation of carbon defect clusters. In contrast, the Gs tail for silicon shows a power law behavior without any noticeable peaks. This absence of peaks is strongly suggestive of small displacements among silicon atoms without forming identifiable defects, while the power law tail is indicative of a self-organized critical state approaching the crystalline-to-amorphous transition. Thus, the current methodology is able to extract a dynamical sub-lattice instability that has not been identified to this point.
11:00 AM - ES08.15.04
Eliminating Helium Bubbles in Nuclear Fuel by Facile Alumina Coating
Shenli Zhang 1 , Erick Yu 1 , Sean Gates 3 , William Cassata 3 , James Makel 2 , Andrew Thron 1 , Christopher Bartel 4 , Alan Weimer 4 , Pieter Stroeve 2 , Roland Faller 2 , Joseph Tringe 3
1 Materials Science and Engineering, University of California, Davis, Davis, California, United States, 3 , Lawrence Livermore National Lab, Livermore, California, United States, 2 Chemical Engineering, University of California, Davis, Davis, California, United States, 4 , University of Colorado Boulder, Boulder, Colorado, United States
Show AbstractHelium gas is generated as one of the main fission product in nuclear fuel rod. Due to its low solubility and low diffusivity in uranium oxide, gas bubbles with high internal pressures are formed that can deteriorate the structural integrity of the fuel. Two strategies have been considered to avoid this high-pressure scenario: either by creating void space at the time of fuel preparation, such that on-site gas retention is allowed, and/or the design of an alternative diffusion path for helium to escape the fuel rod.
Introducing a secondary material can be effective for the above strategies. Previous investigations have been focused on doping additives into the uranium oxide matrix to improve the diffusivity. Nevertheless, it will ideal if the secondary material can be synthesized and optimized independently of fuel chemistry. Direct coating of the secondary material around nuclear fuel particle using atomic layer deposition (ALD) can be promising. Alumina, which is compatible with this technique and the corresponding nuclear environment, is of great interest to be first examined for effective helium transport.
In this work, a new method of employing ALD to realize homogeneous alumina coating around nuclear fuel particles is proposed to facilitate helium transport. This conception is realized on nickel particles as surrogates for uranium oxide, because both materials are similar in low helium solubility.
The effectiveness of alumina as retention phase has been confirmed from both experimental and simulation perspectives. Samples with different alumina layer thickness were made for comparison in experiments. Helium spectroscopy revealed the helium retention capacity of alumina-coated nickel particles could be 2 orders of magnitude higher comparing to pure nickel particle.
Simulation techniques were then applied to predict alumina’s performance beyond experimental test conditions. Alumina’s retention concentration at different temperature and pressure conditions was estimated from Monte Carlo (MC) simulation, which shows the possibility of fully storage of helium in a spent fuel rod. The diffusion coefficient of helium in alumina was further calculated from molecular dynamics (MD) simulation, indicating the improvement of fractional release using alumina coating round uranium oxide particles.
Parts of this work were performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
11:15 AM - ES08.15.05
Ion Irradiation in Oxide Nanoceramics—On the role of the Irradiation Spectrum at Extreme Damage Levels
Matteo Vanazzi 1 , Francisco Garcia Ferre 1 , Alexander Mairov 2 , Luca Ceseracciu 1 , Marco Utili 3 , Marco Beghi 4 , Kumar Sridharan 2 , Fabio Di Fonzo 1
1 , Istituto Italiano di Tecnologia - IIT, Milan Italy, 2 , University of Wisconsin–Madison, Madison, Wisconsin, United States, 3 , ENEA, Bologna Italy, 4 , Politecnico di Milano, Milan Italy
Show AbstractIn the framework of future generation nuclear reactors, structural materials will face environmental conditions even more challenging: higher radiation damage as well as the presence of extremely corrosive media. To deal with these problems oxide nanoceramics have been proposed. Oxide nanoceramics combine the enhanced radiation tolerance of nanocrystalline materials with the chemical inertness of oxides. Following our previous studies in which we demonstrated the stability of nanoceramic alumina (Al2O3) coatings in typical LFRs conditions, in this work we perform further characterization and analyses. The properties of Al2O3 films are evaluated as radiation damage approaches extreme levels, reaching and even exceeding those anticipated for advanced nuclear systems, namely from 150 to 450 displacements per atom (DPA). New irradiation tests are performed using different ions in order to investigate the effect of the irradiation spectrum on the material’s evolution. A comprehensive analysis of the irradiated samples is accomplished by X-Ray Diffractometry (XRD), Transmission Electron Microscopy (TEM), Scanning-TEM (STEM) and nanoindentation. The results show a general grain growth as the main structural change induced by irradiation in oxide nanoceramics. This structural change manifests mechanically through an initial increase of hardness (in accordance with the Hall-Petch relationship) and eventually through softening in the very last part of the experiment. Stiffness increases sub-linearly with damage before reaching a plateau. Further, both hardness and stiffness depend on the phase present. Thus, a deep effort is made to establish a correlation between irradiation spectra and the evolution of the oxide’s structural features and mechanical properties. The phase evolution appears to depend strongly on the ion utilized and on the irradiation spectrum, so the kinetic of the grain growth process. Finally, molecular dynamics simulations of displacement cascades are used to support the collected data and a preliminary model is formulated according to these observations. Last but not least, irradiated samples are exposed to static lead at 550 °C to verify the effectiveness of Al2O3 films after extreme irradiation damages, showing no corrosive attacks or degradative phenomena. To conclude, an extensive characterization campaign is performed on nanoceramic Al2O3 in order to study its behavior at extreme radiation damage levels. Structural and morphological changes are analyzed and supported by modelling. Coatings superior corrosion resistance is retained even after exposure to intense radiation environments.
11:30 AM - ES08.15.06
Interface Dependence of the Nanosize Effect on the Radiation Stability of Nanostructured Apatite
Fengyuan Lu 1 , Jianren Zhou 1 , Tiankai Yao 2 , Jie Lian 2 , Marquis Kirk 3
1 , Louisiana State University, Baton Rouge, Louisiana, United States, 2 , Rensselaer Polytechnic Institute, Troy, New York, United States, 3 , Argonne National Laboratory, Argonne, Illinois, United States
Show AbstractNanostructured materials exhibit great potentials for enhanced radiation tolerance that is crucial for advanced nuclear energy applications. However, recent studies have also shown that the nanostructures are not intrinsically radiation tolerant, and reduced grain size sometimes can lead to decreased radiation stability that have not been well explained. In this study, nanostructured hydroxyapatite samples with two different interfaces – open surfaces in nanoparticles and grain boundaries in densified nanocrystalline bulk materials, were irradiated with 1 MeV Kr2+ in order to investigate the role of interface in the radiation behavior of nanostructured materials. The radiation stability of hydroxyapatite nanoparticles showed a significant size effect in which the nanoparticles became more susceptible to the ion-irradiation-induced amorphization as the size was reduced from 80 nm to 20 nm. In contrast, samples with dense grain boundaries exhibit a significantly higher radiation tolerance than the nanoparticles with the same size, as the result of a lower interface energy at the grain boundaries as compared to open surfaces. This result highlights that the interface energy can significantly reduce the radiation stability of apatite materials, which is a critical factor in the design of radiation-tolerant nanostructured materials under intense radiation conditions.