Symposium Organizers
Dilpuneet Aidhy, University of Wyoming
Kazuto Arakawa, Shimane University
Estelle Meslin, CEA Saclay
Haixuan Xu, University of Tennessee
ES5.1: Metallic Systems I
Session Chairs
Kazuto Arakawa
Charlotte Becquart
Tuesday PM, April 18, 2017
PCC North, 200 Level, Room 223
11:30 AM - *ES5.1.01
Deferred Achievement of a Steady State Microstructure in Radiation-Resistant Alloys
Gary Was 1 , Elizabeth Getto 2 , Zhijie Jiao 1 , Anthony Monterrosa 1
1 , University of Michigan, Ann Arbor, Michigan, United States, 2 , U.S. Naval Academy, Annapolis, Maryland, United States
Show AbstractReactor core materials in both fast reactors and LWRs granted life extension must withstand irradiation to high dpa at high temperature. Ferritic-martensitic (F-M) alloys are attractive candidates for structural components in both fast and thermal reactors that are expected to reach high damage levels. Thus, understanding microstructure evolution of structural materials at high damage levels is critical to the application of these alloys. For relatively radiation resistant materials, microstructures continue to evolve at very high damage levels such that a steady state microstructure may not be reached within the practical lifetime of the reactor component. As such, extrapolation of microstructure evolution from modest to high dpa is invalid. Microstructure evolution at high dpa can be probed using self-ion irradiation with simultaneous He injection at appropriate temperatures. Computational models for defect cluster evolution are also being developed and benchmarked against experimental data to ultimately provide predictive capability for the response of both microstructure (loops, voids, precipitates, etc.), and mechanical properties (hardening, ductility, slip behavior) to irradiation. The dislocation microstructure, precipitate behavior and cavity growth are tracked both experimentally and computationally and are shown to continue their evolution up through 650 dpa in alloy HT9 irradiated with 5 MeV Fe++ ions at 460°C. It is the interplay between these evolving features that delays and may ultimately prevent achievement of a true steady state microstructure.
12:00 PM - ES5.1.02
Microstructural and Micro-Mechanical Changes in Tungsten under High Flux Plasma Exposure
Dmitry Terentyev 1
1 , SCKCEN, Mol Belgium
Show AbstractRecent theoretical and subsequent experimental studies suggest that the uptake and release of deuterium (D) in tungsten (W) under high flux plasma exposure (i.e. under ITER-relevant conditions) is controlled by dislocation microstructure induced by the plasma itself. A comprehensive mechanism for the nucleation and growth of D bubbles on dislocation network under high flux low-energy plasma exposure was proposed and validated. The process of bubble nucleation can be described as D atom trapping at a dislocation line, its in-core migration, the coalescence of several D atoms into a multiple cluster, which eventually transforms into a nano-bubble by punching out matrix atoms on the dislocation line. This view implies that the initial microstructure might be crucial for D uptake and degradation of the sub-surface layer under prolonged plasma exposure. Understanding of the role played by the initial microstructure is the purpose of this work.
In this work, we apply several experimental techniques to investigate the microstructure and mechanical properties of surface and sub-surface layer of W exposed to the high flux plasme. In particular, we use transmission and scanning electron microscopy, as well as nano-indentation measurements. To reveal the impact of the initial microstructure, we have performed exposures in single crystal, poly-crystal and heavily deformed polycrystal tungsten samples. The preliminary TEM study demonstrates that even in single crystal sample, high flux plasma exposure induces high density of dislocations and tangles in the sub-surface area. The presence of the plasma-induced microstructure is well detected by the nano-indentation experiments, which provide reach information about change of material hardness and depth distribution of the irradiation-induced microstructure.
12:15 PM - ES5.1.03
High Temperature Defect Migration Mechanism in Ni and Ni-Based Concentrated Solid-Solution Alloys
Taini Yang 1 , Chenyang Lu 1 , Ke Jin 2 , Hongbin Bei 2 , Yanwen Zhang 2 , Lumin Wang 1
1 , University of Michigan, Ann Arbor, Michigan, United States, 2 , Oak Ridge National Lab, Knoxville, Tennessee, United States
Show AbstractIn this study, pure Ni, single-phase concentrated solid solution alloys (SP-CSAs) including binary (NiCo and NiFe), ternary (NiCoFe and NiCoCr) to quintal (NiCoFeCr) and high entropy alloy (HEA) NiCoFeCrMn have been employed to study defect dynamics by heavy ion irradiation. The irradiation was conducted at the Ion Beam Materials Lab (IBML) at University of Tennessee, using 1.5 or 3.0 MeV Ni+ ions at 500oC to three different ion fluences: 3.0*1015, 1.5*1016, and 5.0*1016. Cross-sectional transmission electron microscopy (TEM) characterization has been conducted to study the defect distribution and migration along the depth from the sample surface. The damage depth, defect cluster size and distributions in alloys of different composition are compared. The preliminary result can be explained with defect migration mechanisms from a long-rage one-dimensional or a short-range three-dimensional mode incorporating with the injected interstitial effect.
This work was supported by Energy Dissipation to Defect Evolution (EDDE), an Energy Research Frontier Center supported by the U.S. Department of Energy, Basic Energy Sciences.
12:30 PM - ES5.1.04
Defect Formation in Helium Irradiated Y2O3 Doped W-Ti Alloys Studied by Positron Annihilation and Nanoindentation
Asta Richter 1 , Wolfgang Anwand 2 , Chun-Liang Chen 3 , Roman Boettger 2
1 Engineering Physics, Technical University of Applied Sciences Wildau, Wildau Germany, 2 , Helmholtz Center Dresden-Rossendorf (HZDR), Dresden Germany, 3 Material Science and Engineering, National Dong Hwa University, Hualien Taiwan
Show AbstractTungsten-titanium alloys are considered as promising materials for future fusion devices, in particular for the divertor and other first wall components. The microstructure and the mechanical properties of the material are dependent on the amount of Ti present in the alloy. In particular, titanium reduces the brittleness of the tungsten alloy, which is manufactured by mechanical alloying. The addition of Y2O3 nanoparticles increases the mechanical properties at elevated temperature and enhances irradiation resistance.
Helium ion implantation was used to simulate irradiation effects on materials considered for applications in fusion reactors. The irradiation was performed using the 500 kV He ion implanter at fluences of 2.5 x 1015 cm-2 for sample series both at room temperature and elevated temperature of 600 °C, respectively. SRIM calculations result in a maximum helium concentration at about 650 nm.
The microstructure and mechanical properties of the pristine and irradiated W-Ti-ODS alloy are compared with respect to the titanium content. Radiation damage is studied by positron annihilation spectroscopy analyzing the life time and the Doppler broadening. Two types of helium-vacancy defects were detected after helium irradiation in the W-Ti-ODS alloy, small defects with high helium-to-vacancy ratio (low S parameter) for room temperature irradiation and large defects with low helium-to-vacancy ratio (high S parameter) for thermally treated tungsten-titanium alloys.
Defect induced hardness has been studied by nanoindentation. A drastic hardness increase is observed after He ion irradiation both for room temperature and elevated irradiation temperature of 600°C. The Ti alloyed tungsten-ODS is stronger affected by hardness increase after irradiation compared to the pure W-ODS alloy. Vacancy-type defects created by helium implantation in W-Ti-ODS alloys and their impact on nanohardness characteristics are correlated with the investigations of microstrucral changes analyzed by positron annihilation.
12:45 PM - ES5.1.05
Exploring the Stability of TPBAR Liner with In Situ, Triple-Beam, Ion Irradiation TEM
Brittany Muntifering 1 , Clark Snow 1 , David Senor 2 , Khalid Hattar 1
1 , Sandia National Labs, Albuquerque, New Mexico, United States, 2 , Pacific Northwest National Lab, Richland, Washington, United States
Show AbstractPredicting the life expectancy and operation condition of nuclear materials is difficult due to the overlapping and often synergistic interactions from the neutron flux, elevated temperatures, and corrosive elements. This is no more apparent than in the liner of a tritium-producing burnable absorber rod (TPBAR). The TPBAR is a viable method for large scale tritium production using commercial nuclear reactors. The large scale production of tritium is important not only for current niche tritium applications, but also the various fusion based technologies being explored.
To predict tritium and helium location, as well as the rod lifetime, greater insight is needed into the physics of TPBAR liner evolution in the combination of extreme environments. This presentation will demonstrate the scientific benefit of a new experimental technique combining elevated temperature, triple-beam, in-situ, ion irradiation transmission electron microscopy (I3TEM) to better understand the stability of the internal microstructure. In this study, we utilized this newly developed accelerated aging technique that permits with nanometer resolution the direct real time exposure of a sample held at elevated temperature to heavy ion irradiation, while simultaneously implanting 10 keV He and D2. In order to predict the long term stability, we in the time frame of a few hours exposed a Zircaloy-4 TEM sample to the controlled ratios of displacements per atom, He, and H isotopes that is expected in a TPBAR liner after 500 days without activating the sample. In addition, we performed a similar in-situ experiment over a three-day period to explore the liner’s response well past the expected lifetime. During these initial exploratory accelerated aging tests, the liner appeared extremely stable. Finally, concerns and limitations of utilizing these highly accelerated small scale results in predicting the operational response of TPBAR liners in reactors will be discussed.
Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.
ES5.2: Metallic Systems II
Session Chairs
Cristelle Pareige
Tomoaki Suzudo
Tuesday PM, April 18, 2017
PCC North, 200 Level, Room 223
2:30 PM - *ES5.2.01
Point Defect Clusters and Loops in Fe—Interaction with Solute Atoms and Insights from DFT Calculations
Charlotte Becquart 1 , Christophe Domain 2
1 , Univ de Lille 1, Villeneuve D Ascq France, 2 , EDF R & D, Moret sur Loing France
Show AbstractThe ageing and the evolution of mechanical properties of pressure vessel steels under radiation have been correlated with the formation of more or less dilute solute clusters. Tomographic atom probe analysis show that these clusters are mainly enriched in Cu, Ni, Mn, Si and P. The formation of these features is governed by the migration and clustering of point defects and atomistic simulation can provide insights on the diffusion and agglomeration of both point defects and solute atoms. In this presentation, Density Functional Theory calculation results on the interaction of substitutional solute atoms as well as carbon atoms with different extended defects (small vacancy and interstitial clusters, as well as interstitial loops) will be examined, and discussed. The role of magnetism on the properties will be examined and the validity of empirical potentials for such calculations assessed. Finally, we will examine the impact of these results on the evolution of the microstructure under irradiation modelled by Atomic / Object Kinetic Monte Carlo.
3:00 PM - ES5.2.02
Experimental Measurement of the 1D Diffusivity of <100> Dislocation Loops in Iron
Kazuto Arakawa 1
1 , Shimane University, Matsue Japan
Show AbstractNuclear-fission and fusion materials are degraded primarily due to the accumulation of radiation-produced lattice defects-point defects (self-interstitial atoms (SIAs) and vacancies) and their clusters. To precisely predict the lifetime of these materials, accurate understanding of the structures and dynamic properties of these defects is required.
Typical SIA clusters in bcc iron-based materials are nanoscale prismatic dislocation loops with the Burgers vector, b, of 1/2<111> and <100>. It has been revealed that 1/2<111> loops undergo 1D glide diffusion in their b direction, by MD simulations and in-situ TEM experiments [1]. Also, the 1D diffusivity of the 1/2<111> loops in high-purity Fe has been experimentally measured [1, 2]. However, there have been no consensus on the <100>-loop diffusion, although <100> loops are dominant at comparatively high temperatures such as those for the reactor operation.
In this presentation, we will present the results for the in-situ TEM observation of <100>-loop diffusion. It will be shown that the <100>-loop diffusion is categorized into two-temperature regimes, and the measured activation energies for the <100>-loop diffusion will be provided.
References
[1] K. Arakawa et al. Science, 318 (2007) 956.
[2] K. Arakawa et al. ISIJ International, 54 (2014) 2421.
3:15 PM - ES5.2.03
OKMC Simulation of Absorption Kinetics of Point Defects by Dislocations and Defect Clusters
Denise Carpentier 1 , Thomas Jourdan 1 , Yann Le Bouar 2 , Mihai-Cosmin Marinica 1
1 , CEA Saclay, Gif-sur-Yvette France, 2 , LEM, CNRS/ONERA, Châtillon France
Show AbstractThe behaviour of materials under irradiation crucially depends on the fluxes of point defects to the different sinks of the microstructure, such as dislocations, cavities and loops. For example, void swelling is generally explained by a net flux of interstitials to dislocations and dislocation loops due to the stronger elastic interactions of these sinks with interstitials than with vacancies. A quantitative assessment of the sink strengths is therefore important to develop a predictive model of microstructure evolution. However, up to now their values are still badly known.
This work aims at performing a precise evaluation of sink strength values by Object Kinetic Monte-Carlo (OKMC) simulations. The elastic interactions between sinks and point defects are taken into account and the atomistic migration mechanisms are explicitly modelled. Defects are described by their elastic dipoles at stable and saddle points, which are computed by ab initio calculations.
The case of a straight dislocation in aluminum is first studied. The results show that the elastic interactions strongly affect the sink strengths. In particular, the anisotropy of point defects at saddle points is identified as a key parameter in sink strength calculations. Similar conclusions are drawn for the other types of sinks that are investigated (cavities and loops). More generally, considering the atomistic migration mechanisms with full account of interactions at stable and saddle points appears essential to derive reliable sink strengths and thus to perform predictive kinetic simulations.
3:30 PM - ES5.2.04
Microstructure and Mechanical Property Evaluation of Oxide Dispersion Strengthened Ferritic Steels
Wahida Ilaham 1 , Tapas Laha 1 , S.K Pabi 1
1 , Indian Institute of Technology, Kharagpur, Kharagpur India
Show AbstractNanostructured oxide dispersion strengthened (ODS) ferritic steels with two different nominal compositions of Fe-14Cr-2W-0.3Ti-0.3Y2O3 (alloy A) and Fe-14Cr-2W-0.3Ti-0.3Y2O3-1Si (alloy B), (wt. %) were synthesized by mechanical alloying for 30 h and consolidated by Spark plasma sintering at 900, 1000, and 1050 oC with sintering time of 5 min at 60 MPa pressure. The effect of Si on the microstructure, as well as on the mechanical properties was investigated and compared by combined study of X-ray diffraction, SEM, TEM, EDS analysis. No major changes in crystallite size (~18 nm), lattice strain (~ 0.96%) and dislocation density (~ 7×1015 m-2) have been observed in both the alloys during mechanical alloying up to 30 h. The relative density, Compressive strength and yield strength of alloy A were only around 81%, 1006.4 MPa and 840.3 MPa, while that of alloy B could reach up to 97.3%, 2226.3 MPa and 1576.6 MPa after sintering at 900 oC. Mechanical properties of the alloy B were ~2 times higher than the alloy A. EDS analysis confirmed that alloy B has a unique microstructure with darker contrast phase in comparison to alloy A, which is rich in Fe-Si-O and were homogenously distributed in the matrix. The considerable improvement in the mechanical properties of the alloy B can be attributed to higher density, typical microstructure and uniform dispersion of Y-Ti-O (20-40 nm) rich spherical precipitates.
Keywords: Mechanical alloying; Spark plasma sintering; Oxide dispersion strengthened steels; microstructure; Mechanical properties.
ES5.3: Metallic Systems III
Session Chairs
Tuesday PM, April 18, 2017
PCC North, 200 Level, Room 223
4:15 PM - *ES5.3.01
On the Correlation between Damage and Ion Injected Profile and Clustering in FeCr(NiSiP) Alloys
Cristelle Pareige 1 , Begonia Gomez-Ferrer 1 , Olivier Tissot 2 1 , Brigitte Decamps 3 , Estelle Meslin 2 , Jean Henry 2
1 , University of Rouen, Saint etienne du rouvray France, 2 , CEA, Saclay France, 3 IN2P3, CNRS, Orsay France
Show AbstractFeCr alloys are model alloys of high-Cr Ferritic-Martensitic steels that are candidates for structural alloys for the future GEN IV and fusion reactors. Their in-service behaviour is thus a key issue. Nevertheless, neutron irradiations are expensive and access to facilities is restricted. The community thus focuses on model irradiation experiments using alternative irradiation sources. However, transferability issues arise. Given the widespread use of ion-irradiation as a surrogate for neutron irradiation, understanding of the origin of differences and commonalities is of high importance.
Depending on the particles used (ions, electrons or neutrons) to irradiate Fe-Cr alloys, at the same dose and temperature, a' precipitation may occur or not [1–4]. It is not the case for impurity clusters (enriched in Ni, Si, P and Cr) that were observed in all investigated FeCr alloys of low purity whatever the irradiating particles used [1,5,6].
Among the parameters which are specific to ion irradiation, the influence of injected ions on α' and impurity cluster formation has never been considered up to now.
In order to gain further insight into the role of these injected interstitials on the precipitation and segregation kinetics in FeCr alloys under ion irradiation, Fe irradiations was performed on a high purity Fe15Cr alloy and on Fe14CrNiSiP alloy within the European FP7/MatISSE project. Atom Probe Tomography was used to investigate the influence of both damage and injected Fe profiles on α/α' decomposition and impurity clusters formation [7].
This paper shows that characterisation of the microstructure with depth enables to explain the absence of α' precipitation under ion irradiation reported in previous studies. It also reports on the influence of injected interstitials and damage profile on the formation of the NiSiPCr-enriched clusters.
[1] V. Kuksenko et al. J. Nucl. Mater. 432 (2013) 160–165..
[2] M. Bachhav et al. Scr. Mater. 74 (2014) 48–51..
[3] F. Bergner et al. Scr. Mater. 61 (2009) 1060–1063.
[4] O. Tissot,et al. Scr. Mater. 122 (2016) 31–35.
[5] C. Pareige et al. J. Nucl. Mater. 456 (2015) 471–476..
[6] O. Tissot, PhD thesis, CEA and University of Rouen, France, 2016.
[7] O. Tissot et al. Mater. Res. Lett. 0 (2016) 1–7
4:45 PM - *ES5.3.02
How Rhenium and Osmium in Tungsten Crystals Suppress Radiation-Induced Defects
Tomoaki Suzudo 1 , Akira Hasegawa 2
1 , JAEA, Tokai-mura Japan, 2 , Tohoku University, Sendai Japan
Show AbstractTungsten (W) is a promising candidate of plasma-facing materials (PFM) in fusion reactors because it has high melting temperature, high resistance to sputtering, and high thermal conductivity. However, radiation-induced vacancies and their clusters generated in W materials seem to trap hydrogen isotopes such as deuterium, and the feasibility of PFM made of W materials is still questionable because such retention may seriously deteriorate safety and efficiency of the fusion devices. It is also known that inclusion of rhenium (Re) and osmium (Os) into W crystals suppresses swelling and void growth; such experiments suggest that Re or Os addition makes the material radiation-resistant. In addition, a recent study made another discovery that inclusion of Re reduces deuterium retention associated with the suppression of surviving vacancies. This implies that Re and Os enhance vacancy-interstitial recombination that probably causes the suppression of void swelling. In the current study, we apply the density functional theory combined with atomic kinetic Monte Carlo method to indicate a mechanism to mitigate the effect of radiation on W crystals by adding particular solute elements that change the migration property of interstitials. The modeling study suggests that, when Re and Os atoms are present, most interstitials produced by radiation displacement form 3D-migrating mixed-dumbbells, which enhance vacancy-interstitial recombination probability and decrease the total number of surviving radiation-induced point defects. We also make comprehensive analyses of related experimental database, which support the above claim. The resultant mechanism is applicable to any body-centered-cubic (BCC) metals whose self-interstitial atoms become a stable crowdion and is expected to provide a guideline for computational design of radiation-resistant alloys in the field of nuclear applications.
5:15 PM - ES5.3.03
Integrated Computational and Experimental Study of Radiation Damage Effects in Alloy 709
Haixuan Xu 1 , Zizhe Lu 1 , Luis Casillas 1 , Li He 2 , Kumar Sridharan 2 , Tianyi Chen 3 , Lizhen Tan 3
1 , University of Tennessee, Knoxville, Tennessee, United States, 2 , University of Wisconsin, Madison, Wisconsin, United States, 3 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractDisplacement cascade simulations have been carried out in Alloy 709, a candidate alloy being developed for the Sodium-cooled Fast Reactor (SFR). Specifically, we focus on the influence of dislocations on the cascade processes and subsequent defect evolutions. The resulting defect density and size distributions are characterized and compared with a reference austenitic steel 316H. In addition, the radiation-induced mechanical property changes have been investigated using atomistic nanoindentation simulations and tensile test simulations. The simulation results are directly compared with ion irradiation experiments and corresponding mechanical tests. Through these synergistic efforts, this study provides initial evaluation of radiation effects on microstructural evolution and mechanical properties of this strategic alloy.
5:30 PM - ES5.3.04
Investigation of Creep-Fatigue Damage in Alloy 617 at High Temperatures
Fraaz Tahir 1 , Yongming Liu 1
1 , Arizona State University, Tempe, Arizona, United States
Show AbstractAlloy 617, a solid solution strengthened nickel base superalloy, is a primary candidate material for the intermediate heat exchanger tubing of the next generation of high-temperature gas-cooled reactors. ASME design codes for nuclear components recommend linear damage summation and time fraction rule for life-prediction. However, the creep-fatigue interaction in this alloy cannot be modelled accurately by this method, particularly at temperatures higher than 850°C. Development of improved models requires test data in the creep-dominant regime. Traditional strain-controlled loading waveforms, which cause stress relaxation during the hold period, show a saturation in cycle life with increasing hold periods due to the rapid stress-relaxation of Alloy 617 at high temperatures. Therefore, longer hold time tests cannot provide creep-dominated failure. In this study, loading waveforms with force-controlled hold periods are used to produce creep-dominated interaction at 850 and 950°C. Analysis of micro-scale damage features, such as fatigue cracks and creep voids, is then used to find a correlation between creep and fatigue damage. A phenomenological life-prediction model that agrees with the micro-scale damage observations is proposed as an alternative to the time fraction rule.
5:45 PM - ES5.3.05
Ion Irradiation Defects in Austenitic Alloy 709 and Ferritic-Martensitic Steel Grade 92 for Nuclear Applications
Li He 1 , Rigen Mo 1 , Beata Tyburska-Pueschel 1 , Haixuan Xu 2 , Tianyi Chen 3 , Lizhen Tan 3 , Kumar Sridharan 1
1 Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin, United States, 2 Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee, United States, 3 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractAustenitic alloy 709 and ferritic-martensitic steel grade 92 are being considered as candidate advanced structural materials for fast reactors and other Gen IV reactor concepts. However, there are relatively few studies on irradiation damage effects in these materials. We have used proton irradiation up to 1.5 dpa (flat damage vs. depth regime at 10 μm depth) at 670 °C and Fe2+ ion irradiation up to 100 dpa at 360 °C (peak damage at 1 μm depth) to study the irradiation defects in these materials. Conventional alloys 316H and grade 91 were also irradiated under the same conditions as base-line comparisons for alloys 709 and G92, respectively. Transmission electron microscopy (TEM) study showed that the main defect type is Frank dislocation loops in alloy 709 under Fe2+ irradiation at 100 dpa, 360 °C. Fewer dislocation loops were observed in alloy 709 than in 316H as a consequence of Fe2+ irradiation. Precipitates of Nb(CN) (rocksalt type structure) in 709 appear to grow larger and decrease in number density under Fe2+ irradiation. Fe2+ irradiation also induced nickel segregation and chromium depletion at high-angle grain boundaries in 709. Cracks were observed in 1.5 dpa 670 °C proton-irradiated 709 along surface grain boundaries.
A new TEM method has been developed to measure the swelling of materials, using energy loss of high energy electrons in interaction with valence electrons. No appreciable swelling was detected in alloy 709 after Fe2+ irradiation.
Symposium Organizers
Dilpuneet Aidhy, University of Wyoming
Kazuto Arakawa, Shimane University
Estelle Meslin, CEA Saclay
Haixuan Xu, University of Tennessee
ES5.4: Complex Behaviors in Ceramics I
Session Chairs
Haixuan Xu
Yongfeng Zhang
Wednesday AM, April 19, 2017
PCC North, 200 Level, Room 223
9:30 AM - *ES5.4.02
On the Relationship between Chemical Disordering and Mass Transport in Complex Oxides
Blas Uberuaga 1 , Romain Perriot 1 , Richard Zamora 1 , Danny Perez 1 , Arthur Voter 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractComplex oxides, such as pyrochlores and spinels, have been proposed for use in various nuclear energy applications, including nuclear waste forms and inert matrix fuels. A key criterion for the performance of these materials is mass transport, as it dictates the response of the material to irradiation, including whether the material amorphizes or the rate at which extended defects such as interstitial loops form. However, irradiation also induces chemical disorder, in which cations of different types begin to mix across their own sublattices, creating a complex and very heterogeneous energy landscape for defects, severely complicating the ability to study these systems at the atomic scale. In particular, while various efforts have examined anion transport in disordered complex oxides, very few have studied cation transport.
Here, using a combination of long-time scale simulation techniques in the class of the accelerated molecular dynamics methods, we interrogate how cation transport depends on cation disorder in pyrochlore. We find that the relationship is rather complex, involving a competition between an enhancement of cation diffusion with disorder and a corresponding enhancement in the healing of that disorder due to the migration of cation defects. We find that the migration of cations in disordered pyrochlore is caused by a percolation effect in which a critical level of disorder is needed to enhance cation migration. We discuss the implications of these results for the performance of these materials under irradiation.
10:00 AM - ES5.4.03
Radiation Damage in Borosilicate and Iron Phosphate Glass
Roger Smith 1 , Kenny Jolley 1 , Kitheri Joseph 2
1 , Loughborough University, Leicestershire United Kingdom, 2 , Indira Gandhi Centre for Atomic Research, Tamil Nadu, Kalpakkam, India
Show AbstractBorosilicate and iron phosphate glass are candidate materials for the containment of high level nuclear waste over long time scales. However there are few good models of these materials. The talk will introduce some fixed charge classical potentials suitable for radiation damage modelling and then, using molecular dynamics, compare their relative stability when collision cascades induced by alpha particle recoils occur.
Since there is no concept of an interstitial in a glass, radiation damage will be analysed in terms of the atoms that are displaced and by the changes in co-ordination number. It will be shown that for the borosilicate glass, over and underco-ordinated regions have high energy barriers to reconstruct indicating that local damage can be frozen in for relatively long periods of time compared to metals. For the iron phosphate glasses, the PO4 tetrahedra rapidly reconstruct after a radiation event and preliminary results indicate that the glass with a low Fe2+ content is the most radiation resistant.
10:15 AM - ES5.4.04
Effects of Ionization in Ion-Irradiation Studies of Damage Evolution in Nuclear Ceramics
William Weber 1 2 , Eva Zarkadoula 2 , Haizhou Xue 1 , Yanwen Zhang 2 1
1 , University of Tennessee, Knoxville, Tennessee, United States, 2 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractAt ion energies typically used to study fast neutron damage, fission damage or alpha-decay damage in nuclear ceramics, the electronic and nuclear energy losses are often comparable, and local ionization along the ion path can affect damage production and evolution. Experimental and computational approaches are used to investigate the separate and combined effects of nuclear and electronic energy loss on radiation damage in ceramics relevant to nuclear applications. Experimentally, ion mass and energy control the ratio of electronic to nuclear energy loss, and large-scale atomistic simulations that combine ionization-induced thermal spike and atomic collision processes are used to model these effects. The results demonstrate that electronic energy loss, typical of MeV ions, can lead to competitive damage recovery processes or additive and synergistic effects on damage production in many nuclear-relevant ceramics. These results have significant implications for interpreting and modeling the radiation response of nuclear ceramics in accelerated testing using MeV ion irradiation. This work was supported by the U.S. DOE, BES, MSED.
10:30 AM - ES5.4.05
Atomic Resolution STEM Imaging of Epsilon Phase Formation in Doped Ceria by Irradiation and Thermal Annealing
Michele Conroy 1 , Weilin Jiang 1 , Ram Devanathan 1
1 , Pacific Northwest National Lab, Richland, Washington, United States
Show AbstractThe mechanism of formation of metallic phase fission products in nuclear fuel and their morphological evolution is still not well understood. The development of these phases within uranium oxide matrices over time in fuel under irradiation has been linked to many important phenomena over the life cycle of nuclear fuel, from in-reactor operations to long-term spent fuel disposition. In this study we focus on the irradiation and thermal annealing conditions under which the metallic phase forms in CeO2 (surrogate material for UO2). CeO2 film doped with Mo, Re, Ru, Rh and Pd was grown on a polycrystalline yttria-stabilized zirconia substrate at 550 °C by pulsed laser deposition. These films were irradiated with 90 keV He+ ions at 400°C to a fluence of 4x1017 ions/cm2 to accumulate a peak damage level of 13 displacements per atom. Scanning transmission electron microscopy (TEM) elemental mapping characterization of the films revealed that small (<3 nm) Pd enrichments/particles were formed after the irradiation, but not observed in the pre-irradiation film. After annealing the same film the other 4 metal dopants are also enriched in some areas along the defects and voids within the film. In-situ thermal annealing TEM experiments will be undertaken to watch the movement of these metals through the film in real time. In an effort to determine the nature of radiation damage that is a driving force for the phase formation, we have initiated molecular dynamics simulations of fission track damage in CeO2 and UO2. Our integrated effort using synthesis, ion irradiation, chemical imaging, and atomistic simulation is providing valuable insights into metallic phase formation in irradiated nuclear fuel surrogate.
10:45 AM - ES5.4.06
Defect Kinetics and Long-Term Evolution of Grain Boundaries in Irradiated Silicon Carbide
Hao Jiang 1 , Xing Wang 1 , Izabela Szlufarska 1
1 , University of Wisconsin Madison, Madison, Wisconsin, United States
Show AbstractWhile interfaces, such as grain boundaries, are known to act as sinks of defects in irradiated materials, much less is known about the effects of radiation-induced defects on the evolution of atomic structure and therefore also of the sink strength of grain boundaries. In particular, in SiC our rate theory models predicted an unbalanced flux of carbon interstitials to grain boundaries as compared to the flux of silicon interstitials, but it is unclear how this excess of carbon is accommodated by grain boundaries. Such knowledge would potentially enable design of interfaces with increased sink efficiencies. In this study, we combined molecular dynamics simulations, rate-theory modeling, scanning transmission electron microscopy (STEM) and electron energy loss spectroscopy (EELS) to investigate the defect kinetics in grain boundaries. More specifically, using molecular dynamics and nudged elastic band calculations, we find the diffusion of intrinsic defects at small angle tilt grain boundaries in SiC is slower as compared to bulk diffusion. This finding is contrary to many other materials where grain boundary diffusion is faster than bulk diffusion. This effect was attributed to the strain effect on the migration of dumbbell-type interstitials. Furthermore, we find a high preference for interstitials to nucleate jog pairs and get annealed on dislocation lines at the boundaries. Based on the migration and reaction barriers calculated by atomistic simulations, a rate-theory model considering the defect flux to grain boundary, the defect diffusion along grain boundary and defect-jog interactions has been developed. The model predicts that there is a transition between the roles of grain boundaries acting as defect diffusion channel and as defect clustering reservoir as the irradiation conditions (e.g., temperature, dose rate) changes from case to case. These predictions are consistent with the composition changes near GBs observed in our STEM/EELS experiments on samples irradiated under different conditions.
ES5.5: Microstructure and Irradiation
Session Chairs
Kazuto Arakawa
Estelle Meslin
Wednesday PM, April 19, 2017
PCC North, 200 Level, Room 223
11:30 AM - *ES5.5.01
Solving the Puzzle of Nonequilibrium Precipitation in Irradiated Tungsten Using Parameter-Free Modeling
Chen-Hsi Huang 1 , Leili Gharaee 2 , Yue Zhao 1 , Paul Erhart 2 , Jaime Marian 1
1 , University of California, Los Angeles, Los Angeles, California, United States, 2 , Chalmers University, Gothemburg Sweden
Show AbstractNeutron irradiated tungsten is known to evolve through a sequence of microstructural changes depending on temperature and total dose. At high temperature and relatively high total dose, a sizable Re concentration emerges from transmutation reactionts, leading to the formation of e of Re-rich clusters that eventually transform into σ-phase W-Re intermetallics. These precipitates result in very high hardening levels, further weakeningthe and embrittling the alloy. These Re precipitates begin to form at overall Re concentrations of >≈2% at., inconsistent with equilibrium thermodynamics, and even with a picture governed by vacancy-mediated RED and RIP. Recently, first-principles calculations have suggested mixed interstitial solute transport as a posible enabler and enhancer of Re precipitation. In this work we use mesoscale modleing based on kinetic Monte Carlo simulations to study the mechanisms of Re precipitation in neutron irradiated W as a function of temperature and dose, incorporating electronic structure calculations of material constants and diffusion mechanisms.
12:00 PM - ES5.5.02
Plastic Straining in Presence of Radiation-Induced Defects—A 3D Dislocation Dynamics Investigation
Christian Robertson 1
1 , CEA Saclay, Gif-sur-Yvette France
Show AbstractFerritic alloys are widely used as structural nuclear materials, thereby submitted to various dose-dependent evolutions. This paper focuses radiation-induced effect on screw dislocation mobility, bearing in mind that «dislocation motion appears to be the factor defining the nature of the brittle to ductile transition» (and therefore, the fracture toughness response). Screw dislocation velocity fields in tensile strained Fe-Cr grains are evaluated quantitatively, by means of 3D Dislocation Dynamics (DD) simulations, where individual screws move according to thermally activated kink-pair nucleation and subsequent kink-pair propagation at finite velocity [1]. Radiation effects are highlighted by performing different simulations representative of post and pre irradiated Fe-Cr grains. It is shown that the radiation defect populations induce dislocation velocity field changes compatible with an apparent straining temperature downshift, to be called «DDIAT» or Defect Induced Apparent straining Temperature shift. The DDIAT results obtained herein are then presented in the form of a simple, practical, semi-quantitative model. Predicted DDIAT values proved comparable to the DBTT changes obtained in various ferritic alloys subjected to different irradiation conditions, characterized by a given
defect size and a defect number density.
[1] K. Gururaj, C. Robertson, M. Fivel, J. Nuclear Materials, 459, 194 2015.
Acknowledgement: This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053.
12:15 PM - ES5.5.03
Concurrent Atomistic-Continuum Simulations of Sequential Dislocation/Obstacle Interactions in Face-Centered Cubic Metals
Shuozhi Xu 1 , Liming Xiong 2 , Youping Chen 3 , David McDowell 1
1 , Georgia Tech, Atlanta, Georgia, United States, 2 , Iowa State University, Ames, Iowa, United States, 3 , University of Florida, Gainesville, Florida, United States
Show AbstractIn nuclear power systems, the metallic components of reactors are subject to irradiation by a flux of fast neutrons, inducing clustered defects in materials that strongly alter their mechanical properties. When the obstacle is a precipitate with a large size, it is bypassed by an edge dislocation in molecular static (MS) simulations following the continuum model. For a small precipitate that is easily sheared, the depinning stresses given by MS are much lower than the continuum predictions. Assuming an impenetrable, unshearable precipitate, an edge dislocation bypasses it via either Orowan mechanism or Hirsch mechanism, depending on the shearing conditions and the relative locations of the precipitate and slip plane. Dislocation dynamics simulations excluding dislocation cross-slip, however, predict that sequential dislocations bypassing an impenetrable obstacle result in multiple Orowan loops around it. Nevertheless, whether these Orowan loops are still stable in simulations allowing the cross-slip remains unexplored. While linear elasticity theory can provide a description of dislocation/obstacle interactions, approximations have to be made for processes that are controlled by atomic mechanisms, such as the dislocation core structure. Thus, the direct contact between the dislocation and obstacle is more amenably studied in atomistic resolution. On the other hand, dislocation pile-up/obstacle interactions would be exceedingly expensive for atomistic simulations because the stress fields involved are long range. In this work, we employ large scale concurrent-atomistic continuum simulations [1-3] in face-centered cubic metals for parametric study of sequential dislocations bypassing an obstacle, which is either a void or a deformable precipitate which contains the same material as the surrounding matrix but is made impenetrable. It is found that in dislocation/void interactions, both edge and screw dislocations leave simple shear steps on the void surface; for a precipitate, Hirsch looping involving cross-slip is observed for each dislocation bypass, leaving behind lattice defects such as vacancies, dislocation loops, and stacking fault tetrahedral.
[1] L. Xiong, G.J. Tucker, D.L. McDowell, Y. Chen, Coarse-grained atomistic simulation of dislocations, J. Mech. Phys. Solids, 59, 160 (2011)
[2] L. Xiong, S. Xu, D.L. McDowell, Y. Chen, Concurrent atomistic-continuum simulations of dislocation-void interactions in fcc crystals, Int. J. Plast., 65, 33 (2015)
[3] S. Xu, R. Che, L. Xiong, Y. Chen, D.L. McDowell, A quasistatic implementation of the concurrent atomistic-continuum method for FCC crystals, Int. J. Plast., 72, 91 (2015)
12:30 PM - ES5.5.04
Epsilon Phase Fission Products in Spent Nuclear Fuel Cladding—A Complimentary Atomic Resolution Electron Microscopy and Atom Probe Tomography Study
Michele Conroy 1 , Kristi Pellegrini 1 , Edgar Buck 1 , Daniel Perea 1 , Daniel Edwards 1 , Richard Clark 1
1 , Pacific Northwest National Lab, Richland, Washington, United States
Show AbstractThe formation of metallic particles composed of Mo, Tc, Ru, Rh, and Pd (epsilon phase) during irradiation of nuclear fuel is a well-established proposition. Post irradiation studies to date show that these fission products form within the nuclear oxide fuel itself. It has also been hypothesized that epsilon phase formation can affect fuel performance, control cladding erosion, and cause stress corrosion through the physical disruption of the fuel matrix. In this article we report the first experimental evidence of epsilon particle formation within the cladding itself. Our objective is to better characterize the epsilon metal phase and reveal the driving factors that produce them within the cladding. In this study we utilized various electron microscopy techniques along with atom probe tomography (APT) to characterize the particles and their surrounding Zr cladding environment. Elemental mapping and imaging in the scanning electron microscope allowed us to identify the location of the epsilon particles and then prepare electron transparent thin foils and APT needles of these areas using focused ion beam milling. The SEM elemental mapping showed a cloud/halo of uranium enrichment present around each epsilon particle. The particles have a wide scale range from 50 nm to > 2 µm and varied in shape. Using an abberration probe corrected JEOL ARM200CF in scanning transmission mode elemental mapping of individual particles revealed the distribution and concentration ratio of the metals. We found that the concentration of each individual element was mostly non uniform, with pockets of higher concentration varying independently of each other. The exception was Ru and Pd, which had an inverse concentration relationship seen across all particles analyzed. The epsilon phase particle not only included the 5 metal consituents described above, but also contained Te and U. Conventional transmission electron microcopy (TEM) analysis showed that the large particles not only exhibited large single crystal grains, but also smaller crytallites around the periphery of the partciles. TEM nano beam diffraction identified the hcp structure of the grains, further confirming epsilon phase. STEM elemental mapping was also done for APT needles made from the Zr cladding before being characterized by APT.
12:45 PM - ES5.5.05
Multiscale Approach on the Understanding of the Growth Mechanisms of Zirconium Alloys under Irradiation
Benjamin Christiaen 1 2 3 , Alexandre Legris 2 3 , Christophe Domain 1 3 , Ludovic Thuinet 2 3 , Antoine Ambard 1 3
1 , EDF R&D, Moret sur loing France, 2 , UMET (Unité matériaux Et Transformation), Villeneuve d'Ascq France, 3 , EM2VM (Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux), Rouen France
Show AbstractES5.6: Thermodynamics and Modeling
Session Chairs
Jaime Marian
Blas Uberuaga
Wednesday PM, April 19, 2017
PCC North, 200 Level, Room 223
2:30 PM - *ES5.6.01
Modeling Cu-Mn-Ni-Si Precipitate Evolution and Hardening in Reactor Pressure Vessel Steels
Dane Morgan 1 , Huibin Ke 1 , Mahmood Mamivand 1 , Shipeng Shu 1 , Henry Wu 1 , Peter Wells 2 , Nathan Almirall 2 , G. Robert Odette 2
1 , University of Wisconsin-Madison, Madison, Wisconsin, United States, 2 , University of California, Santa Barbara, Santa Barbara, California, United States
Show AbstractIn order to maintain the our current levels of nuclear power it is critical to license present light water reactors for operation beyond their present 40 years to 60 years or more. To enable such licensing, one major factor is assuring the mechanical integrity of the reactor pressure vessel, which in turns requires a detailed model of their embrittlement. Irradiation enhanced precipitation hardening is the primary cause of in-service embrittlement of reactor pressure vessel (RPV) steels, making it critical to understand and predict embrittlement effects over 60 years or more in these materials.
It has been well established that for Cu containing RPV alloys Cu-rich precipitates form rapidly under irradiation and can lead to significant hardening. However, it has become increasingly clear that so-called “late-blooming” Mn-Ni-Si phases, first predicted and measured by Odette and coworkers, may be the dominant cause of RPV embrittlement at high fluence.[1] The need to understand the precipitate embrittlement in RPVs under LWR life-extension conditions, which include low-flux, high-fluence, and time scales of 60 or more years, means that direct experimental characterization is impractical and accurate models are essential to allow prediction for the multitude of RPV conditions relevant to light water reactor life extension.
In this talk we describe a cluster dynamics model we have developed to predict the formation and evolution of Cu-Mn-Ni-Si precipitates under a range of compositions, temperature, flux, and fluence. The model integrates CALPHAD thermodynamics, radiation enhanced diffusion models, and both homogeneous and heterogeneous nucleation terms. We demonstrate that the model semi-quantitatively reproduces a wide range of both irradiation and annealing experiments and provides a number of valuable insights, including the critical role for impurities in Cu precipitate evolution, the necessity of heterogeneous nucleation on cascade damage to explain observed precipitation, the strong sensitivity of precipitate evolution to small changes in composition and temperature, and an understanding of how embrittlement is likely to proceed under life-extension conditions.
Finally, we also describe a complimentary approach to physics based modeling of embrittlement that uses machine learning techniques to predict the hardening of RPVs. We demonstrate the approaches effectiveness on the over 1500 entries in the UCSB Irradiation Variables (IVAR) database and validate its predictive ability by modeling the newly measured results of hardening from the UCSB ATR2 database, which was not used in the fitting of the data.
[1] P. B. Wells, T. Yamamoto, B. Miller, T. Milot, J. Cole, Y. Wu, and G. R. Odette, Evolution of manganese–nickel–silicon-dominated phases in highly irradiated reactor pressure vessel steels, Acta Materialia 80, p. 205-219 (2014).
3:00 PM - ES5.6.02
A Comprehensive Thermodynamic Study of the U-Am-O Ternary System—Multiple Experimental Investigations and Calphad Modelling
Enrica Epifano 1 , Ondrej Benes 2 , Dario Manara 2 , Thierry Wiss 2 , Romain Vauchy 1 , Florent Lebreton 1 , Christine Gueneau 3 , Christophe Valot 1 , Philippe Martin 1
1 , CEA Marcoule, Bagnols-sur-Cèze France, 2 , JRC Karlsruhe, Karlsruhe Germany, 3 , CEA Saclay, Paris France
Show AbstractMinor actinides (MA) are created in the nuclear fuel through successive neutron captures during irradiation. One of the options envisaged for reducing the long-term radiotoxicity of the nuclear waste is the transmutation of these elements in fast neutron reactors. In this frame, the attention is mainly focused on mixed U and Am dioxides (U,Am)O2.
The possibility of using advanced fuels containing MA requires investigating their structural and thermodynamic properties, in order to foresee the in pile behavior. For instance, thermal conductivity and melting temperature are fundamental parameters to determine the safety limits of a fuel. In actinide dioxides, that generally admit a large existence domain, thermal properties are affected by the Oxygen/Metal ratio (O/M). The latter also affects the chemical interaction between the fuel and the cladding. Therefore, a thorough knowledge of the existence domain of the dioxide phase is necessary.
The aim of this work is to develop a thermodynamic model of the U-Am-O ternary system through the semi-empirical CALPHAD method, based on the Gibbs energy assessment of each phase. Unfortunately, experimental data on (U,Am)O2 are rare, because of the scarcity and high radioactivity of Am. For this reason, an extensive experimental campaign on these compounds was carried out in collaboration with the JRC-Karlsruhe. (U,Am)O2 samples with Am/(U+Am) ratios ranging from 0.1 to 0.7 were manufactured by a powder metallurgy process.
Structural data, necessary for a reliable CALPHAD description of each phase, were obtained combining X-ray Diffraction (XRD) and X-ray Absorption Spectroscopy (XAS). XRD confirmed the maintaining of the fluorite-type structure, common to all the actinide dioxides. Nevertheless, XAS investigations highlighted a peculiar charge distribution, with Am reduced to trivalent state and U partially oxidized.
Thermodynamic data have been obtained through different techniques. Drop calorimetry was performed to measure enthalpy increments as a function of temperature and derive the heat capacities of (U,Am)O2 with various Am contents. Melting temperature were measured both under reducing and oxidizing conditions using a laser-heating technique. In order to obtain reliable data, various post-melting characterizations (XRD, XAS, Raman, SEM) were performed to determine the composition and the phases present in the samples after the measurements. Finally, study of the oxygen pressure-temperature-composition equilibria were conducted by thermogravimetry in order to derive the oxygen potentials of (U,Am)O2, whose knowledge is necessary to foresee the O/M under different conditions.
Thanks to the new experimental data acquired on the U-Am-O system, a CALPHAD assessment will be performed and the resulting model will be integrated in the Thermodynamic of Advanced Fuel International Database.
3:15 PM - ES5.6.03
Molecular Dynamics Analysis of Thermodynamic and Kinetic Properties of Bulk PdHx
Xiaowang Zhou 1 , Tae Wook Heo 2 , Brandon Wood 2 , Vitalie Stavila 1 , Shinyoung Kang 2 , Mark Allendorf 1
1 , Sandia National Labs, Livermore, California, United States, 2 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractThis work uses molecular dynamics simulations to study PdHx bulk properties relevant to hydrogen storage applications including lattice constants, Gibbs free energies, elastic constants, and diffusivities as a function of temperature and composition. During the course of the calculations, we demonstrated robust molecular dynamics methods to calculate highly converged finite temperature elastic constants, and highly converged overall diffusion properties accounting statistically for all possible atomic jump mechanisms. The robust calculations reveal ideally linear Arrhenius plots of hydrogen diffusion at low compositions, and abnormally nonlinear plots at high compositions. The fundamental cause for this behaviour has been identified. All of our calculated results are compared with available experimental data. Remarkably, the non-Arrhenius behaviour is validated by experiments. While elastic constants calculated at the 0 K temperature agree well with those measured at ~ 4 K, they differ at higher temperatures. Simulations using a vibrational loading condition suggests that this discrepancy may origin from the difference in the sonic methods typically used to measure elastic constants in experiments and the simulated mechanical testing conditions for PdHx where significant hydrogen diffusion could occur in responding to deformation.
Acknowledgement: Sandia National Laboratories is a multi-mission laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the US Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. The authors gratefully acknowledge research support from the U.S. Department of Energy, Office of Energy Efficiency and Renewable Energy, Fuel Cell Technologies Office, under Contract Number DE-AC04-94AL85000.
ES5.7: Chemical Effects
Session Chairs
Wednesday PM, April 19, 2017
PCC North, 200 Level, Room 223
4:30 PM - *ES5.7.01
Atomistic to Mesoscale Understanding of Hydride Formation in Alpha-Zr
Yongfeng Zhang 1 , Xianming Bai 1 , Chao Jiang 1 , Bulent Biner 1 , Simon Phillpot 2
1 , Idaho National Lab, Idaho Falls, Idaho, United States, 2 , University of Florida, Gainesville, Florida, United States
Show AbstractThe formation of hydrides degrades the toughness of Zr-based claddings and is one the primary concerns regarding cladding integrity during used fuel storage. To assess the effect of hydrides on the mechanical integrity of fuel pins require fundamental understanding on hydrogen transport, hydride formation, and the consequent degradation in mechanical properties. In this talk, the atomistic to mesoscale effort made at INL in collaboration with UFL on hydride formation is presented. At the atomic scale, density functional theory, molecular dynamics, and kinetic Monte Carlo are used to elucidate the thermodynamics of hydrides, kinetics of hydrogen diffusion, and mechanism for hydride nucleation. Here, density functional theory calculations serve to assess the thermodynamic stabilities of H solid solution and various hydride phases and the diffusion mechanisms of H, with the effect of stress included. The results are also used to parameterize a kinetic Monte Carlo model, which gives the effective H diffusivity under different stress states and with impurities that may appear in claddings. Assisted by density functional theory calculations, a recently developed charge-optimized-many-body potential is further improved and used to investigate the possible nucleation paths of hydrides. It’s shown that the formation of face-centered hydrides starts with coherent zeta-phase, shearing of which in the basal plane changes the hcp stacking into fcc and leads to the formation of gamma-hydride nuclei. The shearing involves a negligible barrier and negligible strain accumulation in the matrix by adopting three equivalent shear partials. The information obtained at the atomic scale is used to assist the phase field modeling of hydride formation and reorientation under various stress conditions at the mesoscale. Such a multiscale approach is taken for the purpose of developing a physics-based mechanistic model describing hydride formation and evolution in Zr at finite temperatures and loading conditions.
5:00 PM - ES5.7.02
Fast and Ultrahigh-Capacity Uranium Extraction from Sea Water by Electrochemical Method
Chong Liu 1 , Yi Cui 1
1 , Stanford University, Stanford, California, United States
Show AbstractNuclear energy, as a mature energy source, possesses potential to supply in part the massive global energy demand with minimal greenhouse gas emission. Therefore how to secure vast resources of uranium for nuclear energy fuel is inevitably important. The total amount of uranium in sea water is abundant (hundreds of times more than that in land) and sea water is highly accessible, so it would be attractive to develop sea water uranium extraction technology with large capacity, fast kinetics, and high selectivity. However, the challenge to extract uranium lies in its extremely low concentration (~ 3 ppb) in a high salinity background. Current method based on sorbent materials is limited by their capacities and kinetics due to the surface-based physicochemical adsorption nature. Such method requires large quantity of materials and long period of collection time. Here we introduce, the use of electrochemical method for uranium extraction from sea water based on an amidoxime-functionalized carbon (C-Ami) electrode. The amidoxime functionalization enables the surface specific binding to uranium ions. In the electrochemical method, the electrical field can migrate the ions to the electrode surface and induce electrodeposition of uranium compounds at active sites, forming charge neutral species to avoid Coulomb repulsion. Because of the electrodeposition mechanism, uranium extraction is not limited by the electrode surface area. As a result, the electrochemical method achieved 9-fold higher uranium extraction capacity (1932 mg/g) without saturation and 4-fold faster kinetics than conventional physicochemical methods using uranium spiked sea water as well as high selectivity inherited from the amidoxime functionalization.
5:15 PM - ES5.7.03
Reaction Model for Fluorination of Uranium Dioxide Using Improved Unreacted Shrinking Core Model for Expanding Spherical Particles
Shunji Homma 2 , Artur Braun 1
2 , Saitama University, Saitama Japan, 1 , EMPA, Duebendorf Switzerland
Show AbstractA gas-solid reaction model is developed to represent the fluorination of uranium dioxide (UO2), which consists of a two-step reaction: the formation of a solid intermediate of uranyl fluoride (UO2F2) on thecore 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.
S. Homma, Y. Uoi, A. Braun, J. Koga, S. Matsumoto. 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 (8) 823-837.
5:30 PM - ES5.7.04
In Situ X-Ray Diffraction Study of the Steam Oxidation of Zirconium and Zirconium Alloys
Mohamed Elbakhshwan 1 , Simerjeet Gill 1 , Arthur Motta 2 , Peter Mouche 3 , Weicheng Zhong 3 , Brent Heuser 3 , Randy Weidner 1 , Thomas Anderson 1 , Lynne Ecker 1
1 Nuclear Science and Technology Department, Brookhaven National Laboratory, Upton, New York, United States, 2 Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, State College, Pennsylvania, United States, 3 Nuclear, Plasma, and Radiological Engineering Department, University of Illinois at Urbana Champaign, Urbana, Illinois, United States
Show AbstractZirconium cladding alloys are required to withstand extreme conditions of temperature, pressure and radiation environments, in normal and accident conditions. However, there is a there is a lack of understanding of the reaction mechanism at the solid-fluid interfaces [1]. In our work, a novel in situ sample environment for monitoring structural changes and oxide growth on nuclear cladding steam interfaces using high resolution synchrotron methods was designed and built. The sample environment is made from hastelloy [2] and is highly corrosion resistant and complies with ASTM standards for corrosion testing [3]. It is optimized for in situ XRD studies for structural analysis at XPD beamline at NSLS-II and is available to all users through the proposal process. The capabilities and key design features of the sample environment for monitoring interfacial corrosion phenomenon in real time will be discussed.
In addition, the sample environment was used to study the oxidation of pure zirconium and zirconium based alloys (zircaloy-2, binary zirconium-yttrium) in real time. The in situ results showing the development of the oxide phases from early stages of oxidation up to the stabilization of the monoclinic oxide will be discussed. The ability to couple synchrotron methods for in situ analysis opens up a large number of possibilities for mechanistic corrosion studies and allows for studies of numerous interfacial phenomena including oxidation of thin films, hydride formation, and detection of early oxidation forms.
[1] A. Yilmazbayhan, et. al., International Conference on Environmental Degradation of Materials, 201 (2005).
[2] M. Elbakhshwan, S. Gill, A. Motta, R. Weidner, T. Anderson, L. Ecker, Sample environment for in situ synchrotron corrosion studies of materials in extreme environments, Journal of Review of Scientific Instruments, 87 (2016) http://dx.doi.org/10.1063/1.4964101.
[3] ASTM standards, G2/G2M–06, 2011. Corrosion Testing of Products of Zirconium, Hafnium, and Their Alloys in Water at 680°F or in Steam at 750°F.
5:45 PM - ES5.7.05
Joint Experimental and Theoretical Investigation of Cement Paste under Electron Irradiation
Sophie Le Caer 4 , Lucile Dezerald 1 3 , Khaoula Boukari 2 , Andres Saul 2 3
4 , CEA Saclay, Saclay France, 1 , Institut Jean Lamour, Nancy France, 3 MSE2, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States, 2 , CINAM, Marseille France
Show AbstractOne of the main challenges arising from nuclear waste storage is the design of containers that can ensure a safe confinement of medium-to-long lived radioelements in geological repositories over thousands of years. Cement is an excellent candidate to store large quantities of radioelements since it is inexpensive and easy to manufacture. It is consequently envisioned to store fission products such as 90Sr that represents, along with 137Cs, the largest fraction volume of waste from civil nuclear plants. 90Sr is a β- emitter, so the largest radiation damage expected from its decay is H2 production by water radiolysis, which is likely to occur in cement since it is a porous material filled with water.
Here, we performed systematic measurements of H2 production yields under electron irradiation for samples with three different chemical compositions and at five relative humidities: 0% where the only water in the samples is chemically bound water in hydrated phases, 11%, 43%, 74% and 97% where almost all pores are filled with water. We find that portlandite and calcite, two phases commonly found in cement, inhibit H2 production under electron irradiation. However, Calcium-Silicate-Hydrates (C-S-H), the binding phase of cement, is the stage of very efficient H2 production mechanisms. We used ab initio calculations to investigate the interactions between the electrons and holes produced under irradiation and C-S-H. We find that the electron localizes in the so-called interlayer space of C-S-H, where it can solvate to produce H2 even at 0% relative humidity. The hole does not contribute to H2 production since it localizes in the intralayer space, in which no water is present. This work consequently provides insight on fundamental mechanisms at stake in H2 production in cement, as well as guidelines for optimized cement composition to delay H2 production and accumulation in closed storage facilities.
ES5.8: Poster Session
Session Chairs
Thursday AM, April 20, 2017
Sheraton, Third Level, Phoenix Ballroom
9:00 PM - ES5.8.01
Transport of Hydrogen Isotopes in CuCrZr and the Isotope Effect for Nuclear Applications
Woo Jun Byeon 1 , Hae Won Shin 1 , Hee Soo Kim 1 , Dong Min Kim 2 , Seung Jeong Noh 1
1 Physics, Dankook University, Cheonan Korea (the Republic of), 2 Materials Science, Hongik University, Sejong Korea (the Republic of)
Show AbstractPrecipitation hardened (PH) and oxide dispersion strengthened copper alloys are important materials for nuclear applications in heat sinks because of their high thermal conductivity, thermal stability at high temperature, good mechanical properties, and strong radiation resistance. In this study, transport behaviors of hydrogen and deuterium in ELBRODUR G PH CuCrZr alloy (0.65 wt% Cr, 0.1 wt% Zr) were investigated by using a hydrogen-isotope permeation measurement system built by and located at Dankook University. The ELBRODUR G PH CuCrZr alloy was fabricated into disks (20 mm in diameter and 0.5–1.0 mm in thickness). The feed pressure range was 0.1–1.0 bar, and the temperature range was 300–600 °C. The transport parameters (permeability, diffusivity, and solubility) of hydrogen and deuterium were determined using the time-dependent gas-phase technique. The transport parameters of tritium were calculated from the quantum mechanical model based on a harmonic approximation using the experimentally measured data of hydrogen and deuterium. Similarly, the transport parameters of tritium were also calculated using the data of hydrogen and deuterium obtained from the least-squares fit of the measured ones. Finally, the isotope effect ratios for the transport parameters were estimated.
[This work was supported by the National Research Foundation of Korea (Project No. 2015M1A7A1A01002234)]
9:00 PM - ES5.8.02
Examination of High Burnup Behavior of Nuclear Fuels Using a Modified Potts Model
Richard Hoffman 1 , Chaitanya Deo 1
1 Nuclear & Radiological Engineering and Medical Physics Programs, Georgia Institute of Technology, Atlanta, Georgia, United States
Show AbstractAs nuclear fuel is irradiated in reactors the lattice experiences a high number of displacements per atom. This process is not uniform throughout the fuel and due to geometrical considerations of the reactor setup the outer edges of the fuel experiences signi�cantly more irradiation damage then the center of the fuel. Over time this creates a localized burnup level higher on the rim of the fuel. In turn this damage causes an increase in the fuel's porosity due to fission gas bubbles and a restructuring of the grains in the outer edge of the fuel. The new grain structure is known as High Burnup Structure (HBS). In this work we propose to extend a model of the formation based on the Potts model. Our model will focus on the addition of a fission rate dependent term to the input parameters to make the model more flexible for examining different fuel types. In addition we convert the model from a rejection based model to the n-fold method in order to allow for better comparison with realistic time frames for di�erent fuel types. Finally, we will present results of the new model for several Uranium based fuels.
9:00 PM - ES5.8.03
Multi-Scale Simulation of the Experimental Response of Ion-Irradiated Zirconium Carbide
Jean-Paul Crocombette 1 , Stephanie Pellegrino 1
1 , CEA Saclay, Gif Sur Yvette France
Show AbstractThe response of zirconium carbide to heavy-ion irradiation at room temperature has been studied by X-ray diffraction, ion channeling and transmission electron microscopy. Below 5x1014 cm-2, we observe a build-up of elastic strain with increasing fluences. At this threshold fluence the strain is released and important dechanneling appears as well as visible TEM damage. With increasing fluence, this damage is found to spread in the material deeper than the depth of direct damaging by the ion beam. These experimental observations are reproduced and explained by Density Functional Theory informed Rate Equation Cluster Dynamics simulations. Simulations show that the response of ZrC upon ion-irradiation is driven by the diffusion and clustering of interstitials. The two-step evolution seen in experiments stems from the growth of interstitial clusters with a concomitant starvation of the smallest clusters induced by the continuous accumulation of vacancies. The damaging of the material beyond the range of primary damage is driven by diffusion of interstitials.
Reference :S. Pellegrino, J-P Crocombette A. Debelle, T. Jourdan, P. Trocellier and L. Thomé, Acta Materialia 102, 79-87 (2016)
9:00 PM - ES5.8.04
Irradiation Effects of Helium Ions on the Deuterium Retention and Desorption in the Reduced Activation Martensitic Steel ARAA
Hae Won Shin 1 , Woo Jun Byeon 1 , Hee Soo Kim 1 , CheolEui Lee 2 , Jaeyong Kim 3 , Seung Jeong Noh 1
1 Physics, Dankook University, Cheonan Korea (the Republic of), 2 Physics, Korea University, Seoul Korea (the Republic of), 3 Physics, Hanyang University, Seoul Korea (the Republic of)
Show AbstractDuring the fusion process, the plasma-facing components (PFCs) of fusion reactors are exposed to the ions of hydrogen isotopes and helium. Hydrogen isotopes can be retained to the PFCs and may significantly affect the plasma performance and density control in fusion devices. Reduced-activation ferritic/martensitic steels are candidate materials for the PFCs, and the Korea Atomic Energy Research Institute has been developing an advanced reduced activation alloy (ARAA).
In this study, the irradiation effect of helium ions on deuterium retention and desorption behaviors in ARAA was investigated by using a thermal desorption spectroscopy (TDS) system clustered with an inductively coupled plasma ion source built by and located at Dankook University. The ARAA samples were mechanically polished, chemically cleaned, and degassed at 1173 K for 1 hour. Two samples were irradiated with 1.7-keV deuterium ions at a fixed fluence of 6.5×1021 D/m2 after cleaning and heating. For one sample, the TDS measurement was performed in situ immediately after deuterium irradiation. On the contrary, the other sample was exposed to air for one week after deuterium irradiation prior to the TDS measurement. As a result, the deuterium desorption behaviors between two samples were observed to be apparently different. Furthermore, we also carried out separate experiments. After cleaning and heating, the samples were irradiated with helium ions (1.4 or 7.0 keV) at a fixed fluence of 5.0×1021 He/m2 and were continuously irradiated with 1.7-keV deuterium ions at a fixed fluence of 6.5×1021 D/m2. TDS measurements were performed in situ immediately after deuterium irradiation. Detailed results are provided in the presentation.
[This work was supported by the National Research Foundation of Korea (project No. 2015M1A7A1A01002234)]
9:00 PM - ES5.8.05
Structure of La-U-O Compounds
Luis Casillas 1 , Haixuan Xu 1 , Gianguido Baldinozzi 2 , Kurt Sickafus 1
1 , University of Tennessee, Knoxville, Tennessee, United States, 2 , Centrale Supelec, Paris France
Show AbstractCompounds belonging to the La-U-O compositional space are of technological interest in light water reactor fuel technology. Mixed uranium-lanthanide (Ln) oxide compounds form in the UO2 fuel pellets as Ln fission products are incorporated into the fuel during burn-up. We have used experiments and simulation to determine the structures of selected U-La-O compounds. Diffuse or sharp superstructures along [111]c observed for some compositions suggest that these compounds can be modeled using alternating layers normal to that direction. In this model, the unit cell is composed of alternating planes of anions and cations, which allows us to study in-plane cation ordering. Starting with the parent fluorite phase, and the information either available from existing experimental results or by systematic exploration of the possible solutions involving small supercells, we have determined the symmetry-compatible ordered atomic arrangements of these family of compounds. Density functional theory allows to determine the lowest energy configurations of cations among these layered atom models. Using wet chemistry we have actually synthesized those samples with the most interesting compositions predicted by the numerical modelling, and characterized them by x-ray diffraction and transmission electron microscopy to assess the predictions.
9:00 PM - ES5.8.06
A Complete Study on the Evolution of W and W-Re Alloy under Irradiation Using Monte Carlo Simulation Methods
Chen-Hsi Huang 1 , Jaime Marian 1
1 , University of California, Los Angeles, Los Angeles, California, United States
Show AbstractIn this study, we investigate the evolution of W and W-Re alloy under different scenarios using Monte Carlo (MC) simulations. We first perform kinetic Monte Carlo (kMC) simulations on void formation on pure tungsten under irradiation. In this simulation, vacancy clusters that contain more than 3 vacancies are considered immobile vacancy voids. Several defect mechanisms are included: defect generation, defect jump, recombination, and vacancy attachment. The void formation follows the nucleation and growth mechanism and the increases of voids are observed. The void lattice formation study is still on progress. We then study the defect effects, of vacancies or interstitials, on the equilibrium phase diagram of W-Re alloy by using semi-grand canonical Monte Carlo simulations. The defect-free W-Re alloy shows ordered structure with negative short-range order (SRO) parameters. When vacancies or mixed-interstitial are added to the system, the SRO parameters increases and becomes positive at low solute concentration at temperatures below the threshold temperature, which exhibits segregation induced by defects. The phase diagrams with different defect concentrations are plotted. Finally, we perform kMC simulations on W-Re alloy with both vacancies and interstitials included to study the precipitation at defect sinks. Two types of defect sink are employed. The first one is a perfect plane sink at the middle of the system, where a defect disappears to becomes an atom when jumping onto it. In the second sink simulation, vacuum sites are occupied on the left and right side of the system, creating two surfaces at the contact with bulk material. The surfaces are treated as both defect sink and defect creation sites. With both sink scenarios, enrichment of solute atoms at the sinks is observed. We also perform kMC simulations with no defect sink, and the radiation-induced precipitation in bulk is still under investigation.
9:00 PM - ES5.8.07
Cladding Fretting Wear Comparison of APMT Steel and Zircaloy-4
Thomas Winter 1 , James Huggins 1 , Richard Neu 1 , Preet Singh 1 , Chaitanya Deo 1
1 , Georgia Institute of Technology, Atlanta, Georgia, United States
Show AbstractIn support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been a leading cause cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate and the conventional cladding, Zircaloy-4, under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. Confocal scanning laser microscopy is used to obtain a 3D map of the surface, which is in turn used for wear & work rate calculations on the samples. The wear rate coefficient is used as a measure of the performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.
9:00 PM - ES5.8.08
Microstructural Evolution in Hot Rolled Monotectoid Zr-Nb
Jacob Startt 1 , Tanvi Dave 2 , Hamid Garmestani 2 , Chaitanya Deo 1
1 Nuclear and Radiological Engineering, Georgia Institute of Technology, Atlanta, Georgia, United States, 2 Materials Science and Engineering, Georgia Institute of Technology, Atlanta, Georgia, United States
Show AbstractMicrostructural analysis of interdicted U alloys may suggest processing paths leading to the establishment of the provenance of interdicted nuclear materials. Derivation of the process-path functions of the thermo-mechanical processing of these materials may provide a computational means to analyze the microstructural evolutions. Phase transformations, morphology and crystallographic texture evolution are investigated in hot rolled Zr-Nb, that when considered as a surrogate for U-Nb alloy microstructures may show the formation of metastable phases on casting and rolling. Zr-Nb is chosen because it shares a high temperature bcc phase that is retained upon quenching. Both also share similar phase transition behavior at their monotectoid compositions. The Zr-18.8w%Nb samples are fabricated by arc-melting which results in extra-large grains with no stored strain energy with microstructure. Annealing at 900°C for 2 hours followed by water quenching produces a mixture with precipitates. The specimens are then hot-rolled at 900°C to total reductions of 20%, 40%, and 60% in ~6% increments. X-Ray diffraction (XRD) and Energy Dispersive Scattering (EDS) are used identify phases, while both XRD and electron backscatter diffraction (EBSD) are used to measure the evolution of texture. X-Ray data is finally used with the POPLA software package to determine the coefficients of the process-path function for this system.
9:00 PM - ES5.8.09
Hydrogen Solidification on Template Materials for Nuclear Fusion Applications
Swanee Shin 1 , Luis Zepeda-Ruiz 1 , Alexander Chernov 1 , Bernhard Kozioziemski 1
1 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractSolid hydrogen layers are currently used as a fuel for inertial confinement fusion experiments at the National Ignition Facility (NIF) to initiate nuclear fusion reactions. Routinely making uniform, high-quality solid hydrogen layers, however, still remains a challenge in large part due to the lack of understanding and control of the hydrogen solidification process. To address this challenge, we explored “templating effect” of various materials to find material parameters that are relevant to hydrogen solidification. The effectiveness as a template was quantified as degrees of supercooling required for solidification below the melting point of a hydrogen isotope (H2 or D2). The results showed high supercooling (> 100 mK) for most metallic, covalent, and ionic solids, and low supercooling (< 100 mK) for van der Waals (vdW) solids. Concurrently, we performed molecular dynamics (MD) simulations with varying strengths of substrate-hydrogen interaction potential. Based on the results, we will suggest material parameters important to templated hydrogen solidification.
9:00 PM - ES5.8.10
Modeling Diffusion of Fission Products in SiC High-Energy Grain Boundaries
Hyunseok Ko 1 , Izabela Szlufarska 1 , Dane Morgan 1
1 , University of Wisconsin-Madison, Madison, Wisconsin, United States
Show AbstractTristructural-Isotropic (TRISO) coated fuel particles are a type of micro-fuel planned for the next generation very high temperature reactors. In the current TRISO design, the Silicon Carbide (SiC) layer provides the primary barrier against escape of fission products (FPs). These particles, however, have been reported to release undesirable metallic fission products such as radioactive silver (110mAg) and cesium (137Cs) . To date, the release rate and transport mechanism of these fission products through the SiC remains unsolved. One hypothesis for the mechanism of Ag transport through SiC is diffusion along grain boundaries (GBs), which hypothesis is supported by a number of recent studies demonstrating diffusion of FP is significantly faster in polycrystalline than single crystal SiC (e.g. for Ag, Ref. [1]). In this work we therefore focus on elucidating the Ag/Cs transport behavior in grain boundaries, particularly in the high energy GBs (HEGB).
The Ag/Cs diffusion in HEGBs is simulated using an ab initio based kinetic Monte Carlo (kMC) model. The HEGB is modeled as an amorphous SiC region and density functional theory (DFT) is used to evaluate the formation and migration energies for a number of local environments. The Ag diffusion is simulated with a kMC model using migration barriers and spatial correlations drawn from distributions fit to the DFT data. The Cs diffusion is also simulated with a kMC model, but for Cs the model is based on directly using the transition state energies predicted from the DFT in a finite sized cell. The diffusion coefficient in HEGB (DHEGB) for Ag and Cs are calculated at temperature range of 1200-1600°C. In this range, DHEGB for Ag and Cs are predicted to be 13‒16 and 4‒8 orders, respectively, higher than the diffusion coefficients of Ag and Cs in the bulk. The DHEGB for both Ag and Cs exhibit Arrhenius dependence and show good agreements with experimentally measured diffusion coefficients in unirradiated SiC [1]. In both cases, however, there are remaining discrepancies between DHEGB predictions and post-irradiated measurements, suggesting that other conditions are possibly responsible for release of Ag/Cs (e.g., radiation enhanced diffusion).
Reference
[1] H. Ko, J. Deng, I. Szlufarska, D. Morgan, Comput. Mater. Sci. 121 (2016) 248-257.
9:00 PM - ES5.8.11
Efficient Ab Initio Modeling of Nuclear Materials
George Beridze 1 , Yaqi Ji 1 , Yan Li 1 , Piotr Kowalski 1
1 , Forschungszentrum Juelich GmbH, Juelich Germany
Show AbstractSolid state chemistry of actinide materials is an interesting and challenging research topic investigated in the context of nuclear waste management. Using computational chemistry tools we complement the related experimental research effort by providing an atomic scale insight into the origin of materials behavior and properties. However, Density Functional Theory (DFT), which is currently the only method of choice for simulation of chemically complex materials often dramatically fails for systems containing strongly correlated f-electrons. More accurate but computationally too intensive methods (e.g. hybrid functionals, MP2, CCSD(T)) are proposed in the literature to compute such systems. Even though these approaches give better results than the DFT, their computational cost limits significantly the size and complexity of materials which one could model and makes computing of larger, dozens of atoms bearing solids almost impossible. In this contribution, we will show the performance of simple and computationally cheap modification of DFT, the DFT+U method with the Hubbard U parameter values derived ab initio using recently developed cLDA and cRPA approaches, for prediction of the structural and thermodynamic parameters of Ln and An-bearing materials. We will discuss our benchmarking results on structural and thermochemical parameters of simple uranium-bearing molecules and solids. We will also present the results of our investigation of more complex uranium-bearing minerals such as studtite ([(UO2)O2(H2O)2]2(H2O)), metastudtite ([(UO2)O2(H2O)2]), diuranium pentoxide (U2O5) and strontium uranate (SrUO4). The discussion will be complemented by review of results obtained from investigation of structures, reaction enthalpies, thermodynamic properties and energetics of defect formation of lanthanide-bearing ceramic waste forms such as monazite- and xenotime-type orthophosphates ((Ln,An)PO4) and pyrochlores (A2B2O7).
9:00 PM - ES5.8.13
Multiscale Modelling of Delayed Hydride Cracking
Mitesh Patel 1 , Daniel Balint 1 , Mark Wenman 1 , Adrian Sutton 1
1 , Imperial College London, London United Kingdom
Show AbstractDelayed hydride cracking (DHC) is a failure mechanism affecting zirconium alloy components in nuclear reactors. In the reactor environment, an aqueous corrosion process allows hydrogen to enter the bulk zirconium matrix. Through diffusion under the influence of gradients of stress, chemical potential and temperature, the H atoms form elevated concentration profiles ahead of stress- raisers such as loaded cracks and notches. Once the solvus is exceeded, zirconium hydride platelets precipitate in the vicinity of the flaw tip. The hydride phases are significantly more brittle than the parent metal and hence have a detrimental effect on the mechanical properties of the component. As such, these hydrides are more prone to fracture, which enables the flaw to propagate. The interplay and repetition of diffusion, precipitation and fracture can ultimately lead to structural failure of the component. The overarching aim of DHC research is to quantify this complexity and develop a rigorous failure criterion.
We present a novel multiscale mathematical model of DHC sub-phenomena in which analytical calculations play an important role. In particular, the stress state of the system is determined using theoretical continuum mechanics: the planar elasticity methods of Green tensors, complex potentials and conformal mappings. Subsequently, linear irreversible thermodynamics is used to study stress-driven diffusion and obtain equilibrium hydrogen profiles. As hydrogen atoms preferentially occupy tetrahedral sites in hexagonal close-packed zirconium, the anisotropy exhibited at the atomic scale is described using the elastic dipole model of point defects. A unique simplistic treatment of hydride needles as Somigliana-Volterra dislocation dipoles is discussed. Additionally, various techniques concerning the computational geometry of polygons are employed to study the effects of heterogeneity and incorporate a set of constitutive rules for the morphological evolution of hydrides. The advantage of this framework is that the calculations are less computationally intensive, supporting a statistical treatment of the DHC problem.
Symposium Organizers
Dilpuneet Aidhy, University of Wyoming
Kazuto Arakawa, Shimane University
Estelle Meslin, CEA Saclay
Haixuan Xu, University of Tennessee
ES5.9: Metallic Systems IV
Session Chairs
Chaitanya Deo
Estelle Meslin
Thursday AM, April 20, 2017
PCC North, 200 Level, Room 223
9:00 AM - *ES5.9.01
Toward a Mechanistic Understanding of Interfacial Damage Tolerance
Mitra Taheri 1
1 , Drexel University, Philadelphia, Pennsylvania, United States
Show AbstractThe development of radiation tolerant materials has long been a goal in the nuclear energy community. Many approaches have been taken to tailor microstructure toward in order to defect accumulation. The use of high sink density materials, such as nanocrystalline, multilayered, and even porous materials, has been explored extensively. Additionally, progress has been made with advanced chemistries, such as the use of high entropy alloys, and processing methods, such as grain boundary engineering. What remains unclear is what exactly contributes to defect absorption efficiency in these various interfaces. This talk presents results from systematic studies of absorption processes at characteristic interfaces, or sinks, using in situ and ex situ TEM imaging coupled with quantitative techniques, such as precession electron diffraction and strain mapping. The results provide a platform from which a new, refined model of sink efficiency can be developed.
9:30 AM - *ES5.9.02
Influence of Chemical Disorder on Defect Dynamics in Concentrated Solid-Solution Alloys
Yanwen Zhang 1 2 , Mohammad Wali Ullah 1 , Shijun Zhao 1 , Eva Zarkadoula 1 , Laurent Beland 1 , Gihan Velisa 1 , Ke Jin 1 , Haizhou Xue 2 , Chenyang Lu 3 , Hongbin Bei 1 , Dilpuneet Aidhy 4 , German Samolyuk 1 , Lumin Wang 3 , G. Malcolm Stocks 1 , William Weber 2 1
1 , Oak Ridge National Lab, Oak Ridge, Tennessee, United States, 2 Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee, United States, 3 Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan, United States, 4 Department of Mechanical Engineering, University of Wyoming, Laramie, Wyoming, United States
Show AbstractHistorically, alloy development with better radiation performance has been focused on dilute alloys with minor alloying elements where enhanced radiation resistance depends on microstructural or nano-scale features to mitigate displacement damage. In sharp contrast to dilute alloys, recent advances of concentrated solid solution alloys (CSAs) have opened up new frontiers in materials research. In these alloys, a random arrangement of multiple elemental species on a crystalline lattice results in disordered local chemical environments and unique site-to-site lattice distortions. Based on closely integrated computational and experimental studies using a novel set of CSAs in a face-centered cubic structure, we have explicitly demonstrated that increasing chemical disorder can lead to a substantial reduction in electron mean free paths, as well as electrical and thermal conductivity, which results in slower heat dissipation in CSAs. The chemical disorder has also a significant impact on defect evolution under ion irradiation due to substantial modifications of the energy landscapes. Considerable improvement in radiation resistance is observed with increasing chemical disorder at electronic and atomic levels. The insights into defect dynamics may provide a basis for understanding elemental effects on evolution of radiation damage in irradiated materials and may inspire new design principles of radiation-tolerant structural alloys for advanced energy systems.
Research supported by EDDE, a DOE-BES Energy Frontier Research Center.
10:00 AM - ES5.9.03
Defect Properties in Concentrated Solid-Solution Alloys
Shijun Zhao 1 , G. Malcolm Stocks 1 , Yuri Osetsky 1 , Yanwen Zhang 1
1 , Oak Ridge National Lab, Oak Ridge, Tennessee, United States
Show AbstractConcentrated solid-solution alloys (CSAs) have recently proven to hold great promise as structural materials due to their extraordinary mechanical properties and excellent irradiation resistance. In contrast to traditional alloys based on one or two principal elements, CSAs are comprised of multiple elements at or near equiatomic concentrations occupying a simple crystalline lattice. These alloys are characterized by disorder from both random arrangement of elements and the atomic displacement fluctuations. As a result, the defect energies in CSAs exhibit distributions rather than single values as in pure metal and dilute alloys. Based on ab initio calculations and special quasirandom structures, the formation and migration energies for a series of face-centered cubic CSAs derived from the Cantor high entropy alloy are investigated. It is found that the distribution of defect energies is strongly dependent on the properties of the constituent elements. In particular, the structural details of the CSAs play a dominant role in determining the defect energies. A preferable diffusion of one or two constituent elements in CSAs is revealed and it largely relies on the correspondingly defect formation energy distributions. Furthermore, molecular dynamic simulations reveal that the diffusion of large vacancy clusters is suppressed in CSAs. The predicted smaller vacancy clusters in CSAs is consistent with experimental observations. The information on the behaviors of defect and defect clusters in CSAs are helpful for the understanding and further design of irradiation-resistant alloys.
10:15 AM - ES5.9.04
Cr-Induced Fast Vacancy Cluster Formation in Ni Alloys
Debajit Chakraborty 1 , Dilpuneet Aidhy 1
1 , University of Wyoming, Laramie, Wyoming, United States
Show AbstractWe perform molecular dynamics (MD) simulations on pure Ni, Ni-Cr and Ni-Fe concentrated solid solutions to elucidate and compare kinetics of vacancy diffusion and cluster formation. We find that diffusion-based vacancy clustering leads to formation of stacking fault tetrahedra (SFT) in all three systems. We also find that the presence of Cr in Ni-Cr alloys leads to – (i) faster SFT formation, and (ii) high Ni diffusion. In contrast the kinetics of Ni diffusion and SFT formation is slower in pure Ni and Ni-Fe systems. This fast kinetics in Ni-Cr systems is due to the significantly low migration barrier of Cr; the low migration barriers first induce vacancy diffusion, which later leads to faster clustering of vacancies into SFT formation. The high Ni diffusivity is also Cr-induced, i.e., Cr-diffusion that induces vacancy diffusion creates small vacancy-vacancy nearest neighbor configurations around Ni atoms. Under this defect configuration, Ni migration barrier decreases from 1.08 eV to 0.44 eV, thus enhancing Ni diffusion. The defect configuration is captured in our MD simulation snapshots, where higher vacancy-vacancy nearest neighbor distribution is observed in Ni-Cr alloys, in comparison to smaller distribution in pure Ni and Ni-Fe alloys. We conclude that addition of Cr enhances vacancy diffusion, whereas Fe has no effect. This work was supported by 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.
10:30 AM - ES5.9.05
Evaluating Irradiation Effects and Defect Kinetics in a Co-Free High Entropy Alloy
Congyi Li 1 , Xunxiang Hu 2 , G. Malcolm Stocks 2 , Steven Zinkle 1 , Brian Wirth 1
1 , University of Tennessee, Knoxville, Tennessee, United States, 2 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractHigh entropy alloys (HEAs) are potential structural materials candidates for high-temperature fission or fusion reactors due to a potentially superior balance of properties. However, limited research has examined the radiation resistance of this class of materials. Here, we present the results of experimental and computational research to evaluate the ion or neutron irradiation damage of Co-free, near equimolar CrMnFeNi HEA as a function of irradiation temperature. Preliminary post-irradiation examination suggests good resistance to radiation-induced segregation and void swelling from 400 to 700 C and doses up to 10dpa. In this presentation we will focus on the results of Ni ion irradiation at somewhat higher temperature of 800 C to 10dpa, and describe the results of analytical transmission electron microscopy to examine HEA’s resistance to radiation. As well, we report results of neutron irradiation to 0.1 and 1dpa at 60 C. Positron annihilation spectroscopy has been used to reveal the evolution of vacancy-type defects following these neutron irradiations, as well as to reveal the post-irradiation annealing recovery kinetics. Macroscopic property changes have been assessed by measuring the micro-hardness and electrical resistivity at various annealing temperatures. These experimental studies are connected to ab-initio modeling to establish a foundation for understanding any unique radiation effects caused by the unique chemical environment of HEA. Both coherent potential approximation and plane wave methods are implemented to compute the ground state of this HEA. In particular, the ab-initio studies have involved nudged elastic band and cluster expansion calculations to estimate the vacancy formation and migration energies. This modeling effort will facilitate the understanding of vacancy kinetics in neutron/ion irradiated HEA, and provides new insights on HEA solute diffusion mechanisms, which will be described in the presentation.
10:45 AM - ES5.9.06
Effect of Sinks on the Kinetics of Cr Precipitation in FeCr Alloys under Irradiation—APT versus AKMC Modeling
Estelle Meslin 1 , Frederic Soisson 1 , Olivier Tissot 1 2 , Cristelle Pareige 2 , Jean Henry 1 , Brigitte Decamps 3
1 , CEA Saclay, Gif sur Yvette France, 2 , GPM, Rouen University, Rouen France, 3 , CSNSM, Orsay University, Gif sur Yvette France
Show AbstractEstelle Meslina, F. Soissona, Olivier Tissota, Cristelle Pareigeb, Jean Henrya and Brigitte Decampsc
aCEA/DEN, SRMP and SRMA, Univ. Paris-Saclay, F-91191 Gif-sur-Yvette, France.
bGPM – UMR 6634 CNRS -Université et INSA de Rouen, 76801 St Etienne du Rouvray
cCSNSM – UMR 8609 CNRS- Université Paris-Saclay, 91400 Orsay, France
FeCr alloys are model alloys of high-Cr Ferritic-Martensitic steels, candidates for structural alloys for the future generation IV and fusion reactors. At low temperatures, Fe-Cr alloys undergo a coherent phase separation between Fe- and Cr-rich phases. This α-α’ decomposition can be considerably accelerated by irradiation, leading to possible hardening and embrittlement of F/M steels. In this work, we have studied the kinetics of phase separation in Fe-Cr alloys under charged particles conditions (electrons or self-ions) in a Fe15Cr model alloys at 573K by means of atom probe tomography [1,2]. These experimental data have been compared to a multiscale simulation approach coupling ab initio calculations (that provide an accurate description of point defect properties), atomistic Monte Carlo simulations (for the modeling of the α-α’ decomposition), and cluster dynamics (for the evolution of point defect sinks). The contribution of sinks on the acceleration of precipitation and on the ballistic dissolution of precipitates is discussed.
[1] O. Tissot, et al. Scr. Mater. 122 (2016) 31–35.
[2] O. Tissot, et al. Mater. Res. Lett. 0 (2016) 1–7.
ES5.10: Fuels
Session Chairs
Dilpuneet Aidhy
Chaitanya Deo
Thursday PM, April 20, 2017
PCC North, 200 Level, Room 223
11:30 AM - *ES5.10.01
Technical Accomplishments in Westinghouse Accident Tolerant Fuel Program
Peng Xu 1 , Edward Lahoda 1
1 , Westinghouse Electric Company, Hopkins, South Carolina, United States
Show AbstractThe Westinghouse accident tolerant fuel (ATF) program is aimed at providing a significant increase in fuel robustness and accident tolerance for beyond design basis accidents at an economically attractive price to nuclear power providers. The Westinghouse ATF program, which began in 2004, advanced significantly during the current Department of Energy (DOE) program which began in 2012. The program will culminate in the manufacture of fuel rods that will be irradiated at steady state conditions in the Advanced Test Reactor and the Halden test reactor for up to 6 years. Fuel transient tests in TREAT and Halden will follow. The goal is to load lead test assemblies that consist of U3Si2 fuel in both coated zirconium cladding and SiC cladding in a commercial reactor in 2019 and 2022. This talk will review the technical accomplishments of the Westinghouse ATF program and outline the tasks that will be performed in the next phase for reactor irradiation.
12:00 PM - ES5.10.02
Irradiation Behavior of the Zirconium Diffusion Barrier in Monolithic U-Mo Fuels
Jan-Fong Jue 1 , Dennis Keiser 1 , James Madden 1 , Brandon Miller 1 , Adam Robinson 1
1 , Idaho National Lab, Idaho Falls, Idaho, United States
Show AbstractIn order to convert US high-performance research reactors to use of low-enriched uranium, a monolithic fuel type is being developed. Based on early fabrication studies and developmental irradiation testing results, a zirconium diffusion barrier between the U–Mo fuel meat and Al–6061 cladding has been incorporated into the design of the down selected fuel system. Utilizing electron microscopy, microstructural characterization has been performed on as-fabricated fuel plates produced by the laboratory-scale co-rolling and hot isostatic press process, as well as on selected irradiated fuel plates. In general, the zirconium diffusion barrier exhibits an equiaxed grain structure with an average grain size less than 20 μm before irradiation. It is also observed that the interaction layer between the zirconium and cladding appeared to be unchanged after irradiation up to a fuel-plate average fission density of 7.5 × 1021 fission/cm3. Even though the interaction layer between the zirconium and U–Mo becomes mostly non-planar at high burnup, no accumulation of large voids/bubbles on the interface or interfacial delamination was observed. Fission products in measureable quantities were found penetrating into only the first 5–10 μm of the diffusion-barrier layer. The grain structure remains discernable in the zirconium diffusion barrier close to the cladding up to a fuel-plate average fission density of 7.5 × 1021 fission/cm3. The effects of irradiation on the zirconium diffusion barrier in monolithic U-Mo fuels will be discussed.
12:15 PM - ES5.10.03
A Modified Embedded-Atom Method Interatomic Potential for U-Si
Benjamin Beeler 1 , Michael Baskes 2 3 , David Andersson 2 , Yongfeng Zhang 1
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 3 , University of California, San Diego, La Jolla, California, United States
Show AbstractUranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy fuel benefits from higher thermal conductivity and higher fissile density compared to UO2. In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. In this study, a semi-empirical Modified Embedded Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential accurately describes not only the primary phase of interest (U3Si2), but also a variety of U-Si phases across the composition spectrum. A ternary U-Si-Xe potential is also developed for the investigation of fission gas behavior in U-Si systems.
12:30 PM - ES5.10.04
Radiation Stability of Spent Nuclear Fuel—In Situ Experimental Simulation
Yara Haddad 1 , Aurelie Gentils 1 , Frederico Garrido 1
1 , CSNSM, Univ Paris-Sud, CNRS/IN2P3, Université Paris-Saclay, Orsay France
Show AbstractEver since the early days of the nuclear industry researches devoted to nuclear fuels enlightened the legendary radiation stability of fluorite-structured oxides. Uranium dioxide does not become amorphous under irradiation but exhibits instead a defective structure, whose specific microstructure depends on several parameters (burnup, local temperature, irradiation conditions, incorporated impurities). Although the basic mechanisms of defects production in irradiated solids are well established, considerable experimental and computational efforts are undertaken to better understand the exact role played by the various relevant parameters in the formation of a specific microstructure and on the final destabilization of the solid.
Experimental simulations based on the use of monoenergetic ions offer the unique opportunity to investigate the behavior of a material under irradiation. Such an approach was successfully applied to urania single crystals irradiated with low-energy ions to examine the contribution of ballistic damage and the contribution of implanted species. Crystals were alternatively (i) implanted at increasing fluence steps with few 100-keV Xe ions at 500°C (the temperature at the periphery of the fuel) and (ii) characterized in situ by Rutherford Backscattered Spectrometry in Channeling geometry (RBS/C) and Transmission Electron Microscopy (TEM). Channeling data were analyzed by Monte Carlo simulations. Two important steps in the disordering kinetics of the solid were established and they were interpreted in terms of the transition from the formation of isolated defects to extended defects at a low dpa number, and due to the aggregation of impurities when their concentration reaches a critical threshold.
Lucie Delauche and staff of the SCALP / JANNuS-Orsay facility (CSNSM, Orsay, France) are gratefully acknowledged for their strong technical support.
12:45 PM - ES5.10.05
Quantitative Image Analysis of Surrogate Fuel Blocks Imaged with X-Rays for the Transient Reactor Test Facility—A Preliminary Study
Jeffery Aguiar 1 , Seongtae Kwon 1 , Karen Wendt 1 , Robert Seifert 1 , Benjamin Coryell 1 , Erik Luther 2
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractThe office of Materials Management and Minimization is seeking to convert the Transient Reactor Test Facility (TREAT) from a high-enriched uranium (HEU) to low-enriched uranium (LEU) fueled reactor. Over the lifetime of the TREAT reactor from 1959 to the present day, a series of historical tests and experiments have disclosed certain material specific requisites that require the fuel design to achieve a desired performance profile, which includes a high thermal conductivity gradient that results in negligible structural damage after exposure to the reactor and transient type experiments. Comparable fuel design, material testing, and qualifications thereby needs to occur for the HEU to LEU fuel conversion (<20% U-235). This includes, retaining a uniform precipitate dispersion of fueled UO2 micron-size particles throughout a graphite-moderating matrix, where the moderator is in direct contact with the fueled particles. On that note, future LEU fuel will consider the relative fuel sizing and spacing in accordance with the expected thermal, neutron, and energy portfolio for future conversion.
In light of the above, the LEU conversion fuel blocks must undergo critical materials testing and examination to determine the properties of manufactured replacement fuel blocks. In this presentationHerein, we will report in this presentation on value-added evaluated particle size distributions using non-destructive X-ray micron-tomography (µCT) of fabricated surrogate fuel blocks. The work supports decisions regarding the fabrication specifications and tolerances associated with the fueled particle size, distribution, and spacing. The developed computational image analysis methods to distill 3-D particle-size distributions, nearest neighbor distances, volume fraction particle curves, and full reconstructions from µCT imagery is of particular interest to the larger materials community, where the techniques can be applied to larger community datasets. At this point, the results provide a current methodology to critique soon to be fabricated LEU TREAT fuel blocks using µCT to inform and optimize fuel fabrication efforts.
ES5.11: Metals and Theory
Session Chairs
Thursday PM, April 20, 2017
PCC North, 200 Level, Room 223
2:30 PM - *ES5.11.01
Molecular Dynamics Simulations of Cascades with Realistic Energies—First Applications of the Cell Molecular Dynamics for Cascade (CMDC) Code
Jean-Paul Crocombette 1
1 , CEA Saclay, Gif Sur Yvette France
Show AbstractThe evolution of materials under irradiation depends on the damage directly created by the atomic displacement cascades initiated by the fast moving neutrons or ions. The exact nature of this so-called primary state of damage pilots the subsequent number and nature of point and extended defects. The modelling of primary irradiation events of realistic energies is a difficult task especially when one wishes to describe ion irradiation. Several strategies exist. First, Binary Collision Approximation (BCA) calculations are very fast but approximate. Molecular Dynamics (MD) simulations are more accurate but much heavier which strongly limits the energy of calculated cascades and number of cascades that are performed in a given study. This gives poor statistics on the nature and amount of damage at high energies and forbids in practice the direct comparison with experimental irradiations. Conceptual divisions of large cascades into subcascades of tractable size for MD simulations have been designed but their general applicability is not guaranteed.
We propose a modified MD simulation of cascades: Cell Molecular Dynamics for Cascades (CMDC). We try to accelerate as much as possible the calculation of cascades by MD without loss of accuracy of the results, considering the specificities of the cascade unfolding. The goal of the method is to be able to model the primary damage created by cascades of realistic energies, e.g. ion irradiations with experimental implantation energies.
CMDC is based on the observation that many parts of the usual MD boxes do not take part in the cascade and are just present in case the cascade would go there. The idea behind CMDC is then to build the box as the cascade develops. More and more atoms are added as the cascade unfolds. Symmetrically, the parts of the materials where the cascade is over, i.e. when the local structure does not evolve anymore on the MD scale (after the ballistic and thermal phases), are removed from the dynamic simulation.
We will present validations of the CMDC code compared to standard MD simulations in the cases of Fe PKA in iron and U PKA in UO2. We shall then present the results of cascade simulations in two ordered alloys (Ni3Al and UO2) for energies ranking between 0.1 and 580keV for both light and heavy constituents. The average number of subcascades and average number of defects per subcascades as a function of ballistic energy exhibit an unexpected variety of behaviors above the threshold for subcascade formation. Finally modeling of irradiations corresponding to real irradiations, (390 keV Xe and 4 MeV Au irradiations in UO2) will be presented with some emphasis on the difference between thin lamellas and bulk samples.
3:00 PM - *ES5.11.02
Uncertainty Quantification and Sensitivity Analysis of Atomistic and Mesoscale Models of Nuclear Materials Behavior
Chaitanya Deo 1
1 , Georgia Institute of Technology, Atlanta, Georgia, United States
Show AbstractAtomistic and mesoscale methods are used to develop fundamental understanding of nuclear materials properties. These include molecular dynamics simulations of atomic evolution and mesoscale kinetic Monte Carlo simulations of defect evolution. The calculated values depend significantly on the choice of model parameters model inputs. This talk describes methods to quantify uncertainty through sensitivity analysis of the commonly used atomistic models – namely molecular dynamics and Monte Carlo simulations of materials behavior.
First, a sensitivity analysis of the modified embedded atom method (MEAM) potential for body-centered-cubic uranium and zirconium is described. The analysis is conducted in order to examine and understand the uncertainty in the parameters and formalism of the interatomic potential. One-at-a-time (OAT) sampling of the potential parameters is used to study how they affect the ground state, thermal, and alloy structural and thermodynamic properties.
Second, uncertainty quantification using sensitivity analysis of kinetic Monte Carlo models of defect behavior in bcc metals under irradiation is conducted. Mesoscale kinetic Monte Carlo simulations of a model bcc metal undergoing irradiation are parameterized with point defect migration energies, 1D to 3D rotational probability and the rate of defect production. This KMC “rate catalog” is used to simulate the evolution of defects – namely the concentration of vacancies and interstitials. Sensitivity analysis is conducted of the effect of varying the rate catalog on the nature of the point defect balance evolution.
In both cases, limits of the applicability of these atomistic/mesoscale models are explored. The work offers opportunity to develop more confidence in atomistic and mesoscale simulation methods and to create a framework on which UQ of multiscale methods of nuclear materials behavior may be conducted.
3:30 PM - ES5.11.03
Development of Molecular Dynamics Potential for Xe-U-Si Ternary System
Jianguo Yu 1 , Yongfeng Zhang 1 , Jason Hales 1
1 , Idaho National Lab, Idaho Falls, Idaho, United States
Show AbstractUse of uranium–silicide (U-Si) in place of uranium dioxide (UO2) is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. However, fission gas bubble growth and fuel swelling on U-Si fuels has been considered as a major determinant in swelling behavior Better atomistic-level understanding the kinetics of gas bubble swelling is of paramount importance to predict fuel performance and to advanced accident tolerance fuel design. Although it is a more feasible way to employ molecular dynamics (MD) simulation to discover the underlying mechanisms of gas bubble induced fuel swelling under irradiation, so far no MD potential is available for the Xe-U-Si ternary system. Here, we will present our recent progress in developing a transferrable Xe-U-Si ternary potential based on experimental and ab initio data. While focusing on the U3Si2 phase, a discussion in light of the transferibility will also be presented so that the potential can be used to study secondary phases, such as (α, β, γ)-U and U3Si, which are one of the most important factors leading to the breakaway swelling in the performance of U3Si2 fuels. This work is supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program funded by the U.S. Department of Energy, Office of Nuclear Energy.
3:45 PM - ES5.11.04
Defect Energetics of Ni-Based Concentrated Solid-Solution Alloys Using DFT
Debajit Chakraborty 1 , Shubham Pandey 2 , Simon Phillpot 2 , Giovanni Bonny 3 , Dilpuneet Aidhy 1
1 , University of Wyoming, Laramie, Wyoming, United States, 2 , University of Florida, Gainesville, Florida, United States, 3 , SCK CEN, Mol Belgium
Show AbstractConcentrated solid-solution alloys (CSAs) contain two or more alloying elements randomly distributed in a single phase. CSAs have shown enhanced properties such as high thermal stability, enhanced mechanical strength and improved irradiation resistance over conventional alloys. In this work, we explore defect energetics in Ni-based CSAs containing Fe, Cr, Pd, Co or Mn as the alloying elements forming concentrated alloys using density functional theory (DFT) calculations. The defect energies of particular interest are: (1) interstitial formation energies, i.e., <100> dumbbell, <111> dumbbell, tetrahedral and octahedral interstitials, (2) vacancy formation energies, (3) binding energies between point defects and different elements, and (4) defect migration energies. We will also report a comparative study with dilute solutions and ordered systems such as L10 and L11.
ES5.12: Metallic Systems V
Session Chairs
Dilpuneet Aidhy
Jean-Paul Crocombette
Thursday PM, April 20, 2017
PCC North, 200 Level, Room 223
4:30 PM - *ES5.12.01
Towards a Multiscale Experiment of Irradiated Material with High-Energy Synchrotron X-Rays
Meimei Li 1 , Xuan Zhang 1 , Chi Xu 1 , Yiren Chen 1 , Jon Almer 1 , Jun-Sang Park 1 , Peter Kenesei 1 , Hemant Sharma 1
1 , Argonne National Lab, Lemont, Illinois, United States
Show AbstractThis paper will present recent results of in situ characterization of neutron-irradiated ferritic and austenitic alloys under tensile deformation using multiple probes of high-energy synchrotron X-rays. Wide-angle X-ray scattering (WAXS) provides rich information of atomic-scale structure of a material, e.g. lattice defects, lattice strain; small-angle X-ray scattering (SAXS) probes nanometer-sized voids, bubbles, and second-phase precipitates; high-energy diffraction microscopy (HEDM) (also called 3D XRD) measures grain- and subgrain-scale structural and micro-mechanical responses; micron-scale pores and cracks can be revealed in 3D by X-ray tomography. When these techniques are combined with in situ thermal-mechanical tests, the evolution of internal structure and micromechanical state of a material during deformation can be probed over a range of length scales. This talk will discuss the new findings of irradiation effects on the deformation process in Fe-based alloys with current in situ thermal-mechanical loading capability of activated materials, and the outlook for in situ time- and spatial- resolved characterization of grain dynamics in neutron- irradiated materials.
5:00 PM - ES5.12.02
Stability and Evolution of Cavities in Aluminum—Experiments and Modeling
Camille Jacquelin 1 , Estelle Meslin 1 , Maylise Nastar 1 , Chu Chun Fu 1
1 , CEA Saclay, Saclay France
Show AbstractAl-based alloys are foreseen for the cladding of the future exprimental reactor Jules Horowitz. Under irradiation, a large amount of point defects (self-interstitials and vacancies) are created. They are mobile and may cluster to form extended defects such as dislocation loops or cavities. This work is focused on cavities, which may induce swelling and embrittlement. Because of its low atomic number, it is possible to create defects in this metal with the typical low-energy electron available within a Transmission Electron Microscope (TEM), as recently obtained at the atomic scale in bulk magnesium [1].
Facetted cavities are indeed observed with in-situ experiments performed with 200 and 300 keV electrons. In-situ observations show that cavities appear after the formation of interstitial loops. Cavities adopt three main 2D projected shapes in [100] zone axis during irradiation: octahedral, square and cross shaped. Octahedral shaped cavities showing stable {111} and {100} planes appearing for large sizes, in agreement with the Wulff construction given by recently published ab initio data [2]. Square shaped cavities are observed for all size range and show very stable {111} planes. Finally, cross shaped cavities appear only for small sizes and develop along <100> directions. Previous and our own ab initio results strongly suggest the pentavacancy as a possible nucleus for the cavity formation, due to its high stability. Thermodynamic and kinetic models based on a pentavacancy cluster [3] are therefore proposed to explain the experimental observations. Finally, the kinetic evolution of the cavities will be discussed in the light of experimental data, and impurity effects on the early stages of vacancy-cluster nucleation will be investigated via ab initio studies.
References
[1]W. Xu et al., In-situ atomic-scale observation of irradiation-induced void formation, Nat. Commun, 4, (2013) 2288
[2]S.S Gupta et al., Depth dependence of vacancy formation energy at (100), (110), and (111) Al surfaces: A first-principles study, Phys. Rev. B 93, (2016)
[3]H. Wang et al., Defect kinetics on experimental timescales using atomistic simulations, Phi.Mag., 93:1-3, (2013),186-202
5:15 PM - ES5.12.03
Scaling Laws of Cascade and Sub-Cascade Formation in High Energy Ion and Neutron Impacts
Andree De Backer 1 , Andrea E. Sand 1 2 , Kai Nordlund 2 , Charlotte Becquart 3 , Christophe Domain 4 , Laurence Luneville 5 6 , David Simeone 6 7 , Sergei Dudarev 1
1 , CCFE, Culham Science Center, Abingdon United Kingdom, 2 Department of Physics, University of Helsinki, Helsinki Finland, 3 , University of Lille, CNRS, INRA, ENSCL, Lille France, 4 , EDF Lab Les Renardi\`eres, Moret sur Loing France, 5 , CEA/DEN/DANS/DMN/SERMA/LLPR-LRC CARMEN, Gif sur Yvette France, 6 , Centralesupelec/SPMS/LRC CARMEN, Chatenay Malabry France, 7 , CEA/DEN/DANS/DMN/SRMA/LA2M-LRC CARMEN, Gif sur Yvette France
Show AbstractIn the treatment of the microstructural evolution of irradiated materials, a critical aspect of the analysis is associated with the definition of the source term describing the generation of radiation defects in collision cascades. Cascade events produce clusters of defects and not just individual defects. The statistics of defect production is a necessary part of any realistic microstructural evolution model.
Computer simulations show that the defect size versus the frequency of occurrence distribution is well represented by a power law [1,2]. This finding is confirmed by electron microscope images of defect microstructure [3]. The outstanding question concerns the part played by cascade splitting into sub-cascades at very high energy. Developing a model taking into account the cascade fragmentation is essential for treating cascade events exceeding a certain threshold energy value.
Using a BCA model for collision cascades, we investigated the spatial repartition of the deposited energy. Using a local energy criterion, the cascades can be described in terms of separated sub-domains. The number of sub-domains as a function of the cascade energy exhibits a transition from a single domain to several sub-domains. Above the cascade splitting energy the cascades form branches and the number of sub-domains as a function of the deposited energy in their volume follows a power law, different from the power law describing the defect size distributions. Applying our method in more than 20 metals, it was possible to rationalize both the effect of the atomic number an the atomic density.
Preliminary results have been reported in [4]. We propose a new model combining the statistics of defect production in low energy compact cascades with the statistics of sub-cascades and which defines the defect size distributions generated by ion and neutron impacts over a broad range of cascade event energies. Our model has been compared with full Molecular Dynamic in several metals including iron and tungsten.
[1] A. E. Sand, S.L. Dudarev and K. Nordlund, EPL, 103 (2013) 46003.
[2] A. E. Sand, M. J. Aliaga, M. J. Caturla and K. Nordlund, EPL, 115 (2016) 36001.
[3] X. Yi, A. E. Sand, D. R. Mason et al., EPL, 110 (2015) 36001.
[4] A. De Backer, A. E. Sand, L. Luneville, D. Simeone, K. Nordlund, S. L. Dudarev, EPL, 115 (2016) 26001.
5:30 PM - ES5.12.04
Proton Irradiation Effect on Nanostructured Half-Heusler Thermoelectric Materials
Nicholas Kempf 1 , Karthik Chinnathambi 1 , Jonathan Gigax 2 , Zhifeng Ren 3 , Lin Shao 2 , Brian Jaques 1 , Darryl Butt , Yanliang Zhang 1
1 , Boise State University, Boise, Idaho, United States, 2 , Texas A&M, College Station, Texas, United States, 3 , University of Houston, Houston, Texas, United States
Show AbstractThermoelectric materials have applications in extreme environments of high radiation flux, such as nuclear reactors and space. Thermoelectric materials can be used in nuclear power plants to power remote sensor networks, offering the potential to expand remote monitoring of facilities for increased safety and cost savings. In space, the materials can be powered by the thermal energy of radioactive materials, providing electricity for decades. Because of these applications, there is great interest in characterizing the irradiation effect on high-performance thermoelectric materials. Under the influence of radiation, it is possible to have exchanges in atomic positions as well as structural changes in bond configuration in thermoelectric materials, thereby altering their properties and performance. Until now, the effect of proton irradiation on nanostructured half-Heusler materials has remained unexplored.
In this work, a wide array of characterization techniques are used to determine and co-validate changes in thermal conductivity, electrical conductivity, and Seebeck coefficient. A significant change in thermal and electrical conductivity is observed as a result of proton irradiation while Seebeck coefficient remains unchanged. Scanning thermal microscopy is used to determine the depth and profile of radiation damage in the material, finding good agreement with simulation. Microstructural causes for the property changes are presented based on scanning electron microscopy, x-ray diffraction, and high-resolution transmission electron microscopy. Through correlation of microstructure and property changes, we gain essential understanding of the nanoscale origin of thermal and thermoelectric property changes due to irradiation. In turn, the design of radiation-tolerant materials and devices for extreme environments is informed.
5:45 PM - ES5.12.05
Effect of Precipitation and Precipitate Distribution on the Strength of Immiscible Cu-W Alloys during Ion Irradiation
Gowtham Sriram Jawaharram 1 , Robert Averback 1 , Shen Dillon 1
1 Department of Materials Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, Illinois, United States
Show AbstractNanostructured metals and alloys are of great interest in the nuclear industry because the high density of interfaces serve as sinks for the point defects created during irradiation. Immiscible multiphase systems are desirable to suppress coarsening during service, which would otherwise reduce the sink density. While high in strength, nanograined alloys often exhibit brittleness that limits their practical application. This work seeks to better understand the relationships between grain size, precipitate distribution, irradiation effects and mechanical properties. We characterized a model highly immiscible Cu-W alloy which can be grown as solid solution films and phase separates in thermal spikes during low temperature irradiation. This allows us to control, somewhat independently the grain size, precipitate size and precipitate density and develop insights into how to simultaneously optimize mechanical properties and sink density. The strength of the films were measured at room temperature as a function of irradiation dose from 0.25dpa to 70dpa. Our results indicate that the onset of precipitation has the greatest effect on indentation hardness and nanopillar yield strength, with the value plateauing for precipitate sizes >1nm and grain size having minimal effect. Select films were either irradiated or annealed at higher temperatures following room temperature irradiation. Through energy dispersive spectroscopy (EDS), it was observed that the films irradiated at higher temperatures had greater grain boundary W concentration and showed greater hardness than the annealed films. Thus, the presence of W solute and precipitates at the grain boundaries also plays a key role in understanding the effects of high temperature irradiation.