Chu Chun Fu, CEA Saclay
Kazuto Arakawa, Shimane University
Sergei Dudarev, Culham Centre for Fusion Energy
Michael Short, MIT
ACS Energy Letters | ACS Publications
ES5.1: What is the Future of High Performance Steels in Nuclear Reactors?
Monday AM, November 28, 2016
Sheraton, 2nd Floor, Back Bay C
9:30 AM - ES5.1.01
Development of Rapidly-Quenched Filler Metals for Brazing of Ferritic-Martensitic Steels for Service in Lead-Cooled Fast Reactors
Boris Kalin 1 , Oleg Sevryukov 1 , Aleksey Suchkov 1 , Alexander Ivannikov 1 , Milena Penyaz 1
1 National Research Nuclear University MEPhI Moscow Russian FederationShow Abstract
Demands on the properties of structural materials are constantly being increased, new alloys are being developed, and the technology of producing permanent joints for these alloys is crucial for advancement of the nuclear industry, especially for the next generation of nuclear reactors.
One of these developing technologies is high temperature brazing, which has several advantages over welding. The major advantages are the homogeneity of the seam and minimization of the influence of thermal effects on joined materials. Currently, nanocrystalline and amorphous filler metals based on nickel, titanium, zirconium and copper, which have unique properties are widely used. An example is the development of joining technology that uses brazing with these filler metals to connect structural elements of fusion reactors (first wall and divertor). Presently this joining method is being developed for a new generation of fast reactors, where heat-resistant, difficult-to-weld, complex-alloyed steels like NT-9, EP-450, EP-823, ChS-139 (12Cr-Si-Mo-W-Ni-V-Nb-C) are being deployed.
Investigation currently focuses on a rapidly-quenched nickel-base alloy (Ni-Cr-Si-Fe-Mo) in the form of a ribbon that was designed for brazing of these steels with a melting interval 1067-1130oC. Production technology of an amorphous ribbon has been optimized. Brazing of layouts of fuel elements (connection: plug-tube) of steels ChS-139 and EP-823 (EP-450) has been conducted. It was established that the shear strength of the brazed layout, exceeds the level of allowable stress for the materials of shells and plugs of the fuel elements. All samples have been observed to fail in the steel and not the joint.
The use of nickel-base filler metals is unacceptable for fast lead-cooled reactors such as BREST-OD-300 due to the low corrosion resistance of nickel alloys in lead. Therefore an iron-based filler metal based on the Fe-Cr-Si-B system with a melting interval 1100-1200oC has been developed, especially for the possibility of making brazed joints of the ferritic-martensitic steel EP-823. The technology of its production in the form of a rapidly-quenched ribbon with amorphous-crystalline structure has been worked out. The optimum brazing mode has also been determined. Strength and corrosion tests, metallographic studies and electron microprobe analysis of the brazed joints have been conducted.
It was established that brazed joints, using iron-base filler metal, have corrosion resistance in lead no worse than that of the base material, and that they can withstand high loads, e.g., tensile strength in uniaxial tension is 750 ± 90 MPa. These results encourage us to recommend the employment of high-temperature brazing using the iron-base rapidly-quenched filler metal as a replacement of currently employed welding techniques, improving the manufacturing technology available for production of new reactor equipment.
The work was supported by the project of improvement the competitiveness of NRNU MEPhI.
9:45 AM - *ES5.1.02
Evolution of Microstructure in Advanced Ferritic-Martensitic Steels under Irradiation—The Origin of Low Temperature Radiation Embrittlement
Sergey Rogozhkin 1 2 , Alexander Nikitin 2 1 , Nikolay Orlov 2 1 , Alexey Bogachev 2 1 , Olesya Korchuganova 2 1 , Andrey Aleev 2 1 , Alexander Zaluzhnyi 1 2 , Timur Kulevoy 2 1 , Rainer Lindau 3 , Anton Moeslang 3 , Pavel Vladimirov 3
1 Moscow Engineering Physics Institute National Research Nuclear University Moscow Russian Federation, 2 Institute for Theoretical and Experimental Physics of National Research Centre, Kurchatov Institute State Scientific Centre of the Russian Federation Moscow Russian Federation, 3 Karlsruhe Institute of Technology Karlsruhe GermanyShow Abstract
Advanced reduced-activation ferritic/martensitic steels and oxide dispersion-strengthened steels exhibit significant radiation embrittlement under low temperature neutron irradiation. Although the phenomenon has been extensively investigated in various materials, including the vessel reactor materials, understanding of embrittlement of f/m steels is unsatisfactory.
Several f\m steels were investigated following neutron irradiation up to ~ 70 dpa. F82H, Eurofer97 and ODS Eurofer were irradiated within the framework of ARBOR projects at 250-340 °C. All the three steels showed large increase in the ductile to brittle transition temperature (DBTT) below 10 dpa. After irradiation to 20-30 dpa the DBTT shift exceeded 200 °C. Ref.  suggests that TEM-visible irradiation defects do not explain the observed hardening of irradiated Eurofer97. Embrittlement of ODS steels is even less well understood.
We focus on atom probe tomography (APT) of Eurofer97 and ODS Eurofer steels irradiated with neutrons and heavy ions at low temperature. TEM studies reveal dislocation loops in the neutron-irradiated f\m steels. At the same time APT shows early stages of solid solution decomposition . High density (1024 m–3) of ~3–5 nm clusters enriched in chromium, manganese, and silicon atoms are found in Eurofer 97 irradiated in BOR-60 reactor to 32dpa at 332°C. In steels, irradiated with Fe ions to the dose of 1.6×1016 ions/cm2, pair correlation functions derived from APT data show Cr-enriched pre-phases. Our APT study of ODS steels irradiated at low temperatures found significant increase in the density of nano-sized clusters. TEM also showed increased density of small (< 5 nm) oxide particles. Nucleation of dislocation loops is suppressed in the regions with high density of nano-oxides.
These results suggest that irradiation-induced nucleation of very small precipitates may represent the main cause of low temperature radiation embrittlement of f\m steels.
1. C. Dethloff, E. Gaganidze, J. Aktaa. Journal of Nuclear Materials 454 (2014) 323–331
2. S.V. Rogozhkin, A.A. Nikitin, A.A. Aleev, A.B. Germanov, A.G. Zaluzhnyi. Inorganic Materials: Applied Research 4 (2013) 112–118.
10:15 AM - ES5.1.03
Effect of Deformation Induced Martensitic Transformation on the Mechanical Properties of Austenitic Steels Irradiated with Neutrons
Mihail Merezhko 1 , Diana Merezhko 1 , Oleg Maksimkin 1 , Frank Garner 2
1 Institute of Nuclear Physics Almaty Kazakhstan, 2 Radiation Effects Consulting Richland United StatesShow Abstract
Strength and ductility of reactor metastable austenitic steels depend on the prior state of the material and are determined by the combined effect of some factors such as radiation and unstable austenitic matrix. Neutron irradiation results in a significant decrease of uniform elongation and increase of material strength. During the deformation at room and low temperatures, additional strengthening can be achieved by forming in the austenite matrix solid plates of α'-martensite. Optimal strengthening of local micro volumes of the material prevents the formation of a stationary neck and increases plasticity. Moreover, second-order effects are those associated with phase stability and consequences of transmutation growing become to first order importance at higher damage levels associated with PWR or VVER plant life extension.
As the object of study, several American (AISI 304, AISI 321, and AISI 316) and the Russian 12Cr18Ni10Ti austenitic steels (analogue of AISI 321) were chosen. These steels vary in element composition such as nickel and manganese. Stacking fault energy as one of the defining parameters of the stability of the austenitic matrix was calculated.
For mechanical testing flat (dimensions of workspace – 7.62×1.52×0.76 mm) and cylindrical (working length - 10 mm, diameter - 1.7 mm) steel samples were machined. Some samples were austenized (1050 °C, 30 minutes) and then irradiated in the core of the WWR-K research reactor at a temperature of about 80 °C to a maximum neutron flux of 1.3x1020 n / cm2 (E> 0,1 MeV). WWR-K is 6MWt Kazakhstan pool type research reactor on the thermal neutrons.
Mechanical testing was performed with strain rate of 0.5 mm / min at room temperature. Simultaneously amount of α'-martensitic phase induced in a sample and its geometric dimensions were registrated.
As a result of experiments mechanical and energy characteristics of the materials were calculated. The influence of the chemical composition, structure and neutron radiation on the strength and ductility of austenitic steels were determined. Accumulation diagrams of martensitic α'-phase were built and the parameters characterizing the kinetics of martensitic transformation were calculated.
It is shown that neutron irradiation lowers the ductility and increases the strength of all test materials. Adding nitrogen and manganese provides a high level of strength characteristics of the steel by solid solution hardening and a decrease in ductility.
It was revealed that the martensitic transformation is much more difficult with increasing content of nickel and manganese in the matrix. The neutron irradiation shifts the beginning of martensitic transformation to lower deformation values.
10:30 AM - *ES5.1.04
Neutron Irradiated Ferritic-Martensitic 9Cr Steels—Effect of Helium and Nanoscaled ODS Particles on Microstructure, Tensile and Fatigue Properties
Anton Moeslang 1 , Ermile Gaganidze 1 , Michael Klimenkov 1 , Rainer Lindau 1 , Rolf Rolli 1 , Hans-Christian Schneider 1
1 Institute for Applied Materials Karlsruhe Institute of Technology Karlsruhe GermanyShow Abstract
A good understanding of structure-property correlations in neutron irradiated advanced steels is indispensable for design, lifetime prediction and economical operation of fast breeder and future fusion power reactors. Irradiation campaigns on ferritic-martensitic 9CrlWTaV steels (mostly EUROFER) have been performed at the HFR reactor in The Netherlands up to an accumulated damage dose of 16.3 dpa at irradiation temperatures between 250 and 450°C. In the first campaign 80-5100 appm He has been produced by 10B doping to analyze after irradiation systematically Helium embrittlement at low and high strain rates. While tensile and fatigue tests at low stain rates show no Helium effect up to ~400 appm, Charpy tests at high strain rates reveal a helium related pronounced increase in DBTT of more than 100°C already at 80 appm He. In the second campaign 50 kg EUROFER-ODS was fabricated by powder metallurgy. Compared to EUROFER and other 9CrlWTaV steels the ODS-steel shows in the entire temperature range convincing tensile and fatigue properties after neutron irradiation: much less irradiation hardening, unprecedented uniform elongation and no dislocation channeling induced strain localization. It is to our knowledge the first time that results on neutron irradiated ODS-steels are presented in such a wide temperature range. Correlations are made between strengthening, plasticity, cyclic softening, fatigue lifetime and fracture behavior on the one hand and microstructure evolution on the other hand. Finally conclusions are drawn regarding the microstructural engineering of irradiation tolerant materials.
ES5.2: Can Nuclear Materials Design Prevent Nuclear Accidents?
Monday AM, November 28, 2016
Sheraton, 2nd Floor, Back Bay C
11:45 AM - *ES5.2.01
Accident Tolerant Fuel Cladding for Light Water Reactor
Kumar Sridharan 1
1 University of Wisconsin, Madison Madison United StatesShow Abstract
Zirconium-alloys have been used successfully as materials for fuel claddings in Light Water Reactors (LWR) for many decades. However, in a loss of coolant event, the rise in temperature can result in severe exothermic oxidation of zirconium alloy in conjunction with hydrogen generation from reaction between zirconium and steam. This scenario was tragically borne out during the Fukushima accident in Japan in 2011. In response to this accident, efforts are underway world-wide to develop LWR fuel claddings that are more resistant to oxidation at high temperatures. Approaches being investigated include total replacement of zirconium-alloy cladding with an FeCrAl alloy, molybdenum, and SiC-SiC composite or deposition of oxidation coatings (e.g., Cr, FeCrAl, Ti-Al-C MAX phase compound, and transition metal silicides) on the present zirconium-alloy. Each approach presents its own set of opportunities and challenges that involve considerations of oxidation resistance, high temperature mechanical behavior, manufacturability, and neutronics. The presentation will highlight the advances in these various approaches to the development of accident tolerant fuel cladding.
12:15 PM - ES5.2.02
Can CRUD Enhance the Thermal-Hydraulics Performance of Nuclear Reactors
Carolyn Coyle 1 , Matteo Bucci 1 , Tom McKrell 1 , Jacopo Buongiorno 1
1 Nuclear Science and Engineering Massachusetts Institute of Technology Cambridge United StatesShow Abstract
CRUD, named for Chalk River Unidentified Deposits, is a hydrophilic porous layer that collects on the surface of nuclear fuel pins during normal operation. It is composed primarily of nickel oxide (NiO), iron oxide (Fe3O4), and trevorite (NiFe2O4). These oxides are corrosion products of the reactor coolant system that are transported as particles throughout the reactor primary loop. As nucleate boiling occurs on fuel rods, the oxide particles precipitate forming CRUD onto the fuel pins. CRUD can grow to 10-100 μm thick with average pore size, porosity, and roughness values between 0.1-1 μm, 40-50%, and 0.5-3.0 μm, respectively. CRUD also contains characteristic boiling chimneys, which are formed as water is pulled into the porous layer through capillary wicking, evaporates and escapes through the chimneys back into the coolant flow.
Although CRUD effects on corrosion and reactivity have been studied extensively, its effects on reactor thermal-hydraulic performance are not well understood. Typically, CRUD buildup is thought to be a purely negative by-product of harsh nuclear environments that increases fuel’s effective thermal resistance and temperatures. However, CRUD is a naturally occurring hydrophilic porous layer that, in surface engineering studies, has been found to enhance CHF and increase the Leidenfrost temperature. Therefore, in thin layers, CRUD has the potential to postpone the boiling crisis in loss of flow accidents (LOFAs) or transient overpower, and accelerate quenching heat transfer following a loss of coolant accidents (LOCA).
In this work, we present the results of a recent investigation of such effects conducted in the MIT Nuclear Science and Engineering department. The boiling performances of CRUD’ed surfaces have been investigated by preparing synthetic CRUD on indium tin oxide-sapphire heaters and testing them in pool boiling and flow boiling conditions. Reactor CRUD was simulated using layer-by-layer deposition of 100 nm silica nanoparticles to form carefully engineered porous, hydrophilic thick films. Photolithography was used to manufacture posts that were then dissolved to create chimneys characteristic of actual CRUD. Synthetic CRUD thickness, wettability, pore size, and chimney diameter and pitch were verified to be representative of reactor CRUD. During testing, IR thermography and high-speed video have been used to shed light on the boiling mechanisms. Measurements enabled by such diagnostics include: time dependent distributions of temperature and heat fluxes on the boiling surface, typically with a <100 µm space resolution and ~1 ms time resolution, nucleation site density, bubble departure frequency and diameter.
Pool boiling experiments have revealed that heaters with synthetic CRUD may have up to 100% enhancement in CHF, while minor effects were experienced in the heat transfer coefficient in both pool boiling and flow boiling conditions. CHF in flow boiling conditions is currently being investigated.
12:30 PM - ES5.2.03
Weld Development for Thin-Walled Cladding for LWR Accident Tolerant Fuel
Jian Gan 1 , Nathan Jerred 1 , Emmanuel Perez 1 , Haiming Wen 1 , D.C. Haggard 1
1 Idaho National Laboratory Idaho Falls United StatesShow Abstract
The development of new cladding materials for light water reactors (LWR) has been seriously considered with improved high temperature oxidation resistance to minimize hydrogen production from an accident such as a Loss of Coolant Accident. FeCrAl alloy has been selected as a candidate material for its improved properties on oxidation resistance over the present LWR zircaloy cladding. Because of higher thermal neutron absorption for FeCrAl alloy, the cladding wall thickness has to be reduced down to ~ 350 µm that is significantly thinner than that of the current zircaloy cladding in LWR. New weld techniques of joining the endplug to cladding need to be developed. This presentation will summarize the recent work on the weld development for thin-walled FeCrAl cladding. The microstructural characterization of the weld-affected zone and mechanical property investigation of the weld zone will be discussed.
This work has been focused on joining the emulated FeCrAl thin-walled cladding to the end plug. Both laser beam weld and pressure resistance weld are investigated. The former is a fusion-based weld technique with high intensity pulsed laser to minimize the heat affected zone while the latter is a solid state joining without melting to avoid significant change to the original microstructure. For laser weld, it requires optimization to ensure sufficient heat penetration with minimum material loss from ablation. The pressure resistance weld requires securing the contact through breaking the surface oxide and small amount of deformation followed by a rapid and uniform heating at the contact with minimum additional deformation on the cladding. Both techniques show promising results based on the tensile tests. The characterization of the weld including X-ray 3D tomography for nondestructive examination, optical microscopy and SEM for cross-section samples, tensile test for the joined cladding-endplug set, micro hardness test for weld zone of the cross-section samples, EBSD analysis for grain texture, and hydraulic pressure burst test of the jointed cladding-endplug set. Post-weld heat treatment reveals significant improvement on mechanical property of the weld.
12:45 PM - ES5.2.04
Designing the Ultimate Fouling Resistant Material
Max Carlson 1 , Ittinop Dumnernchanvanit 1 , Robert Simpson 2 , Reid Tanaka 1 , Michael Short 1
1 Massachusetts Institute of Technology Cambridge United States, 2 SUTD Tampines SingaporeShow Abstract
The particulate fouling of heat exchanger and pipe surfaces in operational power plants is problematic for continued efficient operation. Fouling deposits on the fuel of nuclear reactors, referred to as CRUD, cause additional operational concerns such as axial power shifts and accelerated corrosion. The adhesion of fouling deposits to surfaces is partially defined by the attractive van der Waals (VDW) force, the magnitude of which is given by the Hamaker constant. Therefore, developing materials with minimized Hamaker constants should minimize VDW bonding, subsequently reducing or eliminating fouling at its source.
The Hamaker constant is defined by the strength of electronic interactions between two solids through an intermediating fluid. Theoretically, the interaction of any of the two solids with the mediating fluid can define the magnitude of the Hamaker constant. This magnitude is defined by the similarities of the dielectric spectra of each material and the fluid, which can be approximated by their visible indices of refraction. Therefore, developing a material which dielectrically mimics its surrounding fluid should theoretically lead to a Hamaker constant of zero, no VDW attraction, and reduced fouling.
In the present study, we describe theoretical, computational, and experimental efforts to design material surfaces which mimic high temperature, high pressure water for fouling resistance in nuclear and geothermal plants. The VDW force that enables fouling particles to adhere to surfaces is measured using atomic force microscope force spectroscopy (AFM-FS), and coatings that are expected to minimize this force while retaining stability at operating conditions are suggested. We focus on the fluorine-containing family of minerals, whose calculated Hamaker constants are two orders of magnitude lower than those of the oxides found in nuclear and geothermal energy systems. The theoretical prediction of the adhesive force is compared to AFM-FS measurements in water, nitrogen, and vacuum. The VDW force is shown to be calculable directly from density functional theory (DFT) simulations of materials, opening the way to computational discovery of new fouling-resistant materials by a genetic algorithm approach.
ES5.3: Nuclear Materials at Ultrahigh Temperatures
Kang Pyo So
Monday PM, November 28, 2016
Sheraton, 2nd Floor, Back Bay C
2:30 PM - *ES5.3.01
Modeling the Behavior of Tungsten in the Divertor Conditions
Charlotte Becquart 1 , Andree De Backer 2 , Christophe Domain 3
1 Lille University of Science and Technology Villeneuve D Ascq France, 2 Culham Centre for Fusion Energy Culham United Kingdom, 3 EDF Ramp;D Moret sur Loing FranceShow Abstract
Tungsten is a candidate material for the divertor and for first wall armour of future thermonuclear fusion reactors (ITER and DEMO). In the near surface of plasma facing materials, high densities of interstitials and vacancies are produced in addition to the implantation of high concentrations of helium as well as hydrogen isotopes from the plasma. These defects and gases will induce changes in the microstructure and can lead to dramatic changes in materials properties. In this perspective, the fate of irradiation induced defects and their interactions with helium atoms and hydrogen isotopes has to be understood. We use a combination of modeling techniques, namely density functional theory calculations, molecular dynamics and object kinetic Monte Carlo methods, which span different time and length scales to tackle the problem. In this work we show how modeling and carefully designed experimental investigations can provide a route to the understanding of the microstructure evolution of materials in these conditions. In particular, we will address self trapping trap mutation for helium atoms and the formation of helium bubbles, the formation of nanovoids in the near surface region of 800 keV 3He implanted tungsten, and the trapping of H atoms by self interstitials and dislocation loops.
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. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
3:00 PM - ES5.3.02
Extreme Ion Irradiation of Oxide Nanoceramic Coatings
Francisco Garcia Ferre 1 , Alexander Mairov 2 , Luca Ceseracciu 1 , Daniele Iadicicco 1 , Matteo Vanazzi 1 , Yves Serruys 3 , Frederic Lepretre 3 , Marco Beghi 4 , Lucile Beck 3 , Kumar Sridharan 2 , Fabio Di Fonzo 1
1 Istituto Italiano Tecnologia Milano Italy, 2 University of Wisconsin-Madison Madison United States, 3 CEA Saclay France, 4 Politecnico di Milano Milano ItalyShow Abstract
Oxide nanoceramics combine the enhanced radiation tolerance of nanocrystalline materials with the chemical inertness of oxides, and are promising coating materials for highly corrosive and intense radiation environments. In this work, the properties of oxide nanoceramic coatings are evaluated as radiation damage approaches extreme levels, reaching or even exceeding those anticipated for advanced nuclear systems. In particular, taking Al2O3 as a model coating material, an effort is made to establish a correlation between the irradiation spectrum and the observed evolution of the oxide’s structural features and mechanical properties. The irradiation conditions are relevant for advanced nuclear systems, and consist in low, moderate, high and extreme radiation damage levels at 600°C –namely, 20, 40, 150, 250 and 450 displacements per atom. A comprehensive analysis of the irradiated nanoceramic is accomplished by X-Ray Diffractometry (XRD), Transmission Electron Microscopy (TEM), Scanning-TEM (STEM), and nanoindentation. The results confirm that grain growth is the main structural change induced by irradiation in oxide nanoceramics. This structural change manifests mechanically through an initial increase of hardness that is well fitted by the Hall-Petch relationship, and eventually through softening in the extreme damage range. Stiffness increases sub-linearly with damage before reaching a plateau. Further, both hardness and stiffness depend on the phase present. The phase evolution may be depth-dependent, and depends strongly on the ion utilized and on the irradiation spectrum. To conclude, it is shown that the coatings successfully provide negligible hydrogen permeation and superior corrosion resistance to structural steels at high temperature in chemically aggressive media. All these features are of particular interest for accident tolerant fuels, Generation IV reactors and fusion systems.
3:15 PM - ES5.3.03
Investigation of High-Temperature Oxidation Kinetics and Residual Ductility of Oxidized Samples of Sponge-Based E110 Alloy Cladding Tubes
Yong Yan 1 , Benton Garrison 1 , James Keiser 1 , Mike Howell 1 , Tyler Smith 1 , Gary Bell 1
1 Oak Ridge National Laboratory Oak Ridge United StatesShow Abstract
Fuel rod cladding is the first barrier for retention of fission products and the gross structural integrity of the cladding ensures coolable core geometry. To ensure adequate cladding performance during Emergency Core Cooling System re-flooding and during loss-of-coolant-accidents (LOCA), the current NRC licensing criteria limits the peak cladding temperature and the maximum cladding oxidation during the accident. The purpose of these limits is to prevent cladding embrittlement during a LOCA to mitigate the potential for subsequent catastrophic widespread cladding fracture. The purpose of this work is to obtain experimental data on the kinetics of high-temperature oxidation and residual ductility of sponge-based E110 cladding, and provide the data needed to assess the effects of hydrogen pickup on cladding embrittlement during the accident.
Two-sided oxidation experiments were recently conducted at 1000-1200°C in flowing steam with samples of sponge-based Zr-1Nb alloy E110. Although the old electrolytic E110 tubing exhibited a high degree of susceptibility to nodular corrosion and experienced breakaway oxidation rates at relatively low test times, the new sponge-based E110 has been demonstrated as having steam oxidation behavior comparable to Zircaloy-4. The sponge-based E110 followed the parabolic law, and the derived oxidation rate constant is in good agreement with the Cathcart-Pawel (CP) correlation at 1100-1200°C. For 1000°C oxidation, the weight-gain of sponge-based E110 is much lower than Zircaloy-4. No breakaway oxidation was observed at 1000°C up to 8000s. Ring compression tests were conducted to evaluate residual ductility of oxidized samples at room temperature and at 135°C. All sponge-based E110 were still ductile at 135°C after being oxidized up to 20% equivalent cladding reacted at 1000-1200°C. Microhardness tests, metallographic examinations, and scanning electron microscopy examinations were performed on oxidized E110 specimens to correlate the material performance to their microstructures.
In addition, a post-quench ductility study was conducted at 1200°C with tubing specimens of Zircaloy-4, 347 stainless steels, and the commercial FeCrAl alloy APMT under design basis LOCA and extended LOCA conditions. The parabolic rate constant for Zircaloy-4 tubing was in excellent agreement with the CP correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR=50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to four hours.
3:30 PM - ES5.3.04
Predicting Macroscopic Material Properties from Microstructures for Functional Nuclear Materials
Llion Evans 1 2 , Elizabeth Surrey 1 , Lee Margetts 2
1 Culham Centre for Fusion Energy Abingdon United Kingdom, 2 School of Mechanical, Aerospace and Civil Engineering University of Manchester Manchester United KingdomShow Abstract
The nuclear sector is currently undergoing a renaissance with 160 new fission reactors being built and another 300 planned. The move to Gen-IV and fusion reactors aims to improve on current technology by increasing output and efficiency but will entail creating more extreme environments within the reactors putting added demands on materials. To this end, novel functional and structural materials are being developed capable of withstanding such extremes. Functional materials are often designed to have properties which make them highly suitable for a particular task but can lead to complex material properties which are difficult to predict. During fabrication of the materials small variations in parameters (e.g. heating/cooling rates, alloying percentages etc.) can yield drastic differences in material properties. It is therefore easily possible to create a large number of new materials quickly but characterising them so that they’re well understood, and therefore useful, can be a lengthy and costly process.
The performance of some of the functional materials proposed (e.g. foams and composites) is dominated by their complex and often anisotropic micro-features. This paper uses the ‘image-based finite element method’ (IBFEM) to perform virtual characterisation of thermal properties. By this method an X-ray tomography scan of graphite foam is converted into an ultra-high resolution FEM model which captures the micro-features. Samples are first digitally cut and prepared from a parent block. The virtual samples are used in an in silico reproduction of the laser flash laboratory experiment. This enables measurement of the material’s thermal diffusivity. Variation in the results due to orientation of the sample demonstrates the power of IBFEM to capture anisotropy due to micro-features.
This method could be expanded to a full suite of laboratory experiments, paving the way to fast virtual material characterisation which will increase the turnaround of new material development.
3:45 PM - ES5.3.05
Electrically-Assisted Forming of Mechanical Property Characterization of Oxide Dispersion Strengthened Structural Materials
Zilin Jiang 1 , Adrien Aubrun 1 , Man-Kwan Ng 1 , Qiang Zeng 1 , Kornel Ehmann 1 , Osman Anderoglu 2 , S. Maloy 2 , Jian Cao 1
1 Department of Mechanical Engineering Northwestern University Evanston United States, 2 Los Alamos National Laboratory Los Alamos United StatesShow Abstract
Oxide Dispersion Strengthened (ODS) ferritic alloys have been considered as candidate materials for nuclear applications due to their excellent resistance to radiation damage, high temperature strength and creep resistance. However, the application of ODS ferritic alloys are limited by the low formability. The maximum uniform elongations of ODS ferritic alloys under 200°C before necking occurs are limited to 5-7%, which poses a significant challenge in shaping these alloys for nuclear reactor applications. Conventionally, to deform such high-strength materials with limit ductility, it requires the workpiece to be heated into the temperature region for either warm forming or hot forming. The thermal softening effect decreases the yield strength of the materials and increases ductility. However, such high temperature can decrease the effectiveness of tools and dies, and is generally very expensive and inefficient.
Here, we propose a new idea to deform ODS ferritic alloys, i.e., applying electric current through workpieces during the plastic deformation, which is known as Electrically-Assisted (EA) forming. In EA forming, constant or pulsed electric current passes through the workpiece during the deformation process to increase the formability of high-strength materials that are difficult to be deformed using conventional forming techniques.
We experimentally investigate the mechanical behavior of ODS 14YWT steel subject to EA deformation. The results are compared with conventional thermal tension tests. The flow stress of materials can be decreased significantly by increasing the passing current density. For example, at current densities of 18A/mm2, 30A/mm2, and 35A/mm2, the flow stress is reduced by 12.45%, 35.20%, and 48.73%, respectively from room temperature (1100MPa) while the maximum temperature is at 151°C, 383°C, and 736°C, respectively. The resulting microstructures will be examined and reported. Taking advantage of the instant softening behavior under EA-forming, an EA-tube drawing process is proposed to form hollow tubes made of ODS 14YWT steel. A new tubing machine will be designed and fabricated to perform the tube drawing experiments.
ES5.4: Designing Materials for Extreme Radiation Resistance
Monday PM, November 28, 2016
Sheraton, 2nd Floor, Back Bay C
4:30 PM - ES5.4.01
Mechanisms of Radiation Damage Reduction in Equiatomic Solid Solution Alloys
Kai Nordlund 1 , Fredric Granberg 1 , Flyura Djurabekova 1 , Yanwen Zhang 2 , William Weber 2
1 University of Helsinki Helsinki Finland, 2 Oak Ridge National Laboratory Oak Ridge United StatesShow Abstract
Equiatomic solid solution alloys – solid solution metals of several principal elements in roughly equal elemental fractions – are promising candidate materials for advanced nuclear reactor components due to their high radiation resistance. We have studied by means of molecular dynamics simulations the radiation effects in pure Ni and Ni-based fcc alloys with an increasing number of alloying components, up to four in equal elemental fractions. Our combined experimental and simulation results show that several equiatomic metal alloys are more resistant to radiation damage than the corresponding pure elements.Moreover, our analysis of the underlying mechanism of different equiatomic alloys establishes that alloy effects on significant reduction of dislocation mobility is generic and not specific to the number of elements in the system 
 F. Granberg, K. Nordlund, M. W. Ullah, K. Jin, C. Lu, H. Bei, L. M. Wang, F. Djurabekova, W. J. Weber, , and Y. Zhang, Phys. Rev. Lett. 116, 135504 (2016)
4:45 PM - ES5.4.02
Dispersion of Carbon Nanotubes in Metals Improves Strength and Radiation Resistance
Kang Pyo So 1 , Di Chen 2 , Akihiro Kushima 1 , Mingda Li 1 , Sangtae Kim 1 , Yang Yang 1 , Ziqiang Wang 1 , Jong Gil Park 3 , Young Hee Lee 3 , Rafael Gonzalez 4 , Miguel Kiwi 4 , Eduardo Bringa 5 , Lin Shao 2 , Ju Li 1
1 Massachusetts Institute of Technology Cambridge United States, 2 Texas Aamp;M University College Station United States, 3 Sungkyunkwan University Suwon Korea (the Republic of), 4 Universidad de Chile Santiago Chile, 5 National University of Cuyo Mendoza ArgentinaShow Abstract
One-dimensional carbon nanotubes (CNTs), which are mechanically strong and flexible, possess better structural properties than their counterpart nanoparticles and other one-dimensional nanowires. We report a nano-dispersion strategy of CNTs in Al matrix that leads to enhance mechanical strength, irradiation resistance with a tenable ductility. This was realized by intragranular dispersion of CNTs from atomic-diffusion driven cold-welding. The uniformly dispersed intragranular CNTs inside metal as 1D fillers (high aspect ratio) create prolific internal interfaces with the metal matrix that act as venues for the radiation defects to recombine (self-heal), reducing void/pore generation and radiation embrittlement at high displacements per atom (DPA). Under helium ion irradiation up to 72 DPA, the 1D carbon nanostructures survive, while sp2 bonded graphene transforms to sp3 tetrahedral amorphous carbon. Self-ion (Al) irradiation converts CNTs to a metastable form of Al4C3, but still as slender 1D nanorods with prolific internal interfaces that catalyze recombination of radiation defects, reducing radiation hardening and porosity generation. The 1D fillers may also form percolating paths of “nano-chimneys” that outgas the accumulated helium and other fission gases, providing an essential solution to the gas accumulation problem. We further observed nanoscale plasticity and rupturing processes near CNTs by in-situ mechanical tests inside TEM which can contribute to minimize the radiation embrittlement. CNT in the middle of the grain prevents the rupture tip propagation after necking (called Multi-step Rupturing Process (MRP)), which delays the final fracture.
Therefore, our observation of dispersion mechanism of CNT, MPR process and improved radiation resistance deepens scientific understanding and accelerates researches on various CNT-reinforced metal composites.
5:00 PM - ES5.4.03
Study of Swelling Accumulation in Austenitic Steels at High Irradiation Doses
Stanislav Golubov 1 , Alexander Barashev 1
1 Oak Ridge National Laboratory Oak Ridge United StatesShow Abstract
The modern radiation damage theory of structural materials predicts such experimentally observed phenomena as void and gas bubble super lattice formation, grain boundary and grain size effects etc. In addition it predicts a saturation of void size at very small magnitude in the case when void spatial distribution is random. This phenomenon takes a part in a typical swelling behavior in austenitic steels that normally has the three stages: incubation, transient and steady state. In this talk the effect of void size saturation on swelling accumulation in the dose range up to 100 dpa is studded by use a code RIME (radiation-induced microstructure evolution) based on a mean-field kinetic model, which account for cascade production of one-dimensionally migrating self-interstitial clusters in addition to that of three-dimensionally migrating single vacancies and interstitials. It is shown that the incubation stage is fully determined by the void size saturation. Calculation results obtained for swelling accumulation in austenitic steels at Light Water Reactor conditions are compared with experiment data.
5:15 PM - ES5.4.04
Progress toward Structural Alloys by Design—Compositional Effects on Defect Production and Damage Evolution
Yanwen Zhang 1 2 , G. Malcolm Stocks 1 , Hongbin Bei 1 , Haizhou Xue 2 , Gihan Velisa 1 , Ke Jin 1 , Mohammad Wali Ullah 1 , Shijun Zhao 1 , Laurent Beland 1 , German Samolyuk 1 , Chenyang Lu 3 , Lumin Wang 3 , William Weber 2 1
1 Oak Ridge National Laboratory Oak Ridge United States, 2 University of Tennessee Knoxville United States, 3 University of Michigan Ann Arbor United StatesShow Abstract
As one of the oldest sciences, alloy development has been focused on traditional alloys with one or two principal elements and minor alloying elements. For nuclear applications, alloy design for enhanced radiation resistance has focused on microstructural or nano-scale features to mitigate displacement damage. In sharp contrast to traditional alloys, recent advances for 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. Integrated experimental and modeling studies in a unique set of fcc CSAs with the number of constituent elements varying from 2 to 5 clearly demonstrate links between intrinsic material properties and various dynamic processes. The focus on model fcc alloys eliminates possible confusion or entangled complications from variations of crystal structure, multiphases, or nano-scale features. The systematic variation in the number and concentration of components, as well as the chemical elements, makes this a valuable tactic to de-convolute the effects at electronic and atomic levels. The recent progress suggests a novel concept of how intrinsic chemical disorder can affect defect dynamics at their early stages (low-dose irradiation cases). This has made it possible to build a much-needed knowledge base that may enable reductions in damage accumulation at a later stage (high-dose irradiation cases). The insights into defect dynamics may provide an innovative path forward towards in solving the long-standing challenge in alloy development to achieve much enhanced radiation resistance.
Research supported by EDDE, a DOE-BES Energy Frontier Research Center.
5:30 PM - ES5.4.05
Design of Model Radiation-Resistant Alloys
Thomas Schuler 1 , Dallas Trinkle 1 , Pascal Bellon 1 , Robert Averback 1
1 University of Illinois at Urbana Champaign Urbana United StatesShow Abstract
The design of nanoscale microstructures is a promising way to develop radiation-resistant alloys. Since these microstructures are obtained under out-of-equilibrium processing routes they may be unstable under irradiation. An alternative to nanopatterning is to introduce specific solutes which show attractive binding with point defects. These solutes would act as recombination centers, reducing the driving force for void and self-interstitial loop formation. Setting the solute concentrations below solubility limits ensures the stability of the solid solution in most cases. We combine various techniques to study the thermodynamic and kinetic properties of FCC materials (Cu and Al) with added solutes. Starting from the electronic scale, we use density functional theory to calculate binding and migration energies of point-defects and solutes in various local chemical environments. Then we apply the self-consistent mean-field formalism to compute transport coefficients that contain all the kinetic information of the system. Finally we include these transport coefficients in homogeneous rate equations to study steady state point defect concentrations as a function of various parameters. This multi-scale modeling process predicts the efficiency of point defect trapping under different temperature/radiation flux conditions, guiding the design of dedicated experiments.
5:45 PM - ES5.4.06
Natural Recovery of Alpha-Decay Induced Damage in Actinide Dioxides
Yehuda Eyal 1
1 Department of Chemistry Technion-Israel Institute of Technology Haifa IsraelShow Abstract
At ambient temperature, self-induced alpha-decay damage in actinide dioxides and ancient analogues, thorianite and uraninite, (Th,U)O2, is limited to lattice dilatation of 0.3 to 0.5% even after prolonged exposure to high alpha-decay doses. It is widely accepted that ~1,500 fresh Frenkel pairs are created per alpha-decay event. Most of this damage is attributed to the alpha-recoil atom, traveling a distance of ~20 nm. Here we review two proposed damage recovery mechanisms.
The more important mechanism is radiation annealing - prompt athermal recombination of randomly occurring, closely spaced interstitial-vacancy pairs. Defect ingrowth is modeled as event-by-event production of local damage zones, and simultaneous elimination of close defect pairs. Recombination is governed by a critical interstitial-vacancy annihilation distance. The needed guided defect mobility is controlled by forces that originate from local electrostatic and mechanical distortions. Data pertaining to 18 different actinide dioxide samples yield damage zones having a mean volume of ~3,200 unit cells, and a mean concentration of ~0.1 Frenkel pairs per unit cell at damage saturation. The validity of the model has been confirmed by molecular dynamic simulations on UO2 at < 5 K. Vacancy and interstitial at closest separation and at some more distant sepatations combine spontaneously in both the uranium and oxygen sublattices.
The second recovery mechanism involves thermal diffusion. Before affected by a subsequent alpha-decay event, a local saturated damage region survives for some time before vanishing into the background of more dispersed damage. This mechanism clearly manifests itself under heating at high temperatures. Evidence at ambient temperature has been provided by leaching of 3 thorianite and uraninite specimens characterized by slow damage rates and lifetime displacements per atom yields of 140 to 450. Annealing is evident by highly enhanced and moderately enhanced initial leaching rates of the short-lived 228Th and the long-lived 230Th, respectively, relative to the release rate of the primary very long-lived 232Th. These 3 nuclides are expected to reside in fresh, partly annealed and annealed damage, respectively. Similarly, the long-lived 234U exhibited moderately enhanced leaching rate relative to the leach rate of its primary very long-lived 238U. It is likely that the ratio of the measured 228Th/232Th and 234U/238U activity ratios in excess of congruent isotopic leaching is given by the ratio of the equilibrium fractions of 228Th and 234U in residual damage. These fractions depend upon the nuclides decay constants and the damage elimination rate, assumed to decrease exponentially with time. This study yields a mean annealing time of ~15 kyr.
Testing the integrity of freshly-recycled UO2/PuO2 reactor fuels under damage created by high loads of minor actinides, and understanding long-term alpha-decay effects in spent UO2/PuO2 reactor fuels are of current interest.
Chu Chun Fu, CEA Saclay
Kazuto Arakawa, Shimane University
Sergei Dudarev, Culham Centre for Fusion Energy
Michael Short, MIT
ACS Energy Letters | ACS Publications
ES5.5: Computational Nuclear Materials Science
Tuesday AM, November 29, 2016
Sheraton, 2nd Floor, Back Bay C
9:15 AM - ES5.5.01
The Effect of Overlapping Cascades on Damage Production and Evolution in Concentrated Solid Solution Alloys
Mohammad Wali Ullah 1 , Yanwen Zhang 1 , William Weber 1 2
1 Oak Ridge National Laboratory Oak Ridge United States, 2 University of Tennessee Knoxville United StatesShow Abstract
Structural materials play an important role in safe, sustainable and economical operation of nuclear power systems. The structural materials for Generation IV concept fission reactors and tokamak-like fusion power plants will be subjected to unprecedented fluxes of high energy neutrons and extreme operating temperature. The development of new high performance materials that can withstand such hostile conditions is, therefore, critical for advanced fission and fusion reactor systems, as well as for high-energy physics. In the search for new materials, a new class of single-phase, concentrated solid-solution alloys (SP-CSAs), might play an important role as structural material for nuclear applications where high radiation tolerance is a primary concern. In this work, we have investigated prolonged irradiation of fcc Ni0.8Fe0.2 and Ni0.8Cr0.2 alloys using molecular dynamics (MD) simulations to study damage accumulation in these model SP-CSAs, as compared with pure fcc Ni. Recoil energies of 5 and 25 keV are used. The Ni0.8Cr0.2 and Ni0.8Fe0.2 alloys exhibit higher radiation resistance compared to Ni, and Ni0.8Cr0.2 has better radiation performance than Ni0.8Fe0.2. In the 25 keV cascade simulations, Ni0.4Fe0.4Cr0.2 is included and compared with the other alloys. For the higher-energy recoil events, the difference in Frenkel pair production between the two binaries is less pronounced; while in the Ni0.4Fe0.4Cr0.2 alloy, defect production is about 1.4 and 2 times lower than the binaries and pure Ni, respectively. While large interstitial clusters are formed in Ni, much smaller interstitial clusters are produced in the alloys with Cr. Due to the low mobility of vacancies on the MD time scales, they are found primarily as single point defects and small clusters in all materials.
Research supported by EDDE, a DOE-BES Energy Frontier Research Center.
9:30 AM - *ES5.5.02
The Value of Computational Materials Science for the Design and Operation of Nuclear Power Reactors
Erich Wimmer 1
1 Materials Design S.A.R.L. Montrouge FranceShow Abstract
A deep understanding of the relationships between chemical composition, structure, and materials properties in the harsh environment of nuclear reactors is of fundamental importance for their safe and sustainable operation. Computational materials science offers unprecedented possibilities to elucidate mechanisms and to predict materials properties as will be reviewed in this talk. Three key factors are driving the progress in computational materials science, namely (i) theoretical concepts such as density functional theory, (ii) computational algorithms and their implementation in software, and (iii) computer hardware. While theory advances steadily on the time-scale of half-centuries, software evolves on the scale of decades, and hardware progresses at a relentless pace of few years. The total impact of computational materials science arises from a combination of these three factors. Thus, the potential value is increasing exponentially. The following examples illustrate current capabilities and challenges. (i) Radiation-induced dimensional changes of Zr alloys can lead to effects such as channel distortion. Simulations reveal that preferential diffusion of self-interstitial atoms lead to the formation of nanoclusters, which deform the surrounding material anisotropically, thus providing an interpretation of experimentally observed radiation-induced growth. Alloying elements such as Nb have a significant influence on these diffusion processes, thereby altering the response of the material to radiation. (ii) With improved corrosion resistance of Zr alloys, hydrogen pick-up and the formation of hydrides with the associated degradation of the mechanical properties are becoming an increasing concern. Computational investigations are helping to unravel the complex mechanisms by allowing detailed simulations of specific scenarios at the water/oxide and oxide/metal interfaces. (iii) Computations provide valuable understanding of the degradation of high-performance alloys. For example, the formation of a Ni2Cr long-range ordered phase in Ni-Cr alloys causes embrittlement. Computations are revealing the key driving mechanisms, thus setting the stage for mitigation strategies. Atomistic simulations have matured to the point where reliable quantitative predictions of certain materials properties have become possible. This offers a valuable source of materials data such as elastic coefficients, permeability of gases such as hydrogen, segregation energies to grain boundaries, and diffusion rates of alloying elements, impurities, and defects. In addition, simulations are valuable also in the prediction of properties of fluids including density, viscosity, diffusivity and thermal conductivity as needed, for example, in the design and optimization of molten salt reactors. An assessment of the current capabilities and perspectives for future research and development will conclude this contribution.
10:00 AM - ES5.5.03
Nano Size Effect of Ion Implantation—The Necessity of 3D Simulations
Yang Yang 1 , Yonggang Li 1 2 , Michael Short 1 , Ju Li 1
1 Massachusetts Institute of Technology Cambridge United States, 2 Key Laboratory for Materials Physics Institute of Solid State Physics, Chinese Academy of Sciences Hefei ChinaShow Abstract
Ion irradiation is widely used to study the radiation damage of materials for nuclear energy applications, since it costs much less time than neutron irradiation for the same level of displacements per atom (DPA). One-dimensional (1D) SRIM simulation is often performed to estimate the defects distribution by ion irradiation, based on the assumption that target size is much larger than the penetration depth of ion, which usually ranges from a few nanometers to a few microns. However, recent studies push target size down to nano-scale. For example, nanostructured materials, such as nanoporous materials, multilayer nanocomposites and oxide dispersion strengthened (ODS) alloy are promising to be the next generation of structural materials in the nuclear reactors, in that high volume fractions of super sink interfaces facilitate self-healing after radiation damage. For nano-target ion irradiation, full three dimensional (3D) simulations become necessary. Using a recently developed 3D Monte Carlo simulation code for Ion irradiation in Matters (IM3D), we quantify the error of 1D approach in three classic applications of the nano-scale ion implantation. Because 1D approach fails to consider ion exchange at boundaries, a large discrepancy is found between 1D approach and 3D approach. Then by visualization of the 3D defects distribution created by a nano-beam, we discovered a topological evolution for the first time. We conclude that 3D simulation is necessary for nano-scale ion implantation.
10:15 AM - ES5.5.04
Non-Local Modelling of Dislocation Climb and Cavity Growth
Iacopo Rovelli 1 2 , Sergei Dudarev 2 , Adrian Sutton 1
1 Department of Physics Imperial College London London United Kingdom, 2 Culham Centre for Fusion Energy Abingdon United KingdomShow Abstract
The problem of irradiation damage in metals is an important issue in the fields of fusion and fission power generation. High energy neutrons generate large amounts of free vacancies and interstitials in structural materials. The migration of these point defects gives rise to the nucleation and growth of nanometric clusters, such as prismatic dislocation loops and cavities, which modify the mechanical, thermal and electrical properties of the material. Such damage can be recovered in an annealing process, via the climb motion of dislocation loops and evaporation of cavities. These processes are diffusion-driven, meaning that the relevant time scales are much longer than the ones typically achievable with atomistic simulations. In addition, dislocation dynamics (DD) simulations usually employ descriptions of dislocation climb based on mobility laws that disregard non-local effects due to the vacancy distribution, which are likely to be important when investigating configurations with multiple sinks and sources. Here we present a unified continuum model for the annealing of dislocation loops and cavities, which makes use of boundary integral equations to solve the diffusion problem, expanding on recent work by Gu et al. Preliminary results show that non-local effects play an important role when dislocations are closely packed together, or, in general, when a dislocation segment is close to other sources or sinks of vacancies.
10:30 AM - ES5.5.05
Interplay between Magnetism and Energetics in FeCr Alloys from
a Predictive Non-Collinear Magnetic Tight-Binding Model
Romain Soulairol 2 , Cyrille Barreteau 1 , Chu Chun Fu 2
2 DEN-Service de Recherches de Metallurgie Physique, CEA, University Paris-Saclay Gif sur Yvette France, 1 SPEC, CEA, CNRS, University Paris-Saclay Gif sur Yvette FranceShow Abstract
Fe-Cr steels are promising candidates for structural materials in advanced fission and future fusion reactors. Magnetism is a key driving force controlling several thermodynamic and kinetic properties of Fe-Cr alloys. Considering the necessity of performing accurate atomistic simulations with large supercells, not accessible with first-principles methods, we have developed a new tight-binding (TB) model for Fe-Cr, where magnetism is treated beyond the usual collinear approximation.
A major advantage of this model consists in a rather simple parameterizing procedure. In particular, no specific property of the binary system is explicitly included in the fitting database. The present model is proved to be accurate and highly transferable for electronic, magnetic and energetic properties in a large variety of structural and chemical environments: surfaces, interfaces, embedded clusters, and the whole compositional range of the binary alloy. The occurrence of non-collinear magnetic configurations caused by magnetic frustrations is successfully predicted.
The present TB approach can also apply to other binary magnetic transition-metal alloys. It is expected to be particularly promissing if the size difference between the alloying elements is rather small and the electronic properties prevail.
10:45 AM - ES5.5.06
Properties of Defects in Fe-Cr-Ni Alloys from First Principles
Jan Wrobel 1 2 , Duc Nguyen-Manh 2 , Mikhail Lavrentiev 2 , Sergei Dudarev 2 , Krzysztof Kurzydlowski 1
1 Faculty of Materials Science and Engineering Warsaw University of Technology Warsaw Poland, 2 Culham Centre for Fusion Energy Abingdon United KingdomShow Abstract
The significance of Fe-Cr-Ni system stems from the fact that it forms the basis for austenitic, ferritic and martensitic steels exhibiting diverse magnetic, thermodynamic, mechanical and irradiation properties. Ferritic-martensitic steels like EUROFER and ferritic ODS steels are the most promising candidates for the first wall fusion reactor applications. Despite the fact that Fe-Cr-Ni alloys are one of the most extensively studied ternary alloy systems, the relation between their magnetic phase stability and properties of defects formed under irradiation is not known.
We investigate magnetic properties as well as phase stability of ternary Fe-Cr-Ni alloys at finite temperatures using a combination of Density Functional Theory (DFT) with Cluster Expansion methods. We derive cluster interaction parameters characterizing bcc and fcc crystal structures [1,2]. Enthalpies of formation of ternary alloys predicted by DFT-based Monte Carlo (MC) simulations combined with a correction derived from Magnetic Cluster Expansion at 1600K are in excellent agreement with data derived from experiments performed at 1565K. The relative stability of fcc and bcc phases is assessed by comparing the free energies of alloy formation.
Point defect properties of Fe-Cr-Ni alloys are investigated using representative bulk structures generated using DFT-based MC simulations. The formation energies of vacancies and dumbbell configurations as well as migration energies of vacancies defects in bulk structures are studied as functions of alloy composition the local environment of a defect.
 J. S. Wróbel, D. Nguyen-Manh, M. Yu. Lavrentiev, M. Muzyk, S. L. Dudarev, Phys. Rev. B 91, 024108 (2015).
 M. Yu. Lavrentiev, J. S. Wróbel, D. Nguyen-Manh, S. L. Dudarev, Phys. Chem. Chem. Phys. 16, 16049 (2014).
ES5.6: Ramping Up Materials Research for the Nuclear Renaissance
Tuesday AM, November 29, 2016
Sheraton, 2nd Floor, Back Bay C
11:30 AM - *ES5.6.01
Materials Research for Sustained Nuclear Growth
Rajagopala Chidambaram 1 , K.V. Manikrishna 1 , G.K. Dey 1
1 Bhabha Atomic Research Centre Mumbai IndiaShow Abstract
Energy Security and Climate Change are universally recognised as the major challenges of the 21st century. Nuclear energy is accepted as a mitigation technology in the context of the climate change threat and, for it to be a sustainable mitigation technology, the nuclear fuel cycle has to be closed. Closed fuel cycle, however, brings its own set of challenges of fuel reprocessing, apart from requirement of handling various types of reactors (thermal and breeder reactors).
India has a long-term strategy of following a closed nuclear fuel cycle. Its nuclear energy programme is based on three stages viz., U based 1st Stage, Pu based 2nd stage followed by Th based 3rd Stage. In addition, India is actively pursuing high temperature nuclear reactors and fusion technologies (in collaboration with the international community). The talk shall cover various materials-related issues and challenges in these programmes and research and development activities carried out in India to address them. These include materials development activities in thermal reactors (e.g. Zr-based alloys), fast breeder reactors (e.g. oxide dispersion steels), high temperature reactors (e.g. Nb alloys), nuclear fuel reprocessing and waste management. The talk shall also deliberate on the material degradation issues specific to various reactor technologies, and how these are addressed by modifying the material composition and microstructures. It will be noted that the knowledge in the nuclear system can have significant spill-over into the other energy fields, such as renewable energy and the ultra super-critical thermal power plants.
12:00 PM - ES5.6.02
Nanostructure Evolution of High-Chromium Ferritic/Martensitic Alloys under Neutron and Ion Irradiation—An Object Kinetic Monte Carlo Model
Monica Chiapetto 1 2 , Lorenzo Malerba 1 , Charlotte Becquart 2 , Nicolas Castin 1
1 SCK-CEN Mol Belgium, 2 Lille University of Science and Technology Lille FranceShow Abstract
High-chromium ferritic/martensitic steels are considered for fuel cladding and other core components in commercial GenIV reactors, as well as for the breeding blanket of fusion systems. However, despite exhibiting an optimal resistance to radiation-induced swelling, the structural integrity, service lifetime and mechanical properties of these materials are threatened, among others, by low-temperature hardening and subsequent embrittlement caused by neutron irradiation damage. To mitigate this problem it is important to understand which radiation-induced nanofeatures are responsible for it, by exposing these materials to irradiation and performing suitable characterization.
The limited availability of suitable neutron irradiation facilities, as well as costs and time expenditure of neutron irradiation experiments, have given rise to a continuously increasing interest in alternative irradiation sources, e.g. ion irradiation. Yet by using a surrogate for neutron irradiation a transferability issue is introduced, mainly due to the displacement rate, which is several orders of magnitude larger in case of ion irradiation, but also due to differences in the damage produced: for example, gradients caused by the self-ion limited penetration which can induce different defect concentrations in the irradiated materials, as well as the injection of extra interstitials.
As part of a multiscale modeling approach, in this work we use an object kinetic Monte Carlo model initially designed for neutron irradiation to study the response of ferritic/martensitic Fe-Cr-C model alloys to a range of conditions. Irradiation temperatures from values relevant for the water-cooled design of the fusion DEMO (i.e. ~250°C) to temperatures of the order of ~400°C, closer to those envisaged for GenIV applications, are explored. The effect of dose-rate and the differences to be expected depending on whether ion- or neutron-irradiation is applied are also investigated. For the simulation of the microstructure evolution under self-ion irradiation, in comparison with neutrons, the irradiation damage profile and the distribution of implanted ions as a function of depth are taken into account, as well as the effect of the ion beam scanning the sample, by comparing continuous versus pulsed irradiation.
Specific reference experiments performed under both ion and neutron irradiation were reproduced and the buildup of the defect populations, in terms of number density and mean size of defect clusters in the model, provided keys to interpret the experimental results.
12:15 PM - *ES5.6.03
Nuclear Materials Research in the UK
Jonathan Hyde 1 , C. A. English 1
1 National Nuclear Laboratory Oxfordshire United KingdomShow Abstract
After a long period without investment, the UK Government is now directly
supporting nuclear materials research. The initiatives including setting up
a National Nuclear User Facility, a Nuclear Fuels Centre of Excellence and
the Sir Henry Royce Institute for advanced materials. The partners involve
UK academic institutions and National Laboratories. Further, in parallel
with this, there has been a real resurgence of interest in the UK in
materials research for nuclear power applications which has stimulated
increased R&D activity in Industry and academia. This has also derived
impetus from significant advances in microstructural and modelling
capability, particularly in academia. The rationale behind these
investments, the recent advances in UK nuclear materials research, and the
vision for future nuclear materials research in the UK will be presented.
12:45 PM - ES5.6.04
Improving Nuclear Power Plant Safety with FeCrAl Alloy Fuel Cladding
Raul Rebak 1
1 GE Global Research Schenectady United StatesShow Abstract
To develop fuels with enhanced accident tolerance, the US Department of Energy (DOE) is partnering with fuel vendors to study alternatives to the current UO2 – zirconium alloy system. The proposed alternative should better tolerate loss of cooling in the core (similar to the Fukushima accident) for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric is proposing to replace zirconium based alloy cladding in current commercial power reactors with a FeCrAl cladding such as APMT. For the last three years extensive testing has been conducted to determine the suitability of the FeCrAl concept under normal operation conditions and under accident conditions. Results show that APMT performs better than zirconium alloys under normal operation conditions and it is several orders of magnitude more resistant to degradation by steam under accident conditions, generating less heat and lower amount of combustible hydrogen. The use of FeCrAl cladding is a direct path to improve the safety of commercial light water reactors.
ES5.7: Are Advanced Fuels Ready for Deployment?
Chu Chun Fu
Tuesday PM, November 29, 2016
Sheraton, 2nd Floor, Back Bay C
2:30 PM - *ES5.7.01
Atomic Scale Investigation of the Influence of Radiation Damage on the Properties of GEN IV Nuclear Fuels
Gerald Jomard 1 , Michel Freyss 1 , Yaguang Li 1 , Emerson Vathonne 1 , Emeric Bourasseau 1 , Marjorie Bertolus 1
1 CEA, DEN, DEC Saint Paul Lez Durance FranceShow Abstract
Nuclear power is one of the largest sources of low-carbon electricity and is considered worldwide as one of the options for mitigating climate change with renewable energy sources. Nuclear fuels play a key role for the objectives of resources optimization, waste minimization, safety and efficiency of nuclear energy and one challenge for the development of future nuclear reactors is to improve significantly the effectiveness of the design and selection of innovative fuels.
Fuel behavior under irradiation is extremely complex due to the combined effect of radiation and temperature. In particular, actinide fission produces large quantities of fission gases which have a significant influence on the structural and mechanical properties of nuclear fuels, as well as chemically active fission products, which induce very complex compositions and interact with other core materials.
An enhanced prediction capability of the fuel behavior in normal operation and accidental conditions is dependent upon improving our detailed understanding of the physical phenomena involved and the induced property changes and upon building more physically based models. Atomic scale methods are efficient tools allowing one to elucidate the elementary mechanisms governing the behavior of materials and their application to UO2 in recent years has shown that they enable one, combined with separate effect experiments, to get further insight on the behavior of nuclear fuels under irradiation [1-6].
We will show the application of atomic scale simulations to mixed oxide fuels and especially U-Pu oxide fuel or MOX, which will be used in the first European Generation IV prototypes, in particular the French reactor ASTRID. Among the various properties of interest, we will focus on defect and fission gas behavior in MOX and thermodynamic properties. The results will be compared to experimental data from the literature, obtained by X-ray absorption spectroscopy and X-ray diffraction, and to the results obtained in pure UO2.
 B. Dorado et al., “First-principles and experimental study of oxygen diffusion in uranium dioxide”, Phys. Rev. B 83, 035126 (2011)
 P. Garcia et al., “Nucleation and growth of intragranular defect and insoluble atom clusters in nuclear oxide fuels”, Nucl. Instrum. Meth. Res. B 277,98 (2012)
 J. Wiktor et al., “Coupled experimental and DFT+U investigation of positron lifetimes in UO2”, Phys. Rev. B 90, 184101 (2014)
 P.M. Martin et al., “Behavior of fission gases in nuclear fuel: XAS characterization of Kr in UO2”, J. Nucl. Mater. 466, 379 (2015)
 R. Bès et al., “Experimental evidence of Xe incorporation in Schottky defects in UO2”, Appl. Phys. Lett., 106, 114102 (2015)
 M. Bertolus et al., “Linking scales: atomic and mesoscopic scale modelling of the transport properties of uranium dioxide under irradiation”, J. Nucl. Mater. 462, 475 (2015)
3:00 PM - ES5.7.02
Multiscale Material Model Development and Simulations for Accident Tolerant Uranium–Silicide Fuels
Jianguo Yu 1 , Yongfeng Zhang 1 , Jason Hales 1
1 Idaho National Laboratory Idaho Falls United StatesShow Abstract
Use 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. This is due to a higher thermal conductivity than UO2 that results in lower centerline temperatures. U-Si also has a higher fissile density, which may enable some new cladding concepts that would otherwise require increased enrichment limits to compensate for their neutronic penalty. However, many critical material properties for U-Si have not been determined experimentally. It is anticipated that modeling and simulation may deliver guidance on the importance of various properties and help prioritize experimental work. In this talk, we will present our recent progress on multiscale material model development for accident tolerant U-Si fuels, spanning density functional theory calculations, molecular dynamics potential development, phase field simulation and engineering scale modeling. Our ultimate goal is to develop knowledge-based models for use at the engineering scale with a minimum of empirical parameters.
3:15 PM - *ES5.7.03
Fuel Development, Testing, and Schedule for the Traveling Wave Reactor
Kevan Weaver 1
1 Terrapower Inc. Bellevue United StatesShow Abstract
As the nations of the world seek to reduce emissions of greenhouse gases and other airborne pollutants, the reliance on carbon intensive fuels will have to be decreased dramatically. The use of nuclear energy as a reliable and scalable option to combat climate change and pollution, while increasing a nation’s energy security, is possible. France provides proof that nuclear energy can displace fossil energy at large scale, where they successfully transitioned from coal to nuclear in two decades. TerraPower has been developing a sodium-cooled fast reactor called the Traveling Wave Reactor (TWR) that aims to greatly improve the perceived shortcomings of nuclear energy by enhancing safety, and providing secure, economic, carbon-free electricity. The TWR will be capable of utilizing either depleted or natural uranium as its main source of fuel. This can be achieved without reprocessing, while reducing the volume of used fuel at the end-of-life. In order to achieve this, the fuel will need to achieve high peak burnups (on the order 30% FIMA), and the materials will need to survive high neutron fluences (on the order of 1.1 x 1024 per cm2, or ~550 dpa). These burnups and fluences are much higher than have been achieved historically, and will require test and qualification programs to demonstrate fuel and material performance. TerraPower’s development and test programs are working towards this goal, where both heavy ion and neutron irradiations are being performed, including irradiation of metal fuel. This presentation will describe the program and the results achieved to date.
3:45 PM - ES5.7.04
Post-Irradiation Examination of a U0.17ZrH1.6 Fuel with Electron Microscopy
Edgar Buck 1 , Michele Conroy 1 , Andrew Casella 1
1 Pacific Northwest National Laboratory Richland United StatesShow Abstract
UxZrHy fuels provide a number of design and performance advantageous over conventional UO2. These advantages include self-moderation (allowing for a more compact core design), enhanced thermal conductivity, and a large negative temperature reactivity coefficient. Both unirradiated and irradiated specimens of a U-Zr-Hydride fuel bonded to Zircaloy cladding with a Pb-Bi liquid metal (LM) were examined with SEM-FIB. The FIB 3D ‘slice and view’ prevented oxidation of the U metal and provided a unique insight into the materials changes following irradiation. Examination revealed significant phase separations in the LM and corrosion/erosion of the cladding and to a lesser extent fuel. In the unirradiated specimen, Pb and Bi appeared to be well structured within the inner LM, Zr appeared to be confined to its expected locations within the cladding and the fuel matrix, and uranium was fairly evenly dispersed within the Zr matrix. The structures in the irradiated specimens were still largely unchanged, but that the inner LM was restructured, the Zr showed some evidence of corroding/eroding away, and there were indications of the presence of Zr in the inner LM. Uranium was also observed to be moving as a bulk phase away from the fuel toward the cladding, smaller amounts of uranium appear dispersed within the inner LM, the fuel surface was fractured and penetrated by inner LM, and the cladding was highly corroded/eroded.
ES5.8: New Techniques to Quantify Radiation Damage
Tuesday PM, November 29, 2016
Sheraton, 2nd Floor, Back Bay C
4:30 PM - *ES5.8.01
Irradiation Damage and Microstructure Evolution in Zr-Based Nuclear Fuel Cladding Studied by Ultrahigh Resolution EDX Spectrum Imaging and Synchrotron X-Ray Diffraction
Michael Preuss 1
1 University of Manchester Manchester United KingdomShow Abstract
In water-cooled reactors zirconium alloys have been the material of choice for fuel assemblies due to a combination of low neutron cross-section, excellent corrosion performance and good mechanical properties. However, fuel cladding performance, or our ability to predict its performance, remains the limiting factor in an effort to push for increased fuel burnup, i.e. the energy extracted from a fuel assembly before it is removed from the core.
Important in-rector properties of fuel cladding are the resistance to corrosion and hydrogen pick up as well as dimensionally stability of the material. These materials performance aspects are particularly affected by irradiation damage and the associated micro/nano structure evolution, which need to be characterised in great detail.
I will present results of detailed studies using a multiscale characterisation approach by employing diffraction and novel electron microscopy techniques. In order to develop a better understanding of the evolution of damage and micro-segregation, proton as well as neutron irradiated zirconium alloys have been investigated using STEM-based ultrahigh resolution EDX spectrum imaging and synchrotron x-ray diffraction. These investigations have been carried out on two types of zirconium alloys, Zircaloys (no Nb) and Zr-Nb type alloys. These studies have provided a detailed picture and have enabled quantitative analysis of the evolution of and dislocation loops as a function of dpa. In addition, ultrahigh resolution EDX mapping provides new insight in the possible role of micro-segregation and the formation of nano-precipitates or nano-clusters on dislocation loop formation. These new observations are interpreted in the view of dimensional instabilities observed for zirconium alloys, particularly growth, as well as the potential impact on corrosion and hydrogen pick-up
5:00 PM - ES5.8.02
Continuous Thermo-Mechanical Property Monitoring under Ion Irradiation Using Transient Grating Spectroscopy—Benchmark Studies on Pure Copper
Cody Dennett 1 , Sara Ferry 1 , Di Chen 2 , Jonathan Gigax 2 , Alejandro Vega-Flick 1 , Vikash Mishra 1 , Alexei Maznev 1 , Keith Nelson 1 , Lin Shao 2 , Michael Short 1
1 Massachusetts Institute of Technology Cambridge United States, 2 Texas Aamp;M University College Station United StatesShow Abstract
The use of ion beam irradiations as a meaningful emulation tool for high-dose neutron irradiations is rapidly being qualified in the nuclear materials community. However, current schemes of materials examination are largely unchanged from the standard processes developed for neutron exposures. Namely, samples are characterized, irradiated, and then undergo post-irradiation examination (PIE). Given the flexible, low-activation environment present in many ion beam facilities, the opportunity exists for the development of continuous, on-line characterization techniques to monitor material property changes in-situ, as irradiation is taking place. The use of transient grating spectroscopy (TGS), a photoacoustic technique sensitive to several thermo-mechanical properties of interest in irradiated materials, is being developed for this purpose. Initial ex-situ characterization of self-ion irradiated pure single crystal and polycrystalline copper samples showed distinct changes in measured TGS signals, related surface morphology changes, and a reduction in elastic modulus and thermal diffusivity. These initial studies motivate the development of an in-situ TGS ion beamline experiment, the conceptual design of which has been completed. Such a facility would allow for not only vastly increased dose resolution of material properties affected by irradiation, but also for targeted PIE at transition points in irradiated material response, such as at the onset of breakaway void swelling. This technique therefore has applications as a tool to pinpoint the onset of radiation void swelling, allowing for the far more rapid qualification and study of families of alloys designed for extreme radiation resistance.
5:15 PM - ES5.8.03
Multi-Layer Nano-Indentation Method for Evaluating Dose Dependence of Radiation Hardening after Ion-Irradiation
Somei Ohnuki 1 2 , A. Sawa 2 , Hiroshi Oka 2 , Naoyuki Hashimoto 2
1 Materials Science and Engineering University of Science and Technology Beijing Beijing China, 2 Hokkaido University Sapporo JapanShow Abstract
Ion-irradiation is an important method for damage simulation with short term experiment, but a disadvantage is inhomogeneous damage profile as a function of depth. In this study the multi-layer nano-indentation method was applied for evaluating the dose dependence of radiation-hardening from a piece of irradiated specimen. Austenitic and F/M steels were irradiated by 10.5 MeV Fe3+ to 20 dpa (in average) at 300 C.
The range of plastically deformed region due to the nano-indentation was confirmed by Nix-Gao plots and transmission electron microscopy. The multi-layer nano-indentation method was based on the following assumptions; (1) the irradiated region can be divided into sub-layers having their own local hardness, (2) the local hardness can be depended on the fraction of each sub-layer, and (3) the deformation zone is a hemisphere shape.
Eventually, through the FIB sectioning followed by nano-indentation, local hardness in each sub-layer was experimentally evaluated, and then we got a dose dependence of radiation-hardening RH up to 60 dpa with reasonable relation of HV = A + B (dose) n. Furthermore the correlation between ion- and neutron irradiation data will be discussed in this study.
5:30 PM - ES5.8.04
Hearing Pulsed Laser Melting in Heat-Resistant Materials for Applications in Post-Irradiation Examination of Nuclear Fuels
Zhandos Utegulov 1 , Azat Abdullayev 1 , Baurzhan Muminov 1 , Almas Rakhymzhanov 1 , Nessipbek Mynbayev 1
1 Nazarbayev University Astana KazakhstanShow Abstract
There is a strong need to develop advanced instrumentation for measurement of thermal properties of nuclear materials in extreme environments for the analysis of the operation limits of nuclear fuels, prediction of possible nuclear reactor accidents and for development of next generation safe and efficient nuclear fuels. Thermal arrest of nuclear fuels suffers from fuel-furnace interaction along with volatilization and contamination during melt testing. To minimize these problems an ample progress has been made in developing experimental techniques enabling investigating the response of nuclear fuels to millisecond laser pulse heating. In this work we explore the response of refractory materials to nanosecond pulse laser surface heating, combined with in-situ laser vibrometry based on two-wave mixing photorefractive interferometry technique.
We demonstrate automated nanosecond pulsed laser heating of metals across melting, but below ablation regime. We have measured laser pulse-induced melting thresholds for different refractory metals by detecting laser-generated ultrasound. The experiments were conducted with motorized stepper motor to control the incident laser pulse energies and simultaneously measure both the varying laser pulse energy and epicentral acoustic waveforms. Finite element modeling based on coupled elastodynamic and thermal conduction governing equations were found to be in good agreement with experimental results. The onset of melting was attributed to the formation of the molten pool leading to preferential generation of shear waves from acoustic sources surrounding the molten mass resulting in the delay of shear wave arrival times. Observed trend in the increased melting thresholds for given refractory metals is in agreement with the increase of their corresponding melting point temperatures. It was also observed that the relative delay in shear wave arrival times above melting threshold increases in metals with higher melting points.
Demonstrated pulsed laser technique has proven to be a suitable and promising to study high temperature phase transitions. A nanosecond pulse laser melting technique using in-situ laser vibrometry can be installed in a hot cell where time- and spatially-resolved examination of solid-to-liquid phase transition in fresh and irradiated fuel samples can be examined. In-situ laser vibrometry based on two-wave mixing photorefractive interferometry determining phase transitions in refractory metals provides an independent, reproducible test to determine the onset of melting and can investigate the high temperature properties of nuclear materials. Latest measurement results on pulsed laser melting thresholds in various refractory materials will be discussed in the light of acoustic and thermal phenomena.
This research was funded by Nazarbayev University and target program № 0115РК03029 "NU-Berkeley strategic initiative in critical state of matter, advanced materials and energy sources" from the Ministry of Education and Science of the Republic of Kazakhstan.
5:45 PM - ES5.8.05
Characterization of Nuclear Materials with a Combination of Atom Probe and Transmission Electron Tomography
Peter Wells 1 , Stephan Kraemer 1 , Yuan Wu 1 , Soupitak Pal 1 , Takuya Yamamoto 1 , G. Robert Odette 1
1 University of California, Santa Barbara Santa Barbara United StatesShow Abstract
Advanced reactors will require development of new alloys that can withstand extreme environments, including high temperatures and neutron displacement damage, and must also manage large amounts of He generated in neutron-alpha reactions. These challenging service conditions lead to microstructure changes on the nano-scale, such as precipitation or He bubble/void formation, that are often at the resolution limit of characterization techniques. Atom probe tomography (APT) provides high resolution compositional measurements that track changes in solute locations, such as precipitation and segregation, in what are intrinsically 3D tomographic reconstructions. TEM provides other information, such as crystal structures, and probes larger sampling volumes, but often only in 2D projections. Here we report on insight gained using a correlative combination of APT and TEM tomography on the same sample volumes. Data is presented for tempered martensitic steels containing He bubbles and voids and nanostructured ferritic alloys containing nano Y-Ti-O oxide precipitates, and precipitate associated He bubbles. The correlative combination of these methods provides unique information and understanding of nanostructures and their associations. In addition, the correlated datasets highlight artifacts associated with the individual techniques.
Chu Chun Fu, CEA Saclay
Kazuto Arakawa, Shimane University
Sergei Dudarev, Culham Centre for Fusion Energy
Michael Short, MIT
ACS Energy Letters | ACS Publications
ES5.9: Combined Experiments and Simulation of Radiation Damage
Wednesday AM, November 30, 2016
Sheraton, 2nd Floor, Back Bay C
9:30 AM - ES5.9.01
An Atomistic Description of Severe Irradiation Damage and Associated Property Changes in Nuclear Graphite
Baptiste Farbos 2 3 , Helen Freeman 4 , Trevor Hardcastle 4 , Jean-Pierre Da Costa 3 , Rik Brydson 4 , Patrick Weisbecker 2 , Andrew Scott 4 , Christian Germain 3 , Gerard Vignoles 2 , Jean-Marc Leyssale 1 2
2 Laboratoire des Composites ThermoStructuraux University of Bordeaux/CNRS/CEA/Herakles Pessac France, 3 Laboratoire d'Intgration du Matériau au Système University of Bordeaux/CNRS/IPB/BSA Talence France, 4 Institute for Materials Research University of Leeds Leeds United Kingdom, 1 MultiScale Materials Science for Energy and Environment Massachusetts Institute of Technology Cambridge United StatesShow Abstract
While many graphite-moderated nuclear reactors in the world have seen their production times extended well above initial predictions, there is a crucial need to better understand the effect of prolonged irradiation on the structure of graphite parts as damages might both affect the mechanical integrity of the cores or perturb their thermal management.
In this work we have used electrons as a surrogate for neutron irradiation and have combined in-situ high-resolution transmission electron microscopy (HRTEM) analysis of electron irradiation in graphite with an image-based atomistic reconstruction procedure , which allows us to propose a dynamic time series of atomistic models, up to a total dose equivalent to approximately four months in a reactor core. We show that even at such high doses, the material retains highly anisotropic grains maintaining anisotropic properties such as stiffness and thermal conductivity. The proposed models, after property homogenization, account for the experimentally observed increase in Young's modulus and decrease in thermal conductivity with increasing dose. Further validation of the models is provided via a comparison of simulated and experimental data from irradiated material such as: HRTEM images, carbon K-edge electron energy loss spectra, dose rate and stored energies .
 Leyssale et al., App. Phys. Lett. 95, 231912 (2009); Farbos et al. Carbon 90, 472 (2014).
 Farbos et al. Under review
9:30 AM - ES5.9.02
Electron–Phonon Coupling in Ni-Based High Entropy Alloys with Application to Displacement Cascade Modeling
German Samolyuk 1 , Laurent Beland 1 , G. Malcolm Stocks 1 , Roger Stoller 1
1 Oak Ridge National Laboratory Oak Ridge United StatesShow Abstract
High-entropy alloys (HEAs) have recently been developed as nontraditional alloy systems. They are composed of multiple elements at or near equiatomic ratios that form random solid solutions on simple underlying lattices. In recent years HEAs have attracted significant attention due to their high strength, ductility and possible high radiation resistance. As a prototype system of HEAs we choose Ni-based binary and ternary alloys.
Energy transfer between lattice atoms and electrons is an important channel of energy dissipation during displacement cascade evolution in irradiated materials. On the assumption of small atomic displacements, the intensity of this transfer is controlled by the strength of electron– phonon (el–ph) coupling. The el–ph coupling in concentrated Ni-based alloys was calculated using electronic structure results obtained within the coherent potential approximation. It was found that Ni0.5Fe0.5, Ni0.5Co0.5, Ni0.5Pd0.5, Ni0.33Co0.33Cr0.33 and Ni0.4Fe0.4Cr0.2 are ordered ferromagnetically, whereas Ni0.5Cr0.5 is nonmagnetic. The el–ph coupling values for all alloys are approximately 50% smaller in the magnetic state than for the same alloy in a nonmagnetic state. As the temperature increases, the calculated coupling decreases. The rate of decrease is controlled by the shape of the density of states above the Fermi level. Introducing a two-temperature model based on these parameters in 10 keV molecular dynamics cascade simulation increases defect production by 10–20% in the alloys under consideration.
This work was supported as part of the Energy Dissipation to Defect Evolution (EDDE), an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Basic Energy Sciences. LKB acknowledges additional support from a fellowship awarded by the Fonds Québécois de recherche Nature et Technologies.
9:45 AM - *ES5.9.03
Developing a Coherent View of Helium Implantation Damage in Tungsten—Combining Multi-Technique Experiments and Atomistic Modeling
Felix Hofmann 1 , Duc Nguyen-Manh 2 , Daniel Mason 2 , Mark Gilbert 2 , Sergei Dudarev 2 , Isaure deBroglie 3 , Jeffrey Eliason 4 , Ryan Duncan 5 , Alexei Maznev 5 , Keith Nelson 5 , Christian Beck 6 , Wenjun Liu 7
1 Department of Engineering Science University of Oxford Oxford United Kingdom, 2 Culham Centre for Fusion Energy Abingdon United Kingdom, 3 Ecole Polytechnique Palaiseau France, 4 Department of Chemical Engineering and Materials Science University of Minnesota Minneapolis United States, 5 Department of Chemistry Massachusetts Institute of Technology Cambridge United States, 6 Department of Materials University of Oxford Oxford United Kingdom, 7 Advanced Photon Source Argonne National Laboratory Lemont United StatesShow Abstract
Tungsten-base materials are currently the most promising candidates for divertor armor components in future fusion reactors. During service they will be exposed to high temperatures, intense neutron flux and ion bombardment. Here we use helium ion implantation to study the interaction of injected helium with displacement damage and the effect the resulting defects have on the mechanical and physical properties of tungsten.
Using X-ray micro-diffraction and laser-induced transient grating measurements, we observe both lattice swelling and changes in the elastic modulus after ion implantation. Surprisingly, a fraction of a percent lattice expansion, driven by the accumulation of helium and implantation defects, causes an order of magnitude larger reduction of the elastic modulus. These observations are interpreted using a combined elasticity and density functional theory based model. We find that our experimental observations are consistent with a damage microstructure dominated by Frenkel defects with helium-filled vacancies. Measurements on helium-implanted single crystals further confirm the slight increase in elastic anisotropy predicted by our calculations.
Laser-induced transient grating measurements can also determine the thermal transport properties of the implanted layer. We find that even a modest concentration of injected helium leads to substantial reductions in thermal diffusivity. Using a kinetic theory model this effect can be explained in terms of the implantation-induced damage microstructure. Importantly the changes in thermal transport properties are not a trivial function of the implanted ion dose.
X-ray micro-diffraction measurements of lattice strains in samples heat-treated after ion-implantation show significant evolution of the damage microstructure. Indeed they suggest that at elevated temperatures defects migrate deeper into the material bulk. These multifaceted observations combined with multi-scale modeling allow us to begin to establish a joined-up picture of the complex effects helium-implantation-induced damage has on the structure and properties of tungsten.
10:15 AM - ES5.9.04
Direct Energetic Measurement and Quantification of Radiation Damage in Metals
Penghui Cao 1 , Sara Ferry 1 , Cody Dennett 1 , Ki-Jana Carter 1 , Sean Lowder 1 , R. Scott Kemp 1 , Michael Short 1 , Brian Turner 2 , Kevin Menard 2
1 Massachusetts Institute of Technology Cambridge United States, 2 Mettler-Toledo Inc. Columbus United StatesShow Abstract
The displacements per atom (DPA) is the current concept to quantify radiation damage in materials. The DPA of a material is defined as a calculation, not a measurement, integrating the energy-dependent fluence of energetic particles multiplied by the damage displacement cross section, and is a measurement of the number of times each atom is displaced from its original lattice site. It therefore is a measure of ballistic radiation damage, not the actual time-evolved residual damage that determines irradiation-affected material properties. In addition, no method exists to measure the DPA of a material irradiated in an unknown condition, as it is a fundamentally unmeasurable quantity. Methods such as transmission electron microscopy (TEM), atom probe tomography (APT), and mechanical testing can reveal some of radiation’s effects on materials, but cannot directly yield the defect populations which fully define irradiated material properties.
In order to gain a more fundamental understanding of radiation damage, a measurable unit of radiation damage must be developed which accounts for other irradiation parameters, such as dose rate and irradiation temperature. We propose the Wigner energy spectral fingerprint, a calorimetric measure of defect energy release during annealing at multiple temperatures, to quantify material damage. This directly measures the potential energy stored in defects in materials, by radiation and/or by other means. A combination of molecular dynamics (MD) simulations and calorimetric experiments shows that the Wigner energy approach quantifies material damage, and correctly identifies defect types by their stored energy per atom.
We present the results of neutron irradiation experiments and MD simulations of radiation damage in six different materials (Al, Cu, W, Zr, Fe, 304SS), spanning a wide range of crystal structures and melting points. MD simulations show that the total stored Wigner energies follow a scaling law normalized by the cohesive energy of the material, while defect types differ with varying PKA energy and material. Total stored energy is found to increase with PKA energy, following a power law with an exponent of about 0.85. Finally, we show that histograms of stored energy per atom are useful to automatically determine defect populations in irradiated materials. We also show the results of the experimental irradiation of these six materials by fast neutrons in the MIT reactor, along with differential scanning calorimetry (DSC) measurements of stored energy spectra. Heating rates ranging from 2°C/min to 2,000,000°C/min reveal different aspects about the Wigner energy spectral fingerprint of each material.
Applications of using the Wigner energy spectral fingerprint include determination of unknown dose levels of irradiated materials, providing fundamental understanding about the nature of radiation damage, and verifying historical production of enriched uranium for non-proliferation and treaty verification.
10:30 AM - ES5.9.05
Irradiation-Induced Grain Growth in a Thermally Stabilized Nanocrystalline Metal
Prince Singh 1 , Yoosuf Picard 1 , Maarten De Boer 1
1 Carnegie Mellon University Pittsburgh United StatesShow Abstract
Grain growth in irradiated elemental nanocrystalline (nc) metals follows a power law in which the grain size depends on the dose. The prevailing models assume that the grain growth is due to a thermal spike of ~10 ps duration that follows the collision and relaxation phases. In these models, the driving force for growth is due to the chemical potential difference resulting from grain boundary energy divided by grain boundary curvature, per the Gibbs-Thompson expression. The main features of the models are that the power law exponent, n, is quantitatively predicted (n~3) and that grain growth continues as dose increases. In thermally stabilized nc NiW, it is believed that the grain boundary energy is reduced, thus enhancing its resistance to grain growth at elevated temperature. According to the thermal spike model, irradiated NiW grains should then grow more slowly than pure Ni grains. However, our experimental results are in striking contrast with this expectation. Using Ni self-ion irradation, up to 10 displacements per atom (dpa), NiW nc grains grow much faster than Ni nc grains. The grain size is linear with irradiation and grows from 6 to 50 nm (n~1). However, beyond 10 dpa and up to 100 dpa, no further grain growth is observed (n>10 !). To investigate the source of the abrupt halt in grain growth, we are conducting TEM-based precession diffraction, by which we index the phase and orientation of each grain to gain quantitative information on the nc-metal texture evolution and emergence of the dominant grain boundary types. Initial results indicate a high fraction of low angle grain boundaries, which are likely to resist grain growth. This work will lead to a better understanding of irradiation-induced grain growth mechanisms in thermally stabilized nc metals, as well as conditions required to suppress grain growth.
ES5.10: Can Ions Reproduce Neutron Irradiation?
Wednesday AM, November 30, 2016
Sheraton, 2nd Floor, Back Bay C
11:15 AM - *ES5.10.01
High-Dose Self-Ion Irradiation as a Surrogate to Study the Relative Resistance to Neutron-Induced Swelling of a Wide Range of Ferritic/Martensitic Alloys and Their ODS Variants
Jonathan Gigax 1 , Hyosim Kim 1 , Eda Aydogan 1 , Frank Garner 1 , Lin Shao 1 , Mychailo Toloczko 2 , V. Voyevodin 3
1 Texas Aamp;M University College Station United States, 2 Pacific Northwest National Laboratory Richland United States, 3 Kharkov Institute of Physics and Technology Kharkov UkraineShow Abstract
The development of candidate ferritic-martensitic alloys and ODS variants of these alloys for advanced reactor service is hampered by the lack of neutron irradiation facilities capable of reaching the high doses required to assess their microstructural and dimensional stability under prolonged irradiation. One approach to overcome this problem is to use charged particle irradiation at accelerated displacement rates, with self-ions best suited for this purpose. This paper presents the results of a series of comparative irradiations focusing on void swelling, phase stability and the stability and action of various types of dispersoids. Two major accelerator facilities are involved, with some alloys irradiated in both facilities. One facility is in Kharkov, Ukraine and uses 1.8 MeV Cr ions. The second facility is in College Station, Texas and uses 3.5 MeV Fe ions.
The investigated ferritic-martensitic alloys are several heats of HT9, several heats of EP-450, T91 in several TMT variants, F82H, GA3X, ChS-139, EI-852, EK-181 and several nano-structured variants of EK-181. The ODS alloys are EK181-ODS, EP-450-ODS, MA956, several heats of MA957, 14YWT, PM-2000, FeCrAl-ODS and two groups of unique ODS steels developed at Hokkaido University and Kyoto University. Additional alloys are being added as they become available. Depending on the observed swelling resistance of each alloy, the dpa levels investigated range from 200 to 1000 dpa. While 1000 dpa might seem to be an unrealistically high dose, there is a suppression of void nucleation in the injected interstitial zone that precludes the extraction of swelling data in the higher-dose region of the damage vs. depth curve, producing ~600 dpa as the upper dose limit from a 1000 dpa peak dose.
All investigated alloys eventually exhibit void growth, with ODS steels in general, but not always, resisting the onset of swelling longer than do non-ODS counterparts. Those alloys that exhibit accelerated void growth swell at ~0.2%/dpa, but with vastly different incubation periods. MA 956 is an example of an ODS alloy where dispersoids actually accelerate the onset of swelling.
The results of these studies point to several routes to develop improved swelling resistance. Notably, dispersoid stability appears to be rather variable. One significant result is that the major role of dispersoids for swelling suppression appears to be maintenance of the stability of nano-structured grains with significant denuded zone. Several examples are shown where maintenance of nanostructured grains resists the transition to accelerated void growth to at least 500 dpa.
11:45 AM - ES5.10.02
Reexamination of the Neutron-Preconditioning Technique to Improve Ion Simulation of Neutron-Induced Void Swelling
Frank Garner 1 , Lin Shao 1
1 Texas Aamp;M University College Station United StatesShow Abstract
The limited availability of high flux neutron irradiation sources severely slows development of improved alloys for advanced reactors that require structural materials displaying radiation resistance to very high dpa levels of 300-600 dpa. Therefore the materials community has turned to charged particle irradiation at accelerated dpa rates as a surrogate for neutron irradiation to speed up the development process. This technique has especially focused on void swelling, a non-saturable process that can produce very large volume changes and distortions at high damage levels.
Void swelling can be divided into two major regimes, void nucleation which dominates the transient regime, and post-transient high-rate swelling regime, often referred to as "steady-state" swelling. While the latter is not very sensitive to most environmental and production variables, the former is very sensitive to almost all variables. It is well-known, however, that there are some significant differences, usually referred to as "neutron-atypical" aspects of ion irradiation that complicate the extrapolation of ion-induced swelling to neutron-relevant applications. These aspects not only add additional complexity but sometimes strongly suppress void nucleation.
One of the strongest differences between ion and neutron irradiation is the orders of magnitude difference in damage rate. Not only is the void nucleation process strongly dependent on dpa rate even in the simplest of alloys, there is often a radiation-driven microchemical evolution of the alloy matrix that precedes and defines the duration of the void nucleation regime. The dependence of the microchemical evolution is also sensitive to dpa rate, but not necessarily in the same manner as that of void nucleation, with the consequence that ion-induced swelling is more difficult to extrapolate to neutron-relevant conditions.
One approach to mitigate this problem is to use neutron-irradiated or "preconditioned" material at some moderate damage level and then continue the irradiation with self-ions to larger damage levels. In this case the microstructure and microchemistry of the alloy were already moving in a direction dictated by the neutron environment. Additionally, if void nucleation is largely in progress or complete, then suppression of nucleation by neutron-atypical aspects of subsequent ion irradiation are much less relevant.
In the 1970s and 1980s a series of such preconditioning experiments were initiated to explore the utility of the technique and to define how to select combinations of temperature and dose rate that would yield the best simulation. Many of these experiments were not published or completely analyzed, however, and the communities' perceptions of how to calculate ion-dose profiles were later strongly revised. This paper reviews and reanalyzes these experiments to provide guidance on how best to use this neutron-plus-ion technique for better simulation of void swelling at very high dpa levels.
12:00 PM - *ES5.10.03
Exploring Microstructure Evolution at High Damage Levels with Self-Ion Irradiation
Gary Was 1 , Elizabeth Getto 1 , Anthony Monterrosa 1 , Stephen Taller 1 , David Woodley 1 , Zhijie Jiao 1 , Kai Sun 1
1 University of Michigan Ann Arbor United StatesShow Abstract
Reactor core materials in both fast reactors and LWRs granted life extension must withstand irradiation to high doses at high temperature. Ferritic-martensitic (F-M) alloys are attractive candidates for structural components of fast and thermal reactors, and high Cr and high Ni stainless steels are potential replacement alloys for LWR core materials. To reach high damage levels, self-ion irradiation is conducted with simultaneous He injection into both F-M and austenitic alloys, accompanied by reactor irradiations conducted in the FFTF and BOR-60 fast reactor to assess the capability of ion irradiation to emulate the evolution of microstructure and mechanical properties in reactor. The evolution of microstructure at high damage levels and the interrelationships of the various features will be presented, along with comparisons to reactor irradiated microstructures.
12:30 PM - *ES5.10.04
Standardization of Accelerator Irradiation Testing Procedures for Simulation of Neutron Induced Damage in Nuclear Materials
Lin Shao 1 , Jonathan Gigax 1 , Frank Garner 1 , Mychailo Toloczko 2 , Jing Wang 1
1 Texas Aamp;M University College Station United States, 2 Pacific Northwest National Laboratory Richland United StatesShow Abstract
It is well known that the BOR-60 fast reactor can accumulate only 20 displacements per atom (dpa) or less per year in iron-base structural alloys. No other fast reactors are currently available. Currently available mixed-spectrum reactors (HFIR, ATR) in the USA cannot accumulate more than 10 dpa per year and usually much less in available test volumes. Total damage levels of 200-600 dpa are required to test the swelling resistance of advanced ferritic-martensitic alloy and their ODS variants. Therefore the international radiation materials community has turned its attention to charged particle simulation techniques at vastly accelerated dpa rates as surrogates for evaluation of new alloy concepts and candidates, especially with respect to void swelling and irradiation creep. However, there are significant differences between the spatial and temporal environments of neutron and charged particle irradiations. The various neutron-atypical aspects of ion bombardment preclude, however, an exact one-to-one ion-neutron comparison for confident near-exact prediction of swelling in a neutron environment. Claims were frequently made that predictive capability was possible, but such claims have not been validated and should be taken with a grain of salt.
In this talk, the most important "neutron-atypical" characteristics or "artifacts" of charged particle irradiation will be discussed. Major atypical factors such as defect imbalance, defect gradients, rastering and displacement rate effects are fairly well known and are being appropriately considered in data analysis. For each neutron-atypical factor, experimental evidence, multi-scale modeling, and solution are discussed in this talk. A call is made for urgent attention on standardization of accelerator testing procedures to maximize the credibility of any neutron-ion correlation.
However, as we optimize the ion irradiation to minimize the impact of some variables such as rastering, we can open the door to new challenges. In this presentation we will focus especially on a recently identified new issue, carbon injection from the accelerator beam lines into the irradiated area. Carbon contamination has not been well studied and is usually assumed to be ignorable. However, in non-rastered irradiations carbon contamination can increase significantly, and carbon is well known to participate strongly in void swelling due to strong defect interactions with carbon atoms or carbides. The issue is serious enough to possibly invalidate high dpa testing data of void swelling in stainless steels. We show that near-surface carbon enrichment results from beam-induced surface contamination, and we provide guidance on specific instrumental innovations to solve the contamination issue.
ES5.11: How Do Interfaces Resist Radiation Damage?
Wednesday PM, November 30, 2016
Sheraton, 2nd Floor, Back Bay C
2:30 PM - ES5.11.01
Predicting He Behavior at Cu-V Interfaces to Mitigate He-induced Damage in Plasma-Facing Materials
Dina Yuryev 1 , Di Chen 2 , Nan Li 2 , John K. Baldwin 2 , Yongqiang Wang 2 , Michael Demkowicz 3
1 Massachusetts Institute of Technology Cambridge United States, 2 Los Alamos National Laboratory Los Alamos United States, 3 Texas Aamp;M University College Station United StatesShow Abstract
Solid-state interfaces in metallic composites are preferential sites for helium (He) precipitation. We aim to predict the distribution of He network precipitates at interfaces to improve performance of plasma-facing materials in fusion devices. We use O-lattice theory to determine the initial distribution of He precipitate nucleation sites and phase-field methods to model the growth, interaction, and coalescence of He bubble networks. These methods can be used to design next generation multilayered metallic composites with enhanced resistance to He damage. We demonstrate our computational methodology on a Cu-V interface . We compare our simulations to experimental TEM images of He implanted Cu-V interfaces.
 D. V. Yuryev and M. J. Demkowicz, "Computational design of solid-state interfaces using O-lattice theory: An application to mitigating helium-induced damage," Applied Physics Letters, vol. 105, Dec 1 2014.
2:45 PM - *ES5.11.02
Interfaces, Defects, and Radiation Damage Evolution
Blas Uberuaga 1 , Enrique Martinez 1
1 Los Alamos National Laboratory Los Alamos United StatesShow Abstract
A key goal of materials research in the realm of nuclear energy is to design new materials that can withstand the high doses and extreme temperatures expected of next-generation nuclear reactors. Thus, there has been significant effort to understand the factors that control radiation damage evolution. Several such factors have been identified, including the use of interfaces, both grain boundaries and heterophase interfaces, as sinks to enhance the annihilation of defects produced under irradiation. However, despite a long-standing realization that interfaces can promote radiation tolerance, the precise atomic-scale mechanisms responsible have been murky at best.
Using a combination of accelerated molecular dynamics and kinetic Monte Carlo simulations, we examine defect interactions with grain boundaries in copper, a model system for more complex materials. We find that both the thermodynamics of defect-interface interactions as well as the kinetics of defects at the interfaces themselves are very sensitive to the atomic structure of the interfaces. Thus, the mobility of both interstitials and vacancies vary significantly from interface to interface. This, in turn, impacts the annihilation rate of defects within the interface. Effectively, each interface has a different defect annihilation rate which dramatically affects its sink efficiency and thus its ability to enhance radiation tolerance. We conclude that defect mobilities are interfaces are a key factor in determining whether a given interface is able to improve the overall response of the material to irradiation.
3:15 PM - ES5.11.03
First-Principle Calculations to Study the Role of Grain Boundaries in Self-Healing Properties of Point Defects
Jianbo Liu 1 , Jian Xu 1 , Shunning Li 1 , Baixin Liu 1
1 School of Materials Science and Engineering Tsinghua University Beijing ChinaShow Abstract
Understanding the self-healing mechanisms of defects in nanocrystalline materials is of importance for developing the structural materials that can support extended component lifetime under extremely hostile conditions. It has been revealed that grain boundary can play a critical role in the radiation damage process by acting as sinks of various point defects. In the present work, first principles calculations are carried out to investigate the energetic landscape of point defects (i.e. self-interstitials, He-interstitials, and vacancies) induced by the irradiation damage and their migration properties of the self-healing process in the vicinity of grain boundaries (GBs) in copper, focusing on six symmetric tilt grain boundaries that vary in their energies. Our results indicate that the low-energy GBs are generally accompanied by a higher propensity of self-healing behavior due to the inter-granular interstitials and intra-granular vacancies recombine with each other. The recombination process is proved to be regulated by two mechanisms: interstitial emission mechanism and vacancy mediated mechanism. For the low-energy GBs, the former mechanism demonstrates its efficiency in describing the atomic motion, while for the high-energy ones, the latter turns out to be superior. With the aid of these mechanisms, we conclude that the low-energy GBs are comparatively more radiation-resistant than the high-energy counterparts, which may shed light on the rational design of high-performance structural materials based on nanocrystalline alloys.
ES5.12: Advanced Materials for Fusion Systems
Wednesday PM, November 30, 2016
Sheraton, 2nd Floor, Back Bay C
4:30 PM - *ES5.12.01
Development of Tungsten Alloys with Improved Resistance against Irradiation and Recrystallization Embrittlement for Fusion Application
Makoto Fukuda 1 , Akira Hasegawa 1 , Shuhei Nogami 1
1 Tohoku University Sendai JapanShow Abstract
Tungsten (W) is prime candidate material for a plasma facing components in a fusion reactor. The degradation of the material properties of W due to high heat load and neutron irradiation is predicted during fusion reactor operation. To increase the reliability and lifetime of W, improvement of the mechanical property and the resistance against irradiation and recrystallization embrittlement are needed. W alloys such as potassium (K)-bubble dispersed and/or rhenium (Re) added W were designed based on the results of researches on neutron irradiation and heat treatment effects on microstructure and material properties of pure W and its alloys. Materials examined in this work were supplied by A.L.M.T. Corp., Japan, and was fabricated by powder metallurgy and hot rolling. The microstructure observation and the measurement of the mechanical and thermal properties were performed. The experimental results showed that the finer grain structure and higher tensile strength of W alloys than those of pure W, and the effectiveness of combination of second-phase dispersion and addition of alloying element was found. One of the drawback by addition of alloying element was decrease in thermal property, although this was not occurred by second-phase dispersion. To clarify the trade-off relation between improved and worsened properties, the structural analysis by using finite element method was performed to evaluate stress and strain responses of pure W and its alloys under fusion reactor relevant heat load conditions, and details of this analysis will be presented. To evaluate the long-term reliability of pure W and its alloys, the effects of thermal history and heavy-ion irradiation on microstructural development and mechanical properties such as tensile property and hardness were investigated. The details of the experimental results and the perspective of further high performance W alloys development will be presented.
5:00 PM - ES5.12.02
Interrelation of the Effects of Deuterium Plasma Exposure and Local Thermomechanical Properties of Tungsten
Yevhen Zayachuk 1 , Felix Hofmann 1 , Thomas Morgan 2 , Steve Roberts 1 3 , Sergei Dudarev 3
1 University of Oxford Oxford United Kingdom, 2 FOM Institute DIFFER Eindhoven Netherlands, 3 Culham Centre for Fusion Energy Abingdon United KingdomShow Abstract
Tungsten is one of the primary plasma-facing materials used in fusion technology. Exposure to a plasma of hydrogen isotopes leads to surface and sub-surface modification of tungsten-based materials, such as formation of macroscopic sub-surface cavities resulting in surface blistering. Plasma-induced modification is locally non-uniform – laterally and in depth, with a characteristic length-scale on the order of micrometers.
Several techniques are able to probe the relevant mechanical and thermal properties of tungsten locally at corresponding length scales. Continuous stiffness measurement nanoindentation allows depth-resolved probing of mechanical properties of individual grains in polycrystalline materials. Laser-induced transient grating measurements allow determination of local near-surface (within ~1 µm from the surface) thermal diffusivity. This contribution reports new results applying these techniques, combined with SEM surface imaging, EBSD crystallographic analysis and FIB cross-sectioning, to tungsten exposed to deuterium plasmas under the divertor-relevant conditions – ion flux density ~1024 m-2s-1, ion energy ~50 eV. We will address both the impact of intrinsic properties on the effects of exposure, and the effect of exposure on material properties.
We find that plasma exposure increases material hardness by up to ~1 GPa. The measured hardening decreases with depth, following the diffusional profile of deuterium concentration. Hardness is independent of crystallographic orientation, both in as-received and exposed material, but depends on the degree of internal misorientation within individual grains, being up to ~20% higher in deformed grains compared to recrystallized ones. Lateral dependence of blistering is found to depend on the polishing state of the surface. Mechanically ground samples exhibit a strong crystallographic dependence of blistering (with grains with <111> surface normal orientation the most susceptible to blistering and those with <001> blister-free), while surfaces of electrochemically polished samples form blisters regardless of grain orientation. On the other hand, blister size and areal density depend on internal misorientation, with larger densities of smaller blisters present in deformed grains. Plasma exposure also degrades thermal transport properties, reducing thermal diffusivity in near-surface region by up to ~15%.
5:15 PM - *ES5.12.03
New Composites and Alloys for Plasma-Facing Components of Fusion Reactors
Christian Linsmeier 1 , Jan Willem Coenen 1 , Johann Riesch 2 , Jan Engels 1 , Gietl Hanns 2 3 , Anne Houben 1 , Bruno Jasper 1 , Felix Klein 1 , Andrey Litnovsky 1 , Yiran Mao 1 , Soren Moller 1 , Rudolf Neu 2 3 , Marcin Rasinski 1 , Rahul Rayaprolu 1 , Alexis Terra 1 , Tobias Wegener 1
1 Forschungszentrum Jülich GmbH Jülich Germany, 2 Max-Planck-Institut für Plasmaphysik Garching b. München Germany, 3 Technische Universität München Garching b. München GermanyShow Abstract
The components and materials facing the plasma in a fusion reactor are subject to extreme and cyclic particle, power and neutron loads. The unique combination of loading conditions requires new “advanced” materials and components in order to realize fusion as an energy source for the future. High heat fluxes and hence surface temperatures establish strong thermal and stress gradients in actively cooled components. The application of different materials as armor towards the plasma, heat sink and structural materials or bonding layers create potential failure zones in these components, especially under cyclic loads. As a very important aspect, hydrogen isotope diffusion and permeation as well as erosion have to be considered in the material selection. The design of a fusion reactor requires a large degree of inherent safety in case of a potential loss-of-coolant event with air ingress.
To accommodate these combined material and system requirements, new material concepts are currently developed and neutron effects in fusion materials are studied by using ion and neutron beams. Damage induced by radiation deteriorates cycle stability and thermo-mechanical properties of the materials. Only in irradiation studies a credible lifetime and operational temperature window of plasma-facing materials can be assessed, as the deterioration is strongly material and temperature dependent.
This talk provides an insight in specific material requirements for first wall materials in future fusion power plants. As examples of current research, the development of tungsten fiber-reinforced tungsten composites, new self-passivating tungsten alloys and effects of nuclear damage (induced by fast particles) on material properties will be discussed. A new “advanced” material concept for a fusion reactor is envisaged, in particular the combination of oxidation-resistant tungsten alloys – which are generally even more brittle than pure tungsten – with a composite concept to establish pseudo-ductile behavior and controlled crack growth. Manufacturing routes of tungsten alloys and composites include powder metallurgy, as the classical route to refractory metals, as well as chemical vapor deposition and infiltration techniques. Further, experiments with >15 MeV proton beams to simulate irradiation damage by 14 MeV fusion neutrons are presented. The advantages and challenges of this technique in comparison to the application of low-energy fission neutrons will be outlined and discussed.
5:45 PM - ES5.12.04
Near Zero Tritium Permeation through Al
3 Diffusion Barrier Coatings
Daniele Iadicicco 1 2 , Francisco Garcia Ferre 2 , Matteo Vanazzi 1 2 , Marco Utili 3 , Fabio Di Fonzo 2
1 Politecnico di Milano Milano Italy, 2 Istituto Italiano di Tecnologia Milano Italy, 3 C.R. Brasimone ENEA Bologna ItalyShow Abstract
Concerning nuclear fusion systems, namely ITER (International Thermonuclear Experimental Reactor) and DEMO (DEMostrator Reactor), which take the tritium deuterium reaction as a reference, tritium breeding from Pb-16Li eutectic is one of the focus point of technological R&D activities worldwide. The permeation of tritium through the foreseen steels of eutectic Pb-16Li Breeder Blanket (BB) concepts is a challenge for the tritium balance in the reactor. Once tritium is produced, it must be appropriately extracted from the Pb-16Li eutectic, precluding losses by permeation through the steels towards the Helium or Water cooling systems. In order to avoid such losses, an adequate permeation barrier is required. We report on the barrier performance of advanced Al2O3 ceramic coatings, which are suitable for this task owing to their chemical inertia, high density and amorphous character.
In particular, hydrogen gas is used in order to simulate tritium. Tests are performed with a fixed barrier thickness (5 µm) and under a fixed hydrogen partial pressure (100 mBar). The temperature of the tests is varied within the range 523-923 K. Hydrogen permeation through coated disc specimens is detected using a quadrupole mass spectrometer. Our results show that the Al2O3 coatings tested are a promising tritium permeation barrier, with permeation reduction factors (PRF) approaching 105. Last, but not the least, we show that the coatings perform well as anti-corrosion barriers in harsh environments relevant for BB concepts, such as liquid lead.
ES5.13: Poster Session
Wednesday PM, November 30, 2016
Hynes, Level 1, Hall B
9:00 PM - ES5.13.01
Temperature- and Fluence-Dependent Surface Morphology Evolution of Ta under High-Flux, Low-Energy He+ Ion Irradiation
Theodore Novakowski 1 , Jitendra Tripathi 1 , Ahmed Hassanein 1
1 Purdue University West Lafayette United StatesShow Abstract
In future nuclear fusion devices, plasma-facing components (PFCs) will be subjected to high thermal loads and high-fluxes of helium (He) and hydrogen (H) isotopes. Currently, tungsten (W) is considered to be the leading candidate PFC due to its high melting temperature, high thermal conductivity, and low sputtering/erosion rates. However, recent studies have shown that W undergoes significant surface morphology evolutions (blisters, bubbles, and tendril-like nanostructures (fuzz)) when irradiated with high fluxes of low-energy He+ ions under simultaneous sample heating at elevated temperatures . These nanostructures, in turn, could lead to an additional source of plasma impurities and negatively impact reactor performance. The motivation of this work is then to explore He+ ion irradiation effects in other candidate PFC surfaces. Namely, our preliminary results for He+ ion irradiation of tantalum (Ta) have shown that Ta may exhibit a higher fluence tolerance for nanostructure formation . Mirror-polished Ta samples were irradiated with 100 eV He+ ions at a flux of 1.2 × 1021 ions m-2 s-1 to total fluences in the range 4.3 × 1024 - 3.5 × 1025 ions m-2 at constant temperatures in the range 773-1223 K. Resulting surface morphology changes were monitored and characterized with a combination of scanning electron microscopy (SEM), focused ion beam (FIB) SEM, X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), and specular optical reflectivity. Overall, it was found that Ta has similar structure formation temperature regimes to W and forms similar blister and tendril-like surface structures. However, Ta demonstrates a higher fluence tolerance to fuzz-like nanostructure formation. While Ta is often discredited as a PFC due to its relatively low thermal conductivity compared to W, this tolerance to high aspect ratio surface structure formation suggests that the thermal conductivity of Ta is less likely to degrade with time in an extreme reactor environment, therefore reclaiming its relevance as a PFC.
 S. Kajita, N. Yoshida, N. Ohno, Y. Tsuji, New. J. Phys. 17 (2015) 043038
 T.J. Novakowski, J.K. Tripathi, A. Hassanein, J. Nucl. Mater. 467 (2015) 244-250.
9:00 PM - ES5.13.02
Amorphous-Like Tungsten Coatings—Deposition, Processing and Thermomechanical Characterization under Thermal Loads
Edoardo Besozzi 1 , Andrea Pezzoli 1 , David Dellasega 1 2 , Matteo Passoni 1 2 , Marco Beghi 1
1 Politecnico di Milano Milano Italy, 2 Istituto di Fisica del Plasma quot;P. Caldirolaquot;, CNR Milano ItalyShow Abstract
Thanks to its high melting point and thermal conductivity, tungsten (W) is among the most promising candidates as plasma facing material (PFM) in fusion reactors (e.g. ITER, DEMO), where the resistance to extreme thermal loads during steady and transient operating conditions is a key requirement. Because of their peculiar properties, which can significantly differ from the bulk ones, W coatings are also widely exploited in Tokamaks as protective layers for plasma facing components (e.g. WEST). In addition, amorphous-like W coatings (a-W), with crystallites dimension < 1 nm and mass density about 11 g cm-3, have been used at lab-scale as proxy of redeposited W in the framework of plasma-wall interaction, mimicking bulk W modification during different tokamak scenarios .
In this work, we deposit a-W coatings by Pulsed Laser Deposition (PLD). Thanks to its high versatility in tuning various deposition parameters, PLD allows us to finely control the growth process of the coatings, tailoring their structure down to the nanoscale. The as-deposited a-W samples are then processed by thermal annealing treatments at temperatures up to 1200 K. The recrystallization behavior and its effects on the mechanical properties are thus investigated. The mechanical properties are measured by Brillouin spectroscopy, while X-ray diffraction analysis and Scanning Electron Microscopy monitor the structure and morphology evolution. We find that the recrystallization process for a-W starts at around 900 K, which is well below the bulk W recrystallization temperature (1700 K), where an increase by 60% of material stiffness with a corresponding loss of ductility by about 30% is observed .
As proposed in , the extreme thermal loads proper of nanoseconds laser pulses can be exploited to simulate the thermal effects of plasma transient events (i.e. ELMs) on W. In addition, the thermal fatigue effects and failure mechanisms (e.g. cracking, delamination) on a-W coatings can be investigated by fast thermal annealing induced by nanoseconds laser irradiation. In view of an experimental campaign of this type, we developed a numerical tool aimed at supporting the design of the experiments and at interpreting the measured data. The temperature and stress fields within the multilayer structure, and their evolution during irradiation, are thus derived by a 2D axial-symmetric numerical code based on the finite difference method. Particular attention is paid to the modeling of the interface between adjacent layers. Taking into account the ductile-to-brittle transition temperature of W, different conditions of cracks formation can be numerically deduced. The total coating/substrate interface stress is also derived so that the delamination process can be investigated.
 M.H.J. ‘t Hoen et al., Journal of Nuclear Materials 463, 989 (2015)
 E. Besozzi et al., Materials and Design 106 14–21 (2016)
 N. Farid et al., Nuclear Fusion 54 012002 (2014)
9:00 PM - ES5.13.03
The Defect Microstructure and Element Composition in Denuded Zones of Stainless Steels Irradiated in BN-350 Nuclear Reactor
Oleg Rofman 1 , Kira Tsay 1 , Oleg Maksimkin 1
1 Laboratory of Radiation Materials Science Institute of Nuclear Physics Almaty KazakhstanShow Abstract
Austenitic stainless steels are widely used as structural materials for nuclear energy facilities. Investigations of microstructure and properties of reactor steels under irradiation is an important task for modern materials science. One of the examples of radiation-induced microstructural changes is related to the development of denuded zones along grain boundaries. This research aims to combine data from previous investigations on denuded zones and to present results obtained from studies of stainless steels (1st and 2nd generations) from the fuel assemblies of the decommissioned BN-350 nuclear reactor (Aktau, Kazakhstan). The studied materials of shrouds from the fuel assemblies were 12Cr18Ni10Ti (AISI321 analog) and 08Cr16Ni11Mo3 (AISI316 analog). The stainless steels were irradiated with neutrons to damage dozes of 0.25-59 dpa in the temperature range of 280-423°C and after the decommissioning they were kept in water environment for 4-8 years. The investigated samples were cut from shroud edges at a different distance from the core along the fuel assembly’s height. Transmission electron microscopy (JEOL JEM2100) was used to perform comparison studies of defect structure in the matrix, as well as that of grain boundaries and near-grain boundary regions. The effect of damage dose on the width of denuded zones and pores distribution was determined. EDS analysis gave information on elements distribution (in particular, Cr and Ni) at different distances from a selected grain boundary with a denuded zone. The work discusses factors that affect the development of denuded zones, it also provide quantitative results of defects sizes and redistribution of elements. Some obstacles of performing quantitative EDS analysis for the irradiated steel samples with swelling were discussed. As a result, the findings may help to illustrate microstructural changes taking place in the austenitic stainless steels after irradiation to further improvement of the existing structural materials for nuclear energy applications.
9:00 PM - ES5.13.04
Post-Irradiation Examinations of Structural Alloys Exposed to Molten FLiBe Salt in MIT Reactor
Guiqiu Zheng 1
1 Massachusetts Institute of Technology Cambridge United StatesShow Abstract
Fluoride salt-cooled High-temperature nuclear Reactors (FHRs) is emerging as a leading reactor concept among all Gen IV nuclear reactors because it offers, among other benefits, a high degree of passive safety, high thermal efficiency, and low spent fuel. One primary challenge in the development of the FHR is the selection of structural alloy that is required to be durable at 700°C, and to be compatible with molten Li2BeF4 (FLiBe) salt, as well as to be stable in strong neutron irradiation environment. Among many candidate alloys, nickel-based Hastelloy N® and 316 stainless steel have been selected as the most promising structural alloys for FHR. In order to evaluate the performance of these two alloys in FLiBe salt in simulated environment of FHR, the static in-reactor corrosion tests in 7Li-enriched molten FLiBe salt at 700°C in both metal-lined graphite crucibles and bare graphite crucibles were successfully accomplished in MIT nuclear research reactor for 1000 hours. This study focuses on the post-irradiation examinations of these two in-reactor corrosion tested alloys using a suite of techniques, including scanning electron microscopy (SEM) in conjunction with energy dispersive spectroscopy (EDS), x-ray diffraction (XRD), transmission electron microscopy (TEM). From the microstructural results, the acceleration effects of neutron irradiation and graphite presence on the corrosion of these two alloys in molten FLiBe are understood which provides valuable reference to the materials selection for FHR.
9:00 PM - ES5.13.05
Modelling Long-Term Microstructural Evolution of Pressure Vessel Steel under Irradiation Combining DFT and AKMC
Christophe Domain 2 , Baptiste Pannier 1 2 , Charlotte Becquart 1
2 EDF Ramp;D Moret sur Loing France, 1 Lille University of Science and Technology Villeneuve D Ascq FranceShow Abstract
Radiation-induced embrittlement of reactor pressure vessel is an important issue for the life extension of the nuclear power plant. Understanding the long-term evolution of its microstructure is a major challenge. In these steels, rich solute clusters formed under irradiation contribute to the mechanical properties evolution. According to tomographic atom probe analyses, these clusters are mainly enriched in Cu, Ni, Mn, Si and P. However, the elementary mechanisms leading to their formation are still an open question.
The formation mechanisms of these clusters are tackled thanks to multiscale modeling based on electronic structure calculations (DFT) and Atomic Kinetic Monte Carlo (AKMC). Defect and small cluster energetics are obtained by DFT allowing to develop a cohesive model used in AKMC to simulate diffusion and obtain the microstructure evolution. AKMC is very time consuming for a system in conditions close enough to those in reactor. Strategies used to speed up the simulations will be exposed. Furthermore, improvements of the AKMC parameterization based on recent ab initio calculations will be presented. High dose simulated microstructures will also be compared to experimental results.
9:00 PM - ES5.13.06
Microstructural and Micromechanical Characterization of SiC-SiC Fiber Composites for Fuel Cladding Applications
Yevhen Zayachuk 1 , David Armstrong 1 , Steve Roberts 1 2 , Christian Deck 3 , Peter Hosemann 4
1 University of Oxford Oxford United Kingdom, 2 Culham Centre for Fusion Energy Abingdon United Kingdom, 3 General Atomics San Diego United States, 4 University of California Berkeley United StatesShow Abstract
Silicon carbide is a candidate material for the use in novel accident tolerant fuel cladding in nuclear technology due to its favorable properties, in particular reduced (compared to Zircaloy) oxidation under accident conditions, as well as good neutronic performance, high temperature strength and stability under irradiation. It is suggested to be used in the form of SiC-fiber reinforced SiC-matrix (SiC-SiC) composite.
Highly non-uniform and anisotropic nature of the composite materials means that in order to reliably model their behavior the knowledge of the individual properties of fiber and matrix, and, crucially, the fiber-matrix interfaces, is required. Micromechanical testing techniques, such as microcantilevers beam fracture, allow determination of such localized properties. This contribution for the first time reports the results of micromechanical measurements, coupled with microstructural characterization, on SiC-SiC composite material.
Material used in this study was provided by General Atomics. It consists of the commercially available Tyranno SiC fiber weaved reinforcement structure and matrix grown in-situ using the chemical vapour infiltration (CVI) technique. General structure of the composite, including fiber arrangement and porosity, was assessed using scanning electron microscopy (SEM) and X-ray tomography. Microstructure of fibers and matrix was characterized with electron backscatter diffraction (EBSD), transmission electron microscopy (TEM) and transmission Kikuchi diffraction (TKD) techniques. Fibres are found to be nanocrystalline with a grain size of ~50 – 300 nm. Microstructure of matrix is shown to be complex, with highly elongated grains (~200 – 500 nm wide, up to several microns long), arranged radially forming rings surrounding individual fibers.
Micromechanical studies included hardness measurements on fibers and matrix performed with nanoindentation, and interfacial fracture tests using focused ion beam (FIB) manufactured microcantilevers, both at room and elevated temperatures. The hardness of the fibres is shown to be significantly lower than that of the matrix (~25 and ~45 GPa respectively), this is thought to be due to porosity in the centre of the fibres. As a result of indentation in the fiber, extrusion of the carbon interlayer between fibre and matrix upwards form the surface was observed. Interrupted microscale interfacial fracture tests showed that failure in mode I fracture proceeds through the carbon interlayer, rather than down either SiC-C interface. Understanding how this interlayer fails is key to understanding the macroscopic properties of SiC-SiC composites, and presented results will be related to macro-scale behavior.
9:00 PM - ES5.13.07
Effect of Radiation on Embrittlement and Matrix Cu Content of a RPV Weld with Different PWHT Conditions
Mikhail Sokolov 1
1 Oak Ridge National Laboratory Oak Ridge United StatesShow Abstract
The influence of temperature and cooling rate on the embrittlement and the copper level in the matrix has been investigated on a weld fabricated from the same weld wire used for HSSI Weld 73W. This weld has a relatively high bulk copper content, 0.32% wt. The heat treatment consisted of heating the material to the desired temperature, holding at the post-weld heat treatment (PWHT) temperature, and then cooling down to room temperature. Except for special cases, all PWHTs were performed with a heating and cooling rate of 15oF/h (8oC/h) to simulate the heating/cooling rate of a real vessel. In two special cases, material was heated with 15oF/h (8oC/h) rate but water quenched after holding at the PWHT temperature. The highest PWHT temperature was 650oC/24h, while the other PWHTs were 610oC/24h, typical PWHT of reactor pressure vessels (RPV), and 580oC/100h. Charpy impact properties were measured using sub-size 3x4 mm specimens and matrix Cu content was measured by atom probe tomography before and after irradiation. Small-angle X-ray scattering (SAXS) was used to measure size and volume fraction of copper-rich precipitates in the irradiated specimens. Charpy specimens were irradiated in the Ford Reactor at 288oC to 0.8x1019 neutron/cm2 (E>1MeV). It was found that the higher PWHT temperature resulted in higher Charpy upper-shelf energy (USE) with little effect on the ductile-to-brittle transition temperature (DBTT). The lower PWHT temperature and slower cooling rate were found to be beneficial in reducing the matrix Cu content. The matrix Cu content after irradiation was approximately the same for all three welds measured regardless of their different matrix Cu contents in the unirradiated condition. Consequently, the weld with the lowest PWHT temperature exhibited the lowest shift of DBTT and drop in USE. SAXS results reveal smallest volume fraction of copper-rich precipitates in this weld compared to others.
9:00 PM - ES5.13.08
Molecular Dynamics Simulation of Fission Fragment Damage in Nuclear Fuel and Surrogate Material
Ram Devanathan 1
1 Pacific Northwest National Laboratory Richland United StatesShow Abstract
We have performed classical molecular dynamics simulations of swift heavy ion damage, typical of fission fragments, in nuclear fuel (UO2) and surrogate material (CeO2) with energy deposition per unit length in the range of 10-40 keV/nm. We did not observe amorphization. The damage consisted of isolated point defects at lower ernergies and defect clusters at higher energies. The simulations provide valuable insights on density changes in the ion track and the role of radiation damage in fission product diffusion. These results shed light on features observed by electron microscopy of swift heavy ion irradiated ceria.
9:00 PM - ES5.13.09
Microstructure and Mechanical Properties on of Silicon Carbide Layer in TRISO-Coated Fuel Paticle Deposited by Fluidized Bed Chemical Vapor Deposition at Various Temperatures
Yeon-Ku Kim 1 , Weon Ju Kim 1 , Sunghwan Yeo 1 , Moonsung Cho 1
1 Korea Atomic Enery Research Institute Daejeon Korea (the Republic of)Show Abstract
The influence of microstructure on the mechanical properties of SiC layer coated onto spherical simulated nuclear fuel particles by fluidized-bed chemical vapor deposition (FBCVD) was investigated. While nano-indentation system was utilized to study mechanical properties of SiC layers, their microstructure were characterized by scanning electron microscopy (SEM) with EBSD, raman spectroscopy, and magic angle spinning nuclear magnetic resonance (MAS-NMR). The lower temperature deposition of SiC layer is advantageous factor preventing the property degradation of inner pyrolytic carbon layer which occurs at ~1500oC. It was found that SiC layers deposited at lower temperature than 1400oC obtained small grains sized between 0.2 and 1.0 µm and the non-stoichiometric ratio of Si and C. However, those phenomena were not observed in SiC layers deposited at higher temperature than 1400oC. Young’s modulus and hardness measurements of SiC layers result in 302~358 and 32~40 GPa, respectively.
9:00 PM - ES5.13.10
The Effect of Temperature and Oxygen Concentration on the Stability of Li and Sn Films on Polycrystalline Tungsten
Oluseyi Fasoranti 1 , Bruce Koel 1
1 Princeton University Princeton United StatesShow Abstract
Liquid lithium (Li) and tin (Sn) are potential candidates for plasma facing materials in the divertor and first walls of fusion reactors due to attractive properties for self-recovery and heat-flux management. It is important to develop a strong understanding of how these metals are affected by the high-temperature environments that are typical in fusion devices. Surface science experiments under UHV conditions have proven to be a useful approach to study these materials and individual particle-surface interactions by enabling clean and controlled deposition of films and their subsequent behavior. We communicate recent investigations from UHV experiments on the thermal stability of ultrathin Li and Sn films on a polycrystalline W surface using temperature programmed desorption (TPD), Auger electron spectroscopy (AES) and low energy ion scattering (LEIS). This multitechnique approach provides a good picture of the various surface processes that occur such as diffusion, alloying, and desorption. Comparisons of the oxidation kinetics following exposure to residual gases on the solid and liquid films as obtained from oxygen uptake curves will be discussed. Additionally, the effect of both post-oxidation as well as preadsorbed surface oxygen and carbon on the thermal stability of these films will also be discussed. We also plan to report ongoing work to study deuterium uptake from incident beams of low energy D2+ and atomic D on Sn and Li films as a function of temperature and film thickness using TPD.
9:00 PM - ES5.13.11
The Effect of Neutron Irradiation on Corrosion Resistance of Materials Used in Shrouds of Fuel Assemblies from BN-350 Fast Nuclear Reactor
Alevtina Yarovchuk 1 , Oleg Maksimkin 1 , Ludmila Turubarova 1
1 Institute of Nuclear Physics Almaty KazakhstanShow Abstract
The study presents results of observations and experiments on the effect of neutron irradiation on corrosion resistance of the structural materials from BN-350 fast nuclear reactor. In the process of decommissioning and utilization of the spent fuel assemblies (about 8 years of exploitation and storage in water environment) an observation of shroud’s surface was performed. Materials of the shrouds were 12Cr18Ni10Ti and 08Cr16Ni11Mo3 austenitic stainless steels. The calculated damage doses for the studied samples were in the range of 55-59 dpa, the irradiation temperature did not exceed 430°C. Visual and metallographic observations indicated the presence of corrosion-induced defects on shrouds surface. It was determined that the rate of corrosion was the most intensive at internal surface of the shrouds. There were also observed local defects, such as pitting, cavities, and inter-crystalline cracks which penetration into the material has reached about 45% of the shroud thickness. The performed analysis of the element composition let us to conclude that neutron irradiation resulted in concentration changes of Cr, Ni and Ti in certain areas of the sample, and it was assumed to be a cause for corrosion development. Laboratory examinations of the corrosion rate for the samples irradiated to different damage dozes showed that weight loss in the iron chloride increases in dozens of times with a damage doze increase. It was determined that irradiation with neutrons significantly changes steels microstructure, results in radiation-induced diffusion of alloying elements, development of alpha-martensite which decreases corrosion resistance of austenitic steels.
9:00 PM - ES5.13.12
Martensitic Transformations in AISI321 Type Austenitic Steels Irradiated with High-Energy Charged Particles
Oleg Maksimkin 1 , Kira Tsay 1 , Michael Short 2
1 Institute of Nuclear Physics Almaty Kazakhstan, 2 Massachusetts Institute of Technology Cambridge United StatesShow Abstract
Martensitic α'-phase forming in austenitic stainless steels, the materials for fission and fusion reactors exposed to irradiation, stess and temperature gradients can sufficiently influence on both mechanical and corrosion properties that eventually reduces the operation period for reactor structural components. So it is actual to carry out the detailed investigation of peculiarities of α'-phase formation, growth and annealing in irradaited steels.
The present work reports the results of complex research of structure and properties changes for the reactor 18-10 chromium-nuckel steel caused by both the direct γ→α' and the reverse α'→γ martensite transformations after irradiation with high-energy charged particles (pulsed electron flux, ions of helium and krypton). Plain-shaped specimens for mechanical tests with gauge dimensions of 10×3.5×0.3 mm are comparatively investigated in both nonirradiated and irradiated states.
The “irradiation-induced martensite” was found in 12Cr18Ni10Ti steel directly after bombarding with a powerful pulse electron flux with energy of 150-300 keV and power density of 5×108 – 4×109 W/cm2. . The martensitic nuclei forming in sites of a cellular dislocation network were observed in 15-20 mm depth from irradiation surface. On the opposite surface the lath–shaped martensite was observed forming in the result of shock compression in lattice during the puls electron flux action.
Experiments indicated that irradiation of 12Cr18Ni10Ti steel with high-energy krypton ions 84Kr+14 with 1.56MeV (the DC-60 accelerator, Astana, RK) generated on the bombarded surface numerous carbides, nitrides, gas blisters as well as microinclusions of martensitic α (bcc) and ε (hcp)-phases. At fluences of (1 ÷ 4)×1015 ion/cm2 the martensite can be revealed by means of EBSD-method (JSM-7500F microscope), at higher fluences of (6 ÷ 9)×1015 ion/cm2 the standard methods of phase analysis as X-ray diffractometry and magnetic measurements can be used. It was found that an increase in fluence of ions resulted in an increase of α'-phase volume fraction forming in steel.
For studying the reverse martensitic transformations the specimens of 12Cr18Ni10Ti steel irradiated with helium ions (50 MeV) uniformly in volume (U-150M cyclotron, Almaty, RK) to 10-3 at.% were deformed and subjected with a series of isochronous (30 min) annealings over the temperature range of 300-800oC. As shown, the reverse a¢®g transformations did not change in monotonic way during the temperature increase, at temperatures of 400-500oC there was an increase in a'-phase content caused by relaxation of compression strains in the steel lattice.
9:00 PM - ES5.13.13
In Situ Observation and Atomic Resolution Imaging of Zr-4 Initial Oxidation
Yang Yang 1 , Akihiro Kushima 1 , Huolin Xin 2 , Peter Hosemann 3 , Ju Li 1
1 Massachusetts Institute of Technology Cambridge United States, 2 Center for Functional Nanomaterials Brookhaven National Laboratory Upton United States, 3 Department of Nuclear Engineering University of California, Berkeley Berkeley United StatesShow Abstract
Zirconium based alloy is widely used as fuel cladding materials and some other structural materials in water-cooled reactors. However, degradation of zirconium alloy by waterside corrosion in harsh conditions severely limits the burnup and thus cycle life of nuclear fuels. Extensive researches have been done to understand this problem; however, several key aspects of knowledge related to the corrosion mechanisms of zirconium alloy still remain unclear. So far, most microstrutural analysis of corrosion behaviors of zirconium alloy is performed ex-situ. While few in-situ TEM studies of Zr-4 oxidation have been performed, none of them reaches atomic resolution, hindering the understanding of detailed corrosion process.
We performed in-situ environmental transmission electron microscope (E-TEM) study of zircaloy-4 initial oxidation in gas environments at atomic resolution for the first time. Experiments under water vapor and oxygen gas environments are done for comparison. Our preliminary analysis shown growth of nano-sized zirconium oxide crystals at the sample edge. This research will help discover new corrosion mechanisms of zirconium alloy, and thus provide insights on prediction and prevention of zirconium alloy corrosion in nuclear reactors.
9:00 PM - ES5.13.14
First-Principles Calculation of Hydrogen Absorption to a Planar Vacancy Cluter in Zirconium
Mitsuhiro Itakura 1 , Taira Okita 2
1 Japan Atomic Energy Institute Kashiwa, Chiba Japan, 2 University of Tokyo Kashiwa JapanShow Abstract
"RBWR" （Resource- renewable BWR) is a new type of BWR which concentrate neutron flux to a specific part of the fuel and accelerate transmutation of TRU elements, in order to reduce the nuclear waste. The increased neutron flux may or may not accelerate the oxidization or the hydrogen pickup of the Zircalloy fuel claddings. The precise effect of the irradiation on these degradation phenomena is still unclear, and we must clarify whether or not the increased flux shortens the service life of the claddings. Based on the fact that irradiation induced strain breakaway and hydrogen pickup start at similar dose, we speculate that c-type vacancy loops that induce the strain breakaway are responsible for the nucleation of hydride. Using first-principles calculations, we show that basal surface can be saturated with hydrogen with a binding energy of 0.5eV, and two hydrogen-filled surfaces can close in to a distance only 15% greater than the distance between two successive basal layers in bulk Zr. Closed-in surfaces have a similar atomic structure as ZrH2 and promote further hydrogen absorption to the adjacent layers, acting as a nucleation site for the hydride.
9:00 PM - ES5.13.15
Ab Initio Study of He Embrittlement of Grain Boundaries in Fe-Cr Alloys
Marcin Zemla 1 , Jan Wrobel 1 , Tomasz Wejrzanowski 1 , Duc Nguyen-Manh 2 , Krzysztof Kurzydlowski 1
1 Warsaw University of Technology Warsaw Poland, 2 Culham Centre for Fusion Energy Abingdon United KingdomShow Abstract
Ferritic steels based on Fe-Cr system are perspective structural materials for fusion applications. Helium is produced in the structural materials in nuclear power plants by the nuclear transmutation following the neutron irradiation. Since the solubility of helium in all the metals is extremely low helium tends to be trapped at defects such as vacancies, dislocations and grain boundaries, which causes materials embrittlement. Helium atoms influence both a cohesive energy and magnetic properties of grain boundaries in Fe-based alloys. In these studies, the segregation energy of helium at symmetric Σ3 (111) and Σ5 (210) tilt grain boundaries (GBs) is investigated both in Fe dilute alloys and Fe-Cr alloys as well as in the Fe-Cr structures generated using DFT-based Monte Carlo simulations . In particular, the strengthening / embrittlement energies caused by He and the segregation energies of He are calculated for different concentrations of Cr and He atoms. The results show that the presence of helium atoms strongly influences the magnetic properties of system in the relatively distant neighbourhood. It is observed that the segregation energy of helium atom is influenced by Cr location in GBs. The migration paths and the migration energy barriers of He atoms in GB structures of Fe-Cr alloys are calculated using the Nudged Elastic Band method. Density functional theory calculations are performed by using VASP code, with generalized gradient approximation (GGA) of Perdew-Burke-Ernzerhof (PBE) for exchange-correlation.
The modelling on the atomistic level presented here will enable to improve the knowledge on the phenomena of helium embrittlement and formation of helium bubbles in Fe-based alloys.
 J. S. Wróbel, D. Nguyen-Manh, M. Yu. Lavrentiev, M. Muzyk, S. L. Dudarev, Phys. Rev. B 91, 024108 (2015).
9:00 PM - ES5.13.16
First Principles Theory Calculation of Helium Interaction with Nanoparticles within 14YWT
Yingye Gan 1 , David Hoelzer 2 , Di Yun 3 , Huijuan Zhao 1
1 Clemson University Clemson United States, 2 Oak Ridge National Laboratory Oak Ridge United States, 3 School of Nuclear Science and Technology Xi’an Jiao Tong University Xi'an ChinaShow Abstract
As an advanced oxide dispersion strengthened iron alloy developed through mechanical alloying, 14YWT has exhibited excellent mechanical properties under high temperature, high pressure and high irradiation condition. Other than nanometer size precipitants such as TiN, TiC, and Y2Ti2O7, there exists a high density (>1023) of ultra-fine Y-Ti-O enriched nanoclusters (2~4nm in diameters) within the iron matrix. These oxygen enriched nanoclusters are ultra-stable without coarsening at high temperature, high pressure and high irradiation conditions. After ion/neutron irradiation, the helium bubbles are observed extremely uniform in size (1nm) and quite homogeneously distributed within the matrix and next to these nanoparticles indicating that the microstructure of 14YWT remains remarkably tolerance to radiation damage. With first principles theory calculation, we will perform the energetic study about the helium interaction with different local environments in 14YWT, such as (1) helium atom in iron matrix with/without vacancy; (2) helium atom within a BCC iron unit cell enriched with Y, Ti and O:vacancy pair; (3) helium atom within TiN/TiC matrix with/without vacancy; and (4) helium atom at the coherent/semi-coherent Fe-TiN/TiC interface enriched with vacancy and O:vacancy pair. The objective is to define the preference locations where helium bubbles would like to nucleate, therefore to further understand the formation and growth criteria of helium bubbles within 14YWT.
9:00 PM - ES5.13.17
Phase Field Modeling of Uranium Dioxide Sintering and Densification
Ian Greenquist 1 , Michael Tonks 1 , Yongfeng Zhang 2
1 Mechanical and Nuclear Engineering The Pennsylvania State University State College United States, 2 Idaho National Laboratory Idaho Falls United StatesShow Abstract
In recent years, the phase field approach has become a popular and powerful tool for modeling nuclear materials, including UO2 fuel. Such models are able to efficiently predict fuel properties such as microstructure, heat conductivity, fission-gas release rate, and other properties. However, current fuel-performance phase field codes are often limited by assumptions of the fuel’s initial microstructure. One way to improve the initial microstructures of these models is to develop a phase field model of the fuel manufacturing process. The result of the manufacturing model could then be used as the initial condition of the fuel-performance model.
UO2 fuel pellets are manufactured by sintering to densities of 95% or greater. The remaining porosity has a large impact on fuel performance. Once the fuel is placed in a reactor, it continues to condense due to heat, pressure, and irradiation effects. The current work seeks to build on existing work to develop a mechanistic phase field model that describes densification of UO2 both during sintering and during reactor operation in the MARMOT mesoscale fuel performance code. In this presentation we will illustrate how the model has been implemented in MARMOT and demonstrate how it extends the capabilities of MARMOT to model fuel fabrication.
9:00 PM - ES5.13.18
Interstitial/Substitute Doped Dislocation Core Structures in bcc Materials
Yinan Wang 1 , Ben Xu 1
1 Materials Science and Engineering Tsinghua University Beijing ChinaShow Abstract
A quantum mechanical/molecular mechanical method was employed in the investigation of interactions between impurity atoms and dislocation cores (both screw and edge) in bcc materials, especially iron and tungsten. Hydrogen atoms are particularly considered for tungsten, while Cr, Ni and Cu atoms are considered for iron. The configurations of the dislocation core structures are the results of two competing energies: the interaction between the fractional dislocations and the corresponding generalized stacking fault energy in between the two partial dislocations, which is presented in this work. With this, we can precisely predict the configurations of the interstitial/substitute doped dislocation core structures, which are of significant importance for the future understanding of the dislocation mobile properties.
9:00 PM - ES5.13.19
Radiation-induced Compositional Evolution of Grain Boundaries in 3C-SiC
Xing Wang 1 , Hao Jiang 1 , Cheng Liu 1 , Juan-Carlos Idrobo 2 , Dane Morgan 1 , Paul Voyles 1 , Izabela Szlufarska 1
1 University of Wisconsin-Madison Madison United States, 2 Oak Ridge National Laboratory Oak Ridge United StatesShow Abstract
Grain refinement has been proposed as a promising way to improve the radiation resistance of silicon carbide (SiC), as grain boundaries are known to act as defect sinks. However, the detailed mechanisms for defect annealing at grain boundaries still remain unclear. Characterizing the radiation-induced composition change of grain boundaries can provide important clues for resolving the complicated process of defect-grain boundary interactions. We applied electron energy loss spectroscopy (EELS) to characterize the grain boundary chemistry in CVD (chemical vapor deposition) 3C-SiC bulk samples at different irradiation conditions. The experiments found grain boundaries in CVD 3C-SiC are intrinsically carbon-poor within 2 nm nearby grain boundaries with a relative C composition as low as 45%. This carbon depletion disappeared in samples irradiated to 1 dpa (displacement per atom) at 300 C, which is probably due to the unbalanced flux of C interstitials to defect sinks and low diffusivity of defects at grain boundaries. The C depletion appeared again in samples irradiated at 600 C and the depletion is obviously larger (in about 4nm-wide region near grain boudaries and the relative C composition can be as low as 35%) than that in the non-irradiated sample. This indicates that the role of grain boundaries may change from defect clustering reservoirs at low irradiation temperature to defect diffusion channels at high irradiation temperature. This transition has also been predicted by our ab initio informed rate theory model
9:00 PM - ES5.13.20
Development of a Thermal Conductivity Model for Accident Tolerant UO
2 Reactor Fuel with High Thermal Conductivity Additives Using Multiscale Modeling and Simulation
Floyd Hilty 1 , Michael Tonks 1
1 The Pennsylvania State University State College United StatesShow Abstract
The low thermal conductivity of UO2 ceramics causes large thermal gradients in nuclear fuel pellets, which in turn results in degradation of the fuel from thermal stress induced fracturing and other defects. With this in mind, more accident tolerant reactor fuel is being developed that employs additives to raise the thermal conductivity of UO2 fuel. In this work, we employ the MARMOT mesoscale fuel performance code to evaluate the impact of BeO and SiC additives on the effective thermal conductivity of the fuel. These mesoscale simulation results are then used to inform the development of a thermal resistance model that predicts the effective thermal conductivity as a function of the additive volume fraction and average size. This model can be used in fuel performance codes such as BISON to evaluate the fuel performance of these accident tolerant fuel concepts.
9:00 PM - ES5.13.21
Phase Field Simulations of Solid-State Sintering Behavior of Metallic Uranium Fuel
Bruce Berry 1 , Paul Millett 1
1 Mechanical Engineering University of Arkansas Fayetteville United StatesShow Abstract
We present phase-field simulations of microstructure evolution of metallic Uranium compacts exposed to sintering temperatures. The open source (M)ultiphysics (O)bject-(O)riented (S)imulation (E)nvironment software package is used to perform phase-field simulations of a green compact with varying initital distributions of particle size. The effects of the initial morphology is investigated for the purpose of determining configurations that allow highly stable pore distributions that resist densification and act as capture sites for fission gases. In particular, the sintering behavior of microstructures containing bimodal pore size distributions are studied and found to possess greater resistance to densification relative to monomodal pore sizes. Applications to metallic reactor fuels, prepared by powder metallurgy, that capture fission gases
and improve in-pile swelling are targeted.
9:00 PM - ES5.13.22
Thermal Conductivity Reduction of Tungsten Plasma Facing Material due to Helium Plasma and Cu2+ ion Irradiation
Shuang Cui 1 , Michael Simmonds 1 , Joseph Barton 1 , Yongqiang Wang 2 , George Tynan 1 , Russ Doerner 1 , Renkun Chen 1
1 University of California, San Diego La Jolla United States, 2 Ion Beam Materials Laboratory Los Alamos United StatesShow Abstract
Near-surface region of plasma facing material (PFM) plays an important role in thermal management of fusion reactors. In this work, we measured thermal conductivity of tungsten (W) surface layer damaged by He plasma in PISCES at UCSD and Cu2+ ion beam at LANL.
For He plasma irradiated W, we studied the damage effect on both bulk and thin film of W. The surface morphology of both bulk and thin film was altered after exposure to He plasma with the fluence of 1×1026 m-2 (bulk) and 2×1024 m-2 (thin film). Transmission electron microscopy (TEM) analysis reveals that the depth of the irradiation damaged layer was approximately 20 nm on a bulk W exposed to He plasma at 500 oC for 2000 s (with the fluence of 1×1026 m-2). In order to measure the thermal conductivity of this exceedingly thin damaged layer in the bulk W, we adopted the well-established ‘3-omega’ method and employed novel nanofabrication techniques to improve the measurement sensitivity. For damaged W thin film sample, we measured the reduction in electrical conductivity and used the Wiedemann-Franz (W-F) to extract the thermal conductivity. Results from both measurements showed that thermal conductivity in the damaged layers was reduced by at least ~80% compared to that of undamaged W.
Moreover, Cu2+ ion irradiation with low power density is performed on W to mimic neutron displacement damage. The thermal properties of the damaged layer (~1mm) were measured by the aforementioned adopted 3-omega method. Results show that the thermal conductivity of irradiated surfaces drops from the un-irradiated value of 182±3.3 Wm-1K-1 to 37 ±2.8 Wm-1K-1 for 2dpa, 50±7.8 Wm-1K-1 for 0.6 dpa, 52.5±8 Wm-1K-1 for 0.2 dpa irradiated at 500 oC respectively. For samples damaged at room temperature, decrease in thermal conductivity was observed even at 10-3 dpa. The decrease appears to begin to saturate at a value of roughly 55 W/m-K as the damage level exceeds 10-1 dpa, a factor of ~3x lower than the undamaged value.
The large decrease in thermal conductivity can be attributed to the scattering of electrons, the dominant heat carriers in W, caused by defects created during the He and Cu2+ ion bombardment process. Our result suggests that suppressed thermal conductivity in PFM needs to be taken into account in the thermal design of future plasma-facing components.
9:00 PM - ES5.13.23
Atomic Simulation Investigation on High Entropy Alloy as Possible Accident Tolerant Cladding Material
Alice Hu 1 , Yanhui Zhang 1 , Kayu Fung 1
1 City University of Hong Kong Hong Kong Hong KongShow Abstract
In order to prevent the hydrogen explosion induced radioactive isotopes contamination happened in the Fukushima Daiichi Nuclear Power Plant, we need to design new cladding materials having the following key properties: a reasonably low thermal neutron absorption cross section; a high resistance to radiation damage up to 10 dpa without degradation in property during normal operation; and an excellent high temperature oxidation resistance up to 1000 celsius with a reasonable mechanical strength during the accident conditions. Recent research work has been toward the design of high entropy alloys [1,2] by mixing together the various elements at or near equal molar proportions. Because of this "high-entropy effect", a multi-component alloy is a very stable composition at high temperature. High entropy alloy as a new type of multi-component alloy, not only has a very high intensity (> 1 GPa), but also has excellent toughness (fracture strain of 70%), that is, ultra-high resistance to injury (fracture toughness KIIC＞ 200 MPam1/2). As a result, these multicomponent alloys which have a high strength [1,3,4], sluggish diffusion [5,6], and excellent creep resistance , so that they become promising new materials [8,9] at high temperature.
HEAs’ radiation behavior, however, is scarcely studied although they are potential radiation resistant materials indicated by two intrinsic properties: (a) severe distorted lattice sites in HEAs can trap atoms which may lead to diffusion-limited interstitial-vacancy recombination at high irradiation temperature; (b) low stacking fault energy in some HEAs favors the formation of nanotwins which act as sinks inside each grain to assist interstitial-vacancy recombination too. Thus, swelling in HEAs may be controlled by tuning these two properties to reach an ideal balance of production and annihilation of interstitial and vacancy defects under irradiation, resulting in “self-healing” effect. Atomistic simulation is performed to study radiation defect clusters in HEAs, resulting crystal structure would be detailed analyzed. This investigation will lead to more understandings about the possibility of utilizing HEAs as next generation accident tolerance cladding material.
 Zhang Y, Zuo TT, Tang Z, Gao MC, Dahmen KA, Liaw PK, et al. Prog Mater Sci 2014;61:1e93.
 Yeh JW, Chen SK, Lin SJ, Gan JY, Chin TS, Shun TT, et al. Adv Eng Mater 2004;6:299e303.
 He JY, Liu WH, Wang H, Wu Y, Liu XJ, Nieh TG, et al. Acta Mater 2014;62:105e13.
 Zou Y, Maiti S, Steurer W, Spolenak R. Acta Mater 2014;65:85e97.
 Tsai KY, Tsai MH, Yeh JW. Acta Mater 2013;61:4887e97.
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 Senkov ON, Wilks GB, Miracle DB, Chuang CP, Liaw PK. Intermetallics 2010;18:1758e65.
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 Ye YF, Wang Q, Lu J, Liu CT, Yang Y. Intermetallics 2015;59:75.
9:00 PM - ES5.13.24
Effect of Solute Atoms on Radiation Induced Grain Boundary Creep
Yinon Ashkenazy 1 , Noya Dimanstein 1
1 Hebrew University of Jerusalem Jerusalem IsraelShow Abstract
Irradiation of crystalline structure in conjunction with stress causes the materials to slowly deform (“creep”), resulting in changes to the material’s dimensions and strength. One of the options to limit creep in coarse grained materials is to decrease grain size down to sizes where dislocation mobility is prevented and point defect sinks are abundant. It was previously demonstrated that in such systems creep is controlled by local “string” like atomic relaxations within the grain boundaries as they absorb point defects, e.g., interstitials and vacancies, produced in the grain interior during irradiation. We demonstrate that in such systems grain boundary impurities have strong effect on observed radiation induced creep rates while having insignificant effect on thermal relaxations. Characterization of creep related relaxations within the grain boundary and correlation between these relaxations for various immiscible impurities serve as a basis for creating a mean field model for irradiation induced creep. The model predictions are used to describe various segregation scenarios and sample structures.
 Ashkenazy, Yinon, and Robert S. Averback, "Irradiation Induced Grain Boundary Flow—A New Creep Mechanism at the Nanoscale." Nano Letters 12(8): 4084–4089
9:00 PM - ES5.13.25
Energies and Forces of Elastic Interaction between Nano-Scale Defects in Irradiated Materials
Sergei Dudarev 1 2 , Adrian Sutton 2
1 UK Atomic Energy Authority Abingdon United Kingdom, 2 Department of Physics Imperial College London London United KingdomShow Abstract