Radiation Effects in Metals and Ceramics: Poster Session II
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Djamel Kaoumi, North Carolina State University; Thak Sang Byun, Oak Ridge National Laboratory; Dane Morgan, University of Wisconsin-Madison; Maria Okuniewski, Purdue University; Mahmood Mamivand; Geoffrey Beausoleil, Idaho National Laboratory; Philip Edmondson, The University Of Manchester; Khalid Hattar, University of Tennessee Knoxville; Aurelie Gentils, Université Paris-Saclay; Joel Ribis, Cea

Monday 5:30 PM
February 24, 2020
Room: Sails Pavilion
Location: San Diego Convention Ctr


H-36: Combined Use of In-situ Ion Irradiation and In-situ Nanomechanical Testing for Characterizing Helium Pre-implanted 304 Stainless Steel: Ce Zheng1; David Frazer2; Peter Hosemann2; Djamel Kaoumi1; 1North Carolina State University; 2University of California, Berkeley
    In this study, Helium ions were pre-implanted into pre-FIBed thin pillars of 304 stainless steel in the range of 1017-1018 He/cm2 using an Orion Nanofab He ion beam microscope. The subsequent irradiation was conducted in-situ in a transmission electron microscope using 1 MeV Kr2+ ions up to 5 dpa at 300°C. The subsequent mechanical compression testing on the He pre-implanted and ion irradiated pillars was also conducted in-situ in an electron microscope using a PI-95 TEM holder. Snapshots, video and stress-strain curve were recorded for each tested pillar. The in-situ irradiation allowed to follow the bubble formation and growth in the micro-pillars, and the in-situ mechanical testing allowed to evaluate in-situ the radiation hardening due to radiation damage and the impact of the He bubbles. The presentation will showcase the method and the results will discussed.

H-37: Coupled Bulk and Grain Boundary Compositional Patterning in Binary Immiscible Alloy under Irradiation: A Phase Field Modeling Study: Qun Li1; Pascal Bellon1; Robert Averback1; 1University of Illinois at Urbana–Champaign
    Irradiation of moderately immiscible alloy systems results in competing effects, namely randomization of the composition by ballistic mixing and phase separation driven by thermodynamics. Past modeling and experiments have shown that this competition can lead to bulk compositional patterning, i.e., the stabilization of finite-size intragranular precipitates at steady state. Here we employ phase field modeling to investigate the coupled microstructural evolution of grain interiors and grain boundaries. We find that irradiation can lead to compositional patterning at grain boundaries, and remarkably that it can coexist with bulk patterning. GB precipitates are larger than intragranular ones owing to faster diffusion along GBs. We analyze these results by building steady-state maps as a function of the nominal composition and the forcing intensity γb, the ratio of ballistic mixing rate to thermal mobility. The double grain interior-GB patterning suggests that suitable coupling of these sub-systems can result in self-organization of the whole microstructure.

H-39: Doppler Broadening Positron Annihilation Spectroscopy for Understanding Void Formation in Neutron Irradiated Fe-Cr Alloys: Carly Romnes1; Ming Liu2; James Stubbins1; 1University of Illinois at Urbana-Champaign; 2North Carolina State University
    Positron annihilation spectroscopy (PAS) is a powerful, non-destructive technique used to characterize the defect properties of materials. This work focused on a type of PAS called Doppler Broadening Spectroscopy (DBS). Several samples were irradiated at a wide range of conditions at the Advanced Test Reactor at Idaho National Laboratory. DBS was utilized to better understand materials exposed to low doses at irradiation temperatures of 300°C, 450°C, and 550°C. The goal of this work is to develop a fundamental understanding of the evolution of nanoscale and sub-nanoscale defect clusters in neutron irradiated Fe-Cr alloys. These defect clusters are too small to identify using other experimental microstructural analysis techniques, such as high-resolution transmission electron microscopy, and are not stable enough to identify using atom probe tomography, which makes PAS an ideal technique for this work. PAS is essential for developing a fundamental understanding of void nucleation and void incubation processes.

H-40: Effect of Ordered Helium Bubbles on the Deformation and Fracture Behavior in Zr: Liu Simian1; Han Weizhong1; 1Xi‘an Jiaotong University
    Radiation-induced helium bubbles are detrimental to the mechanical properties of metals, usually causing severe hardening and embrittlement. In this study, we investigated the effect of ordered helium bubble on plastic deformation and the failure of the α-Zr. Helium bubbles prefer to align along the basal plane in α-Zr. Micro-scale in situ tensile tests revealed that helium bubbles less than 8 nm in size can increase the critical shear stress for the prismatic slip; larger bubbles have a higher resistance. However, once the helium bubbles become larger than 8 nm, a bubble-softening effect appears due to an increase in bubble spacing and porosity. Once the Schmid factor for the basal slip is considerably higher than the prismatic slip, the bubble coalescence along the basal plane becomes the major failure model for helium-irradiated α-Zr, rather than the activating prismatic or pyramidal slip.

H-41: Elucidating the Role of Dispersoids on the Bulk and Nano-mechanical Properties of Dispersion-strengthened W Alloys Following Ion Irradiation with In-situ Characterization: Eric Lang1; Quentin Rizzardi1; Robert Maass2; Jean Paul Allain3; 1University of Illinois; 2University of Illinois at Urbana Champaign; 3The Pennsylvania State University
    Tungsten (W) is the plasma-facing material of choice for future plasma-burning tokamak fusion devices for its high melting point and high sputter threshold. However, monolithic W is intrinsically brittle and its low recrystallization temperature poses complications in fusion environments. In this work, dispersion-strengthened W samples are processed via spark plasma sintering with TiC, ZrC, and TaC dispersoids alloyed from 0.5 to 10 wt. % to synthesize fine-grained W alloys with inter- and intra-granular dispersoids strengthening the W matrix. The effect of the dispersoid microstructure on the mechanical performance is systematically studied. Bulk studies elucidate greater recrystallization inhibition and ductility than pure W. Nano-mechanical response is studied via in-situ SEM techniques. Nanoindentation and micro-pillar compression is studied to extract stress-strain relations and understand dispersoid-microstructure stability. Correlations with irradiation parameters including: low energy, high temperature D/He ion irradiation will elucidate the synergy between ion damage and nano-mechanical response.

H-42: Evaluation of Bubble Layers in Single- and Poly-crystal Tungsten after Helium Exposure: Daniel Morrall1; Cierra DellaRova2; Russell Doerner3; Matthew Baldwin3; Chad Parish1; 1Oak Ridge National Laboratory; 2Colorado School of Mines; 3University of California, San Diego
    Understanding the He (D-T) plasma surface interactions of tungsten, the primary candidate material for use as divertor, will be key to the future of fusion reactors. In order to evaluate the performance of the tungsten surface, we exposed both single- and poly-crystalline tungsten pucks to helium plasma at PISCES-A at UC-San Diego over a temperature range of ~300-600 K (polycrystalline) and ~873 K (001 single crystals), in order to study bubble formation and evolution. These temperatures were chosen to prevent surface morphology (i.e., nanofuzz) from forming. We found that in the polycrystalline tungsten samples, a thin, native oxide layer appeared to be co-located with the helium layer. We will draw conclusions about the interactions between oxide and helium layers. In the single crystals, we measured the bubble size, density, and morphology from ~1024 to 1026 He/m2. Nanoindentation was used to evaluate the influence on the near-surface mechanical properties.

H-43: Exploring Stability of Nanocrystalline Metals with Competing Solute Effects under High Temperature Irradiation: Christopher Barr1; Patrick Price1; Nathan Heckman1; Brad Boyce1; Khalid Hattar1; 1Sandia National Laboratories
    Nanocrystalline metals have shown significant radiation tolerance for potential application in next generation nuclear energy systems. However, a major shortfall is low homologous temperature grain coarsening. To mitigate grain growth, researchers have developed thermodynamic and kinetic stabilization routes to stabilize the grain size through solute additions. While this has been successful in minimizing grain growth under thermal conditions, it is still unknown how these stabilization mechanisms evolve under irradiation. In this study, we explore the combined effect of both thermal exposure and irradiation in nanocrystalline Pt and a Pt-Au alloy. Under high temperature thermal conditions, grain boundary Au enrichment is shown to stabilize the grain size. In contrast, under high temperature heavy ion irradiation, significant grain growth was observed associated with de-mixing of Au solutes at grain boundaries. The competing thermal and irradiation effects in Pt-Au highlight the careful consideration needed for designing nanocrystalline alloys for combined extreme environments.

H-44: Generalized Dislocation Mobility Law for BCC FeCrAl Alloys: Sanjoy Mazumder1; Raven Maccione1; Yash Pachaury1; Janelle Wharry1; Anter El-Azab1; Tomohisa Kumagai2; 1Purdue University; 2Central Research Institute of Electric Power Industry, Japan
    We aim to construct a generalized mobility law for edge, screw and mixed dislocations in BCC alloys. The mobility of dislocations in BCC alloys is influenced by their relatively open crystal structure and the non-planar nature of the dislocation core. In addition, the presence of solute atoms and clusters act as obstacles to dislocation motion. We have used Molecular Dynamics (MD) simulations to compute the dislocation mobility in FeCrAl alloys at various compositions and temperatures. It was observed that at 300K and an applied shear stress in the range 500-700 MPa, <111>{110} edge dislocations in Fe-13%Cr-5%Al undergo stick and slip motion due to pinning and de-pinning by the solute atoms. This stick and slip motion becomes less significant at higher temperatures and the extent to which this type of motion persists at those temperatures also depends on stress. Our MD data was analyzed and cast into a generalized mobility law.

H-46: How to Improve an Irradation-simulation Testbed: Younggak Shin1; Sanghyuk Yoo2; Keonwook Kang2; Byeongchan Lee1; 1Kyung Hee University; 2Yonsei University
     Irradiation damages are often simulated with atomistic calculations but the validity depends on several factors: interatomic potentials are known to be a key factor. On the other hand, the importance of numerical setups has been underestimated. Here, we introduce the concept of a “simulation testbed.” A simulation testbed is a structured procedure from preparation of numerical samples to execution of atomistic calculations. First, we prepare numerical samples as close as possible to thin-film tungsten samples with thickness of 370 nm or less. Second, recoil events inside numerical samples are determined based on neutron-transport statistics. Third, constant-energy molecular dynamics is performed with the periodic boundary condition in the direction perpendicular to the free surface.We present the results from billion-atom molecular dynamics results of thin-film tungsten under irradiation, and compare the damages and changes in mechanical properties with experimental measurements. Further possibility of improvement is also discussed.

H-48: Influence of Grain Size and the Presence of Nano Oxides on the Radiation Resistance of a FeCrW Alloy: Bertrand Radiguet1; Auriane Etienne1; Cristelle Pareige1; Nariman Enikeev2; Constantinos Hatzoglou1; Maria Vrellou1; Julia Ivanisenko3; 1University Of Rouen; 2Ufa State Aviation Technical University; 3Karlsruhe Institute of Technology
    In order to investigate the influence of defect sinks on the stability of the microstructure of a Fe-14Cr-1W-0.3Mn-0.2Ni-0.3Si alloy under irradiation, this alloy was studied in hot-extruded state (coarse grain (CG) alloy), after severe plastic deformation, resulting in grains of about 100nm (ultra fine grain (UFG) alloy) and after hot extrusion of the master alloy reinforced by nano-oxides (ODS alloy). CG and ODS alloys were elaborated at CEA Saclay. The UFG alloy was obtained by high pressure torsion (HPT) at USATU in Ufa. The 3 alloys were ion irradiated at 400°C in JANNUS facilities. Their microstructures were characterized by atom probe tomography in order to quantify Cr precipitation, Mn, Ni and Si clustering and to evaluate the stability of the nano-oxides in the case of the ODS alloy.

H-49: Investigating Radiation Damage in Metallic and Ceramic Materials for Advanced Nuclear Systems Using JANNuS Multiple Ion Beams: Aurelie Gentils1; Celine Cabet2; 1CSNSM, Univ Paris-Sud and CNRS/IN2P3, Université Paris-Saclay; 2DEN, Service de Recherches de Metallurgie Physique, CEA, Université Paris-Saclay
    Ion accelerators have been used by material scientists for decades to investigate radiation damage formation in nuclear materials and thus to emulate neutron-induced changes. In France, under the auspices of the Université Paris-Saclay, the JANNuS platform for Joint Accelerators for Nanosciences and Nuclear Simulation comprises five ion implanter and accelerators with complementary performances. At CSNSM Orsay, a Transmission Electron Microscope is coupled to an accelerator and an implanter for in situ observation of microstructure modifications induced by ion beams. At CEA Saclay, the unique triple beam facility in Europe allows the simultaneous irradiation with ions for nuclear recoil damage and implantation of a large array of ions including gasses for well-controlled modelling-oriented experiments. Various examples of metallic and ceramic materials for nuclear applications will illustrate the use of JANNuS ion beams in investigating the radiation resistance of materials for today’s and tomorrow’s nuclear reactors and for waste management.

H-50: Ion-irradiation-induced Structural Disorder and Thermal Conductivity Changes of Intermetallic Compounds: Shradha Agarwal1; Andy Nelson2; Steven Zinkle3; 1University of Tennessee and Oak Ridge National Laboratory; 2Oak Ridge National Laboratory; 3University of Tennessee
     The ordered intermetallic compounds are potential candidates for high-temperature components for both fission and fusion reactors. The irradiation-induced amorphization and structural disorder is a common phenomenon known for this class of materials. The knowledge of experimentally measured critical dose and temperature for amorphization is different for each type of intermetallic compound and is vital for their application. Most importantly, how this disorder and amorphization affects the thermal conductivity of this material is crucial, especially with regard to their application in fuel systems. Under this context, five different intermetallics compounds, namely FeAl, TiAl, NiAl, Ni3Al, and Fe3Al were damaged using 21 MeV Au ions at 1 dpa dose. The irradiations were conducted at very low temperature ~ 80 K. The samples were then investigated using thermal conductivity microscope to evaluate irradiation induced changes in the property. The structural disorder and amorphization were examined using transmission electron microscopy and X-ray diffraction. .

H-51: Ion Irradiation Effects on the Microstructure of PM-HIP Inconel 625: Caleb Clement1; Janelle Wharry1; Xiang Liu2; Megha Dubey3; David Gandy4; 1Purdue University; 2Idaho National Laboratory; 3Boise State University, Center for Advanced Energy Studies; 4Electric Power Research Institute
    The objective of this talk is to compare the microstructure of an ion irradiated Ni-based alloy produced by powder metallurgy with hot isostatic pressing (PM-HIP), and its traditionally manufactured counterpart. PM-HIP is an attractive alternative to castings and forgings in the nuclear industry because components exhibit chemical homogeneity and are produced near-net shape, reducing the need for machining and welding. Irradiation effects must be understood before PM-HIP materials can be employed in nuclear reactors. This work focuses on PM-HIP and cast Inconel 625, self-ion irradiated to a dose of 100 displacements per atom (dpa) at temperatures 400-500°C. Transmission electron microscopy (TEM) is used to characterize irradiation-induced voids, dislocation loops, and changes in pre-existing precipitates. The irradiated microstructures are compared between the PM-HIP and cast material. Results will be discussed in the context of an ongoing PM-HIP neutron irradiation campaign and the progress toward qualification of PM-HIP alloys for nuclear service.

H-52: Irradiation Induced Damage Evolution in Tungsten: Trevor Clark1; Suveen Mathaudhu2; Samuel Briggs3; Robert Dowding4; Jason Trelewicz5; Khalid Hattar1; 1Sandia National Laboratories; 2University of California, Riverside; 3Oregon State University; 4United States Army Research Laboratory; 5Stony Brook University
    The extreme radiation and thermal environments that plasma facing materials for nuclear fusion energy applications are difficult to experimentally explore making it difficult to elucidate the underlying mechanisms or develop predictive modeling. This presentation will present the in-situ ion irradiation transmission electron microscope and the recently developed in-situ ion irradiation scanning electron microscope that have been developed at Sandia to explore these combinations of extreme environments. Initial TEM and SEM results investigating the response of tungsten as a function of processing and microstructure under isolated and combined heating (>1,000 °C), low energy helium implantation, and displacement damage events will be highlighted and compared to the current literature. The effects of grain morphology on surface damage evolution are explored. SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525.

H-53: Material Irradiation and Investigation Capabilities at TRIUMF: Ferran Boix Pamies1; Alexander Gottberg1; 1TRIUMF
    ISAC-TRIUMF operates targets under particle irradiation in the high-power regime of 50 kW to produce radioactive isotope beams using the ISOL method. High-energy protons passing through the ISOL target are available for studies of radiation damage in materials with unique versatility in radiation type and intensity, nature of samples and irradiation conditions. In addition, the existing hot-cell capabilities for the routine maintenance of target components allow immediate in-situ material characterization of highly activated samples as well as preparation of microscopic samples with dose rates that allow hands on material investigation. The research aims to study materials for use in components for accelerator, fusion and fission applications that require specific thermoelectrical or structural properties in a wide range of operating temperatures and radiation fields. The ongoing capabilities and plans for material irradiations will be presented with the achievable DPA levels, proton and neutron fluxes and material characterization capabilities.

H-54: Mechanical Response of FeCr Alloys under Thermal Aging and Irradiation: Pengcheng Zhu1; Yajie Zhao1; Shradha Agarwal1; Steven Zinkle2; 1The University of Tennessee, Knoxville; 2The University of Tennessee
    High chromium ferritic/martensitic steels are promising candidate structural materials for Generation IV fission and fusion reactors. Fe18%Cr and Fe25%Cr were subjected to thermal aging (100-900 hours at 475 °C) to produce Cr-rich α’ precipitates of varying size and density. Nanoindentation hardness testing was performed with Berkovich and spherical indenters to obtain information on yield and ultimate strength. The Berkovich indentations provide information on the bulk hardness of nanoscale irradiated volumes, while spherical indentations provide work hardening capacity information. The aged materials were subsequently irradiated with 8 MeV Fe+ ions (0.35 dpa at 475 °C, dose rates of ~10-5 and 10-4 dpa/s), and the nanoindentation response was compared with the thermal aged specimens. The results of TEM and APT characterization of the α’ precipitates will be summarized, and detailed comparisons will be provided on the predicted (dispersed barrier hardening superposition) vs. measured strength values of the aged and irradiated specimens.

H-55: Microstructural Response of FeCr/Y2O3 Bilayer System to He/H Implantation: Olga Emelianova1; Aurelie Gentils2; Maria Ganchenkova3; Amir Gumarov4; Igor Yanilkin4; Iskander Vakhitov4; Igor Golovchanskiy5; Igor Shchetinin5; Lenar Tagirov6; Vladimir Borodin7; 1CSNSM, Univ Paris-Sud, CNRS/IN2P3, Université Paris-Saclay and National Research Nuclear University MEPhI; 2CSNSM, Univ Paris-Sud, CNRS/IN2P3, Université Paris-Saclay; 3National Research Nuclear University MEPhI; 4Kazan Federal University; 5National University of Science and Technology MISIS; 6Zavoisky Physical-Technical Institute, FRC Kazan Scientific Center of RAS; 7NRC Kurchatov Institute and National Research Nuclear University MEPhI
    The high radiation tolerance of oxide-dispersion strengthened steel is often attributed to the efficient trapping of helium and hydrogen at oxide/matrix interfaces. A clearer picture of the involved mechanisms can be obtained using model FeCr/Y2O3 thin films. Such bilayers were implanted with helium, hydrogen and krypton ions and characterized using TEM techniques. At room temperature, the implantation of either He or H creates nanometer-size cavities in the metal, oxide and at the interface. Consecutive He/H implantation leads to clear synergetic effects and interfacial de-cohesion. Room temperature Kr+ irradiation creates small features, presumably cavities, in the metal only. Finally, helium implantation at 550°C produces in the metal a typical gas cavity pattern with larger cavities at grain boundaries, but no changes in the oxide and at the interface. The implications of results for clarifying the He and H impact on the microstructural evolution at and near the metal/oxide interface are discussed.

H-56: Modeling Slip-induced Crack Initiation in Nickel Containing Nano-scale Helium Bubbles: Tung Yan Liu1; Michael Demkowicz1; 1Texas A&M University
    We present molecular dynamics simulations of plastic deformation in nickel containing nano-scale helium bubbles to study the mechanism of slip-induced crack initiation. We show results from models that are built according to experimental data of defect structures developed under irradiation in a nuclear reactor. We focus on examining the effect of slip band formation in nickel with interstitials clusters, bubbles, grain boundaries and hard inclusions, which may serve as precursors of crack initiation. We discuss the implications of our work for understanding the mechanism of deformation and fracture in irradiated materials, as well as techniques of modelling material property evolution.

H-57: Molecular Dynamics Simulation of Irradiation Damage in Disordered Alloys with Ordered Precipitation: Shijun Zhao1; 1City University of Hong Kong
    Chemically disordered concentrated solid-solution alloys, including high-entropy alloys (HEAs), have demonstrated good mechanical properties and promising irradiation resistance depending on their compositions. Recently, it has been shown that ordered precipitates introduced into the HEA matrix can improve the mechanical performance significantly, due to the formation of L12 phase that affects dislocation movement during deformation. Here we employ molecular dynamics simulations to study how these ordered phases can influence the evolution of irradiation damage during accumulated cascade conditions. We studied two systems, i.e. Ni-Fe and Ni-Al, with L12 ordered phases embedded into the disordered matrix. Our results show that the size and density and the ordered phase affect the irradiation damage states of the alloy, and the ordered phase can reduce defect diffusivity which is related to the specific defect energetics in these two systems.

H-58: Neutron Radiation Induced Patterning of Fe-Cr System: A Phase-field Approach: Bohyun Yoon1; Jeongwhan Lee1; Kunok Chang1; 1Kyung Hee University
    Fe-Cr Ferritic/Martensitic steel is one of the promising structural materials for the nuclear reactor due to their high strength and fracture toughness. During the nuclear power plants operating, the structural material undergoes the neutron irradiation, it means the material degrades. One of the well known degradation mechanism of Fe-Cr system under the neutron irradiation is radiation induced segregation (depletion) in the vicinity of the grain boundary. We found that due to the depletion of the chromium atom near the grain boundary, patterning of chromium rich phase takes place which is consistent with former experimental observation and other computational predictions. We developed the GPU (Graphic Processor Unit) accelerated phase-field code to simulate the microstructural evolution of Fe-Cr system. The effect of the elastic interaction in the radiation induced patterning will be discussed.

H-60: Radiation Damage Mechanisms in the Oxides Formed on Zr Alloys: Junliang Liu1; Guanze He1; Anamul Mir2; Jing Hu3; Stephen Donnelly2; Meimei Li3; Sergio Lozano-Perez1; Chris Grovenor1; 1University of Oxford; 2University of Huddersfield; 3Argonne National Laboratory
    The structure of the complex nanoscale oxides formed on Zr alloys by aqueous corrosion can directly affect the corrosion performance, especially under neutron irradiation. In this work, we have used in-situ TEM and transmission-EBSD to study radiation damage mechanisms in the oxides formed on Zr alloys. The aim has been to determine the relative stability of the different oxide phases formed by corrosion. We report for the first time on the susceptibility to radiation damage of the suboxide phase which may influence the nucleation of new oxide grains and the transportation of oxidation species across the oxide/metal interface, and lead to enhanced corrosion rates. A monoclinic-to-cubic transformation of the bulk oxide is also observed by in-situ experiments, followed by irradiation-induced grain growth. The formation of cubic oxide phase thus needs to be considered as a possible process that can occur at high burnups, and further affect the corrosion rates.

H-61: Radiation Tolerance in Stabilized Alumina Coatings: an In-situ Irradiation Study: Matteo Vanazzi1; Davide Loiacono2; Wei-Ying Chen3; Meimei Li3; Marco G. Beghi2; Fabio Di Fonzo1; 1Center for Nano Science and Technology (CNST) - IIT; 2Politecnico di Milano; 3Argonne National Laboratoy
    Innovative reactors require new material strategies. Recently, coatings have earned great interest, since they could tackle major issues like corrosion or fretting. In this framework, alumina coatings by IIT have been characterized as corrosion-resistant barriers and tested under ion irradiation, being stable up to 150 dpa. The radiation tolerance is related to the amorphousness of the material, which crystallizes under irradiation. To preserve this tolerance and improve the range of operation, the amorphous matrix must be conserved. Here, the stabilization of alumina by doping is evaluated. Irradiation is performed with in situ TEM at the IVEM-Tandem facility (ANL). Samples are irradiated with different ions, up to 20 dpa. Experiments are performed at 600 and 800 °C. Doped alumina retards the crystallization, even under irradiation. Doping stabilizes the metastable phases of alumina, preserving the pristine characteristics. Moreover, while pure alumina suffers from voids formation, no swelling appears in the doped counterpart.

H-62: Radiation Tolerance of Gradient Grain-structured Copper: Heather Salvador1; Yiwei Sun1; Trevor Clark2; Khalid Hattar2; Sina Shahrezaei3; Suveen Mathaudhu1; 1University of California, Riverside; 2Sandia National Laboratories; 3Pacific Northwest National Laboratory
    Radiation tolerance of metals and alloys has been studied as a function of grains size with studies showing enhanced radiation defect tolerance in nanocrystalline materials through increased grain boundary area, while other reports show a loss of radiation tolerance due to the loss of grain boundary stability from the highly energetic nature of nanocrystalline grains. These findings point to using a heterogeneous grain structure to achieve a balance of beneficial properties and microstructural stability. In this study, we investigate the microstructural evolution and radiation damage in gradient grain size Cu exhibiting a grain distribution of <50 nm to ~0.5 mm. Through in situ Cu-ion irradiation, electron microscopy, and hardness measurements parallel and perpendicular to the gradient, damage evolution and resulting mechanical response can be explored as a function of grain size. Advantages and disadvantages of gradient grain-structured alloys will be discussed base on the experimental observations.

H-63: Small Scale Tensile Testing of Grain Boundary Strength of Pristine and Neutron Irradiated Ni Based X-750 Alloy: Yachun Wang1; Xiang Liu1; Daniel Murray1; Fei Teng1; Mukesh Bachhav1; Wen Jiang1; Lingfeng He1; Cheng Sun1; Ziqi Xiao2; Xian-Ming Bai2; John Jackson1; Robert Carter3; 1Idaho National Laboratory; 2Virginia Polytechnic Institute and State University; 3Electric Power Research Institute, Inc
    The degradation of grain boundary (GB) strength under irradiation conditions can cause intergranular cracks inside reactor structural materials and pose a safety threat. Understanding the relationship between GB character and its strength in neutron irradiated Ni based X-750 alloy can shed insight into the impact of irradiation on the grain boundary strength. This work aims, by combining advanced characterizations, small-scale tensile test, and modeling, to quantitatively study effects of element segregation/depletion and precipitates on GB strength in neutron-irradiated and non-irradiated X-750 alloys. The GB microstructure was characterized by utilizing Electron Backscatter Diffraction (EBSD), Transmission Electron Microscope (TEM) equipped with Energy Dispersive x-ray Spectroscopy (EDS) and Atom Probe Tomography (APT). Evaluation on the strength of single GB of different types has been performed through small-scale tensile testing and modeling.

H-65: Tungsten Transmutation Products from Mixed Spectrum Neutron Irradiation: Nathan Reid1; Lauren Garrison2; Yutai Katoh2; Jean Paul Allain1; 1University of Illinois at Urbana-Champaign; 2Oak Ridge National Laboratory
    Tungsten (W) and W-alloys are being considered as armor material for future fusion device plasma facing components (PFCs). Neutrons produced from the fusion reaction will cause transmutation reactions in W-alloys until precipitation of near-neighbors in the periodic table begin to cause embrittlement and degradation of thermal conductivity. The amount of transmutation and thus property changes critically depends on the neutron energy spectrum. Tungsten irradiated in a high transmutation spectrum, the flux trap of the High Flux Isotope Reactor, is compared with tungsten irradiated inside a shield to reduce transmutation. A detailed distribution of elements in neutron-irradiated W is obtained using glow-discharge optical emission spectroscopy (GD-OES). The composition is measured by controlled sputtering of surface layers, while simultaneously ionizing and measuring the optical emission of the eroded material. Depth is determined with in-operando laser interferometry to have quantification of composition with depth.