Microstructural Processes in Irradiated Materials: Poster Session
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Thak Sang Byun, Pacific Northwest National Laboratory; Chu-Chun Fu, Commissariat ŕ l'énergie atomique et aux énergies alternatives (CEA); Djamel Kaoumi, University of South Carolina; Dane Morgan, University of Wisconsin-Madison; Mahmood Mamivand, University of Wisconsin-Madison; Yasuyoshi Nagai, Tohoku University

Monday 6:00 PM
February 27, 2017
Room: Hall B1
Location: San Diego Convention Ctr

Session Chair: Mahmood Mamivand, University of Wisconsin-Madison; Ryuta Kasada, Kyoto University


G-18: Atom Probe Tomography Investigations of Archival Surveillance Steels from the UCSB ATR-2 Irradiation: Nathan Almirall1; Peter Wells1; Takuya Yamamoto1; David Gragg1; Kirk Fields1; G. R. Odette1; Randy Nanstad2; Philip Edmondson2; 1University of California Santa Barbara; 2Oak Ridge National Laboratory
    Extended operation of nuclear reactors to 80 years will require ensuring the integrity of the reactor pressure vessel (RPV). RPV steels experience neutron irradiation-induced hardening and corresponding ductile brittle transition temperature shifts (TTS). One of the goals of the high fluence intermediate flux UCSB ATR-2 irradiation experiment is to support development of physically based models of TTS for high-fluence-low flux conditions (φt >10^20 n/cm2 >1 MeV, φ < 10^11 n/cm2-s) beyond the existing surveillance database. This presentation covers Atom Probe Tomography (APT) investigations on a matrix of archival surveillance materials acquired from various operating LWRS for UCSB ATR-2. A particular focus will be on comparisons to previous test reactor irradiations and surveillance programs to characterize the effect of a wide range of flux and fluence on the volume fraction, number density, average size, composition and character of precipitates and defect-solute complexes that form under irradiation.

G-20: Characterization of Nanoscale Intermetallic Precipitates in Highly Neutron Irradiated Reactor Pressure Vessel Steels: David Sprouster1; E Dooryhee1; S Ghose1; P Wells2; T Stan2; N Almirall2; G. R. Odette2; L Ecker1; 1Brookhaven National Laboratory; 2University of California Santa Barbara
    Massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons. The neutrons cause embrittlement of the vessel steel that increases with dose and service time, as manifested by an increase in the temperature of transition from ductile-to-brittle fracture. Here, we discuss our recent efforts employing synchrotron-based measurements to characterize the nm-scale precipitates that develop at high fluence. Data from the recently completed Advanced Test Reactor-2 (ATR-2) irradiation tests will be presented, where a wide range of materials with different chemical compositions were irradiated at ≈ 290°C to high fluence ≈ 1.4× 1020 n/cm2. We will also discuss developments in our dedicated high-throughput capabilities (experiment and analysis), with particular emphasis on the small angle X-ray scattering measurements. Our results will be compared to complementary characterization methods and highlight the benefits of rapid measurements to examine microstructure evolution in this important structural material.

G-21: Development of Standard Protocols for the Analysis of Atom Probe Data of Radiation Damage in Light Water Reactors: Bertrand Radiguet1; Gérald Da Costa1; John Hyde2; Constantinos Hatzoglou1; Hannah Weekes2; Paul Styman2; François Vurpillot1; Cristelle Pareige1; Auriane Etienne1; Giovanni Bonny3; Nicolas Castin3; Lorenzo Malerba3; Philippe Pareige1; 1GPM UMR CNRS 6634 - Université et INSA de Rouen; 2National Nuclear Laboratory; 3SCK-CEN
    Neutron irradiation results in hardening and embrittlement of reactor pressure vessel steels, due to the formation of nano-scale solute clusters. Thanks to its ability to identify chemical heterogeneities in 3D down to atom scale, atom probe tomography can provide microstructural data that underpin development of atomistic models of radiation damage and provide explanations of the evolution of mechanical properties. However, multiple methods for analysing the atom probe data exist. This can result in un-necessary scatter, even inconsistencies, between results obtained from different groups. This makes interpretation of results and calibration of radiation damage models challenging. In this work simulations of a range of different microstructures are used to directly compare different cluster analysis algorithms and identify their strengths and weaknesses.

G-22: Effect of Helium/dpa Ratio on Microstructure Evolution in Dual Ion Irradiated HT9 Steel: David Woodley1; Elizabeth Getto1; Zhijie Jiao1; Kai Sun1; Gary Was1; 1University of Michigan
    Ferritic-martensitic steels are leading candidates for structural materials in next generation reactors due to their resistance to swelling. The process of cavity nucleation and growth is not well understood but is known to be influenced by temperature and helium generation from transmutation reactions. Dual ion irradiations were performed at the Michigan Ion Beam Laboratory on alloy HT9. A defocused 5 MeV Fe++ beam and a degraded ~2 MeV He++ beam were used for irradiations to a damage level of 188 dpa at temperatures from 440-480oC with helium/dpa ratios from 0 to 0.2 appm He/dpa. The effects of temperature and helium/dpa ratio on cavity, precipitate and dislocation loop evolution were analyzed using advanced electron microscopy. Increasing temperature resulted in an increased diameter and a decreased density for all features. Increasing helium/dpa ratio was found to increase diameter and decrease density for cavities and have minimal effect on most other features.

G-23: Energetic Study of Helium – Nanoparticle Interaction within Nanostructured Ferritic Alloy: Yingye Gan1; Huijuan Zhao1; David Hoelzer2; Di Yun3; 1Clemson University; 2Oak Ridge National Laboratory; 3Xi’an Jiao Tong University
    Advanced nanostructured ferritic alloys have been developed as structural materials for nuclear reactor applications. Their enhanced mechanical properties under high temperature, pressure and irradiation condition are mainly due to the ultrahigh density of nano-particles within the iron matrix. Taking the irradiated 14YWT as an example, the helium bubbles are observed extremely uniform in size (1nm) and quite homogeneously distributed adjacent to these nanoparticles. With first principles theory calculation, we investigate the formation energy variations of helium atom with the following conditions (1) in pure iron matrix with/without vacancy; (2) in iron matrix enriched with Y, Ti and O:vacancy pair; (3) in TiN/TiC precipitants with/without vacancy; and (4) 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.

G-24: Evolution of Irradiation-induced Precipitates in Reactor Pressure Vessel Steels under High-Dose Irradiation: Mikhail Sokolov1; Michael Miller1; Randy Nanstad1; Ken Littrell1; Lynne Ecker2; David Sprouster2; Enrico Lucon3; 1ORNL; 2BNL; 3NIST
    Some of the current fleet of nuclear power plants is poised to reach their end of life and will require an operating life time extension. Therefore, the reactor pressure vessel will be subject to higher neutron exposures. In this study, several RPV steels were irradiated at high doses to study the degradation of the mechanical properties and related microstructural changes. The evolution of copper-, nickel-, manganese- and silicon-enriched precipitates is studied by means of atom-probe tomography, small-angle neutron scattering, X-ray diffraction, and small-angle X-ray scattering. Evolution of these microstructural features is compared to fracture toughness degradation and hardening of these steels.

G-25: On the Effects of Helium-dpa Interactions on Microstructural Evolution in Tempered Martensitic Steels: Analyses of Dual Ion Beam Irradiation Databases: Takuya Yamamoto1; G. Robert Odette1; Yuan Wu1; Kiyohiro Yabuuchi2; Sosuke Kondo2; Akihiko Kimura2; 1University of California Santa Barbara; 2Kyoto University
    The UCSB cavity evolution database on dual beam Fe3+ + He+ ion irradiations (DII) was analyzed as a function of dpa, He/dpa ratio and dpa rate to establish void swelling trends in normalized and tempered martensitic steels. A wide range of DII dpa and He/dpa provides a basis for characterizing the systematic dependences of swelling on these variables. For example at a nominal 1.5x10-3 dpa/s, the swelling incubation dose, dpai, linearly decreases with increasing He/dpa from ≈ 80 dpa for < 5 appm/dpa to ≈ 20 dpa for ≈ 50 appm/dpa prior to post incubation swelling at a rate of ≈ 0.1 %/dpa. Lower dose rate irradiations at ≈ 7x10-4 dpa/s led to ≈ 4 times lower swelling with an ≈ 50% higher incubation dpai. Detailed analyses of cavity size distributions evolution combined with rate theory models are discussed in terms of the critical bubble to void concept.

G-26: In-situ Ion Irradiation Induced Microstructure Evolution in Ferritic/Martensitic Steel T91: Djamel Kaoumi1; Ce Zheng1; 1North Carolina State University
    In-situ transmission electron microscopy (TEM) combined with ion irradiation is a powerful tool to investigate the kinetics of irradiation induced effects because the microstructure evolution can be characterized as the damage proceeds. In this study, in-situ TEM characterizations were performed on ferritic/martensitic steel T91 under 1 MeV Kr++ ion irradiation at different irradiation conditions (50K, 298K, 673K, and 773K up to 10 dpa). The microstructure evolution under irradiation was followed in terms of formation and evolution of clusters and loops and the spatial correlation of the defect structures with the pre-existing microstructure. The size and number density of dislocation loops were measured and compared between different irradiation conditions. Self organization of defects into arrays aligned along certain crystallographic directions were observed under some of irradiation conditions. Efforts were made to establish the relationship between the microstructure evolution and some key parameters (e.g. irradiation temperature, grain size of T91).

G-27: In Situ TEM Cantilever Testing of Irradiated ODS to Determine Grain Boundary Embrittlement and Cohesion: Kayla Yano1; Janelle Wharry2; Xianming Bai3; 1Boise State University; 2Purdue University; 3Virginia Tech
    The objective of this study is to utilize miniaturized cantilever beam tests to link radiation induced segregation (RIS) with grain boundary cohesion. Atomic-level chemical changes associated with RIS affect grain boundary cohesion and subsequently alter mechanical and fracture properties of the bulk material. Here, we utilize a transmission electron microscopic (TEM) in situ cantilever test to connect the atomic and macro scales. Our work focuses on a model Fe-9%Cr oxide dispersion strengthened alloy, which is a candidate structural material for advanced reactors. The ODS is irradiated with Fe++ self-ions to 3 and 100 dpa at 500°C. We fabricate and test sub-micrometer-sized cantilevers, which are fully contained within the ion-irradiation layer. Each cantilever contains a grain boundary oriented transversely at the fixed end. We combine quantitative stress-strain data with RIS measurements, qualitative observation of grain decohesion, and complimentary atomistic modeling to better understand irradiation effects on grain boundary embrittlement and cohesion.

G-28: Microstructural Evaluation of Ion Irradiated Model Binary Alloys: Ling Wang1; 1University of Tennessee
    Phase stability of nanoscale precipitates is of key importance for the design of radiation resistant materials, yet relatively little fundamental information is available regarding the role of particle size, coherency and composition on dynamic stability during irradiation. In order to provide initial information on the ballistic stability of different types of precipitates, three copper-base binary alloys (1%Fe, 1%Co and 0.8%Cr), two nickel-base alloys (0.8%Zr, 0.8%Hf), and two iron base alloys (1%Cu, 1%Zr) were heat treated to produce precipitates of different sizes and densities, and then irradiated with 1 MeV Cu ions to peak doses of 0.1 and 1 dpa at room temperature. The microstructure was characterized using cross-section transmission electron microscopy. Information will be summarized regarding the evolution of the precipitate size and density as a function of irradiation dose and precipitate type, and the effect of precipitate sink strength on the evolution of defect clusters such as dislocation loops.

G-29: Neutron Irradiation and Post Irradiation Annealing Effects on the Microstructure of HT-UPS Austenitic Stainless Steel: Chi Xu1; Xuan Zhang2; Wei-Ying Chen2; Meimei Li2; Jun-Sang Park2; Jonathan Almer2; Yaqiao Wu3; Yong Yang4; 1Argonne National Laboratory / University of Florida; 2Argonne National Laboratory; 3Idaho National Laboratory / Boise State University; 4University of Florida
    High-Temperature Ultra-fine Precipitate Strengthened (HT-UPS) Stainless Steel samples were neutron irradiated to 3dpa at 500°C in the Advanced Test Reactor (ATR), Idaho National Laboratory. Post-irradiation annealing was conducted at 600°C and 700°C for 1 hour, respectively. The microstructures of three specimens, 3dpa/500°C, 3dpa/500°C + 600°C/1h, 3dpa/500°C + 700°C/1h were characterized using synchrotron wide-angle X-ray scattering (WAXS), small-angle X-ray scattering (SAXS), high-energy X-ray diffraction microscopy (HEDM), transmission electron microscopy (TEM), and atom probe tomography (APT). X-ray measurements showed significant lattice parameter changes and peak broadening after neutron irradiation and post-irradiation annealing treatments. TEM revealed Frank dislocation loops and black-dot defect clusters in the as-irradiated specimen, and the existence of nano-sized precipitates in the irradiated-annealed specimens, which were further confirmed by APT.

G-30: Numerical Estimation of Phosphorus Transport for Different Migration Modes in Alpha-iron: Ken-ichi Ebihara1; Tomoaki Suzudo1; Masatake Yamaguchi1; 1Japan Atomic Energy Agency
    We evaluated diffusion coefficients of phosphorous (P) in alpha-iron (Fe) of a mixed interstitial dumbbell (MID) composed of a Fe atom and a P atom using the kinetic Monte Carlo simulation based on the model which determined from first-principles calculation. The evaluated coefficient was compared with that of the interstitial P atom and that for the vacancy (V) migration mode. As a result, it was confirmed that the diffusion of interstitial P atoms and MIDs is almost the same with each other and is faster than that of the V mode. Furthermore, we modified the rate theory model by incorporating the interstitial P atom and the evaluated coefficients and applied it to the simulation of irradiation-induced grain-boundary (GB) P segregation. Then it was found that the previous model of GB is not adequate for the interstitial P atom because it does not consider the detrapping process of P atoms.

G-32: The Effect of Pre-implanted Helium on Cavity Nucleation and Swelling Rate in Ion-irradiated T91: Anthony Monterrosa1; Zhijie Jiao1; Gary Was1; 1University of Michigan
    To compensate for lack of transmutation reactions in ion irradiation of reactor structural materials, helium is often implanted into the samples prior to irradiation. The purpose of this study is to determine the role of pre-implanted helium on the nucleation and swelling rate of cavities after ion irradiation. Samples of ferritic-martensitic alloy T91 were pre-implanted with helium levels of 0, 1, 10, 100 and 1000 appm at room temperature, followed by irradiation with 5 MeV Fe2+ ions at 460°C up to damage levels of 450 dpa at the Michigan Ion Beam Laboratory. The effect of varying levels of pre-implanted helium on the nucleation and swelling rate of cavities was assessed. It was found that helium content above 10 appm dramatically increased nucleation and suppressed growth. The swelling rate exhibited a maximum at 10 appm of helium, where nucleation was enhanced and growth was unhindered.

G-33: The Evolution of Laves Phase in Ferritic-Martensitic Steel Grade 92 under Thermal Aging and Sodium Exposure: Wei-Ying Chen1; Meimei Li1; Krishnamurti Natesan1; 1Argonne National Laboratory
    The Ferritic-martensitic steel grade 92 is candidate structural material for use in sodium fast reactors because of its high temperature strength and creep resistance. This study investigated the effects of aging and sodium-exposure on the microstructure of G92 treated at 550°C, 600°C and 650°C up to 36061 hours. After aging, Laves phase, Fe2(W, Mo), precipitated at grain boundaries. The SEM measurements showed that the size of Laves phase increased with time and temperature. A gradual saturation of areal fraction with time had been observed for 600°C and 650°C, but not yet for 550°C. Sodium exposure resulted in a larger size and lower density of Laves phase at the exposure surface (depth ~40 µm). Correspondingly, an altered composition was observed with EDS locally at the surface. The effect of the Laves phase formation on solid solution strengthening from tungsten and molybdenum will be discussed.

G-34: TEM Observations on He Bubble Nano Oxide Associations in As-Processed and Annealed Nanostructured Ferritic Alloys: Yuan Wu1; Tiberiu Stan1; Takuya Yamamoto1; Jim Ciston2; G. Odette1; 1UCSB; 2NCEM at LBNL
    An important attribute of nanostructured ferritic alloys (NFAs) is their ability to manage high concentrations of transmutation product He in nm-scale bubbes associated with nano oxides (NOs). Here we report details of bubble-NO associations, using a variety of TEM techniques, for peak He concentrations of ≈2000 appm implanted at 700°C. To facilitate these studies, we examine both as-processed and post-processing annealed NFA conditions, where the latter contain coarsened, thus more readily characterized, oxides-bubble features. These observations show a one-to-one association between the oxides and bubbles. Other details, like bubble facet selection, attachment angles, and bubble-NO size correlations, are presented. The crystal structure of the oxides after annealing and implantation were also characterized. Oxides coarsened from about ≈ 2.5 to ≈ 7 nm remain pyrochlore Y2Ti2O7 with a cube-on-edge orientation matrix relationship. However, much larger 25 nm oxides no longer index as pyrochlore.

G-35: In Situ Studies of Nanopore Shrinkage during Heavy Ion Irradiation of Nanoporous Au: Jin Li1; Cuncai Fan1; Jie Ding1; Sichuang Xue1; Youxing Chen2; Qiang Li1; Haiyan Wang1; Xinghang Zhang3; 1Texas A&M University; 2Los Alamos National Laboratory; 3Purdue University
    Nanoporous (np) metallic materials with a large surface-to-volume ratio have unique mechanical and physical and catalytical properties. Recent studies show that np metals may have superior radiation tolerance compared to their bulk coarse-grained (cg) counterparts. However, there is very limited in situ studies on the radiation response of nt metals. Here we show, by using in situ Kr ion irradiation in a transmission electron microscope, the shrinkage of nanopores during radiation. Furthermore, the diffusivity under different dose-rate in np Au has been discussed. This study sheds light on the design of radiation-tolerant nanoporous metallic materials. This study is supported by NSF-DMR-Metallic Materials and Nanostructures Program under grant no. 1304101.

G-36: Irradiation Effects on Diffusivity of Copper in Ferromagnetic Iron Studied by Atom Probe Tomography: Takeshi Toyama1; Masaki Shimodaira1; Keiko Tomura1; Naoki Ebisawa1; Kazuaki Nagumo1; Yasuo Shimizu1; Koji Inoue1; Yasuyoshi Nagai1; 1Tohoku University
    Since Cu precipitation plays an important role in the hardening in nuclear reactor pressure vessel steels, the diffusion coefficient (D) and the solubility of Cu in Fe are essential to the understanding of precipitation kinetics. It is suggested that the diffusivity is enhanced by irradiation due to super-saturated vacancies and interstitials, however, only a few studies are reported about the D in irradiation-environment. In this study, diffusion-couples of Cu-Fe were electron-irradiated at 573 – 773 K to the dose of about 10 mdpa, and then the diffusivity of Cu in Fe was studied by atom probe tomography. The irradiation effects on D and the solubility limit will be presented.

G-38: Nickel Ion Irradiation Damage In GH3535 Alloy Weld Metal and the Temperature Effect: Hefei Huang1; Xiaoling Zhou1; Zhiyong Zhu1; 1Shanghai Institute of Applied Physics, Chinese Academy of Sciences
    GH3535 alloy weld metal have been characterized by TEM and nanoindentation to determine their microstructural evolution and mechanical property changes after 8MeV Ni+ ions irradiation. The irradiation experiments were carried out at room temperature and 600 oC, and the ion fluences correspond to a calculated peak damage dose of 0.5, 2 and 12 dpa, respectively. TEM results show the formation of nano-scaled dislocation loops with a number density of approximately 5-14×1022 m-3 at room temperature. At high temperature irradiation, several same-sized dislocation loops but lower in amount were observed. The calculated indentation values in irradiated samples were found to be much higher in comparison to the unirradiated one. However, in the case of the ion irradiation at 600 oC, the hardness value was significantly decreased. The relationship between ion irradiation induced microstructural evolution and the changes in the mechanical properties of this weld metal is discussed.

G-39: Radiation-induced Segregation in Proton Irradiated Commercial Fe-Cr-Ni Base Austenitic Alloys: Miao Song1; Chad Parish2; Gary Was1; 1University of Michigan; 2Oak Ridge National Laboratory
    Radiation-induced segregation (RIS) behavior of several commercial austenitic alloys was investigated. These alloys were divided into three groups by their Fe content: alloy 316L-with around 70 wt.% Fe, alloys 310 and 800 with 45-55wt.% Fe, and nickel base alloys 625, 690, and 725 with less than 10 wt.% Fe. All the alloys were proton-irradiated to 5 dpa at 360şC in the Michigan Ion Beam Laboratory (MIBL). Energy-dispersive X-ray spectroscopy (EDX) scans were performed on a Talos transmission electron microscopy (TEM). RIS results show that all the random high angle grain boundaries of these austenitic alloys follow the same pattern of significant Cr and Fe depletion, Ni enrichment. RIS of minor elements such as Si in austenitic steels were tracked as well. The segregation behavior will be interpreted in the context of major and minor element compositions and the inverse Kirkendall model for RIS.

G-40: Study of Neutron and Ion Irradiation Damage in Aluminum Alloys: Ziv Ungarish1; Benedicte Kapusta2; Pierre Gavoille2; 1NRCN; 2DEN-Service d’Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay
    Aluminum alloys are common materials in the nuclear industry, used in numerous different components in research reactor cores. During reactor operation components are exposed to a neutron flux generating radiation damage. This can affect the material's microstructure, resulting in degradation of properties. In the current study, Al 6063 and AlFeNi have been irradiated with both Si ions and neutrons. Si ion irradiation creates damage associated with Frenkel pair formation, while introducing silicon in the aluminum matrix. This mimics the formation of silicon through aluminum transmutation, due to the interaction with thermal neutrons. Si ion irradiations were performed at different sets of parameters. The microstructure of both alloys before and after irradiation was studied by TEM. The mechanical properties of the ion irradiated materials were characterized using nanoindentation. The effects of the ion and neutron irradiations on microstructure and mechanical properties will be presented, compared and discussed.

G-42: Ion Irradiation-induced Structural Damage in Different Multi-component Alloys at Elevated Temperatures: Tengfei Yang1; Songqin Xia2; Yuan Fang3; Yong Zhang2; Congyi Li1; Yugang Wang3; Steven Zinkle1; 1Department of Nuclear Engineering, The University of Tennessee; 2State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing; 3State Key Laboratory of Nuclear Physics and Technology, Center for Applied Physics and Technology, Peking University
    A series of four progressively more complex equiatomic alloys (Ni, FeNi, FeCoNi and FeCoCrMnNi) were irradiated by 3 MeV Cu to 2×1015 cm-2 (10 dpa at midrange) at 250~650 oC to study irradiation tolerances of different multi-component alloys. It is found that the four materials exhibit a typical irradiation response of fcc alloys; no phase decomposition was observed in TEM characterization. The three concentrated alloys show good resistance to void swelling. Numerous irradiation-induced voids can be observed in as-irradiated Ni, while three concentrated alloys are nearly void-free. Furthermore, the sizes of irradiation-induced defects are decreased with increasing number of principal components, and the growth of defect sizes with increasing irradiation temperature is much slower for the three concentrated alloys. Detailed high resolution mapping of radiation induced solute segregation at grain boundaries will be presented. The improved irradiation tolerance of multi-component alloys is attributed to the sluggish atomic diffusion.

G-43: Effect of Proton Irradiation on Deformation Mechanisms in Model Alloy Fe–20Cr–25Ni: Tianyi Chen1; Lizhen Tan1; Kumar Sridharan2; Haixuan Xu3; 1Oak Ridge National Laboratory; 2University of Wisconsin–Madison; 3The University of Tennessee
    Alloy 709 (Fe–20Cr–25Ni–1.5MoNbTiN) is down-selected as a candidate material for sodium-cooled fast reactors, for its superior corrosion and creep resistances. To obtain fundamental knowledge about the radiation effects on its mechanical properties, proton irradiations, mechanical tests and microstructure characterizations were conducted to model alloy Fe–20Cr–25Ni. For both unirradiated and proton-irradiated samples, nanoindentation was conducted on different grains whose crystal orientations were determined by electron backscatter diffraction. A systematic observation of the anisotropy in hardness and elastic modulus was made for both irradiated and unirradiated samples. Proton irradiation was found to affect the mechanical properties. Cross-section samples were lifted-out underneath the nanoindents using focused-ion beam. Transmission electron microscopy characterizations revealed the radiation effects on microstructure and deformation mechanisms, which was related to the crystallographic orientations. This study will provide essential insights into radiation resistance of Alloy 709 and a critical bridge between experiments and multiscale modeling.

G-44: Deformation of He Bubble Superlattice in FCC Cu: Ian Winter1; Daryl Chrzan1; 1University of Calfifornia, Berkeley
    Recent work on He-irradiated nano-twinned copper has shown that the formation of a helium bubble superlattice can increase the yield strength by approximately a factor of two. This finding motivates the current work, to give a greater understanding to the mechanisms governing this dramatic change in the mechanical response of the material after helium irradiation. Molecular dynamics simulations show that the presence of a helium bubble superlattice has a profound effect on Cu's mechanical properties both in terms of twin nucleation as well as twin propagation. By modeling the response of a step in a twin boundary to an applied shear stress, critical resolved shear stress values have been obtained that are in qualitative agreement with experiments. Evidence from this work suggests that the strengthening by helium is a result of precipitation hardening. Funding is provided by the Nuclear Energy University Program through the DOE, project No. 13-5161.

G-45: Simulations of Irradiated-enhanced Segregation and Phase Separation in Fe-Cu-Mn Alloys: Boyan Li1; Ben Xu2; Wei Liu2; Chuck Henager3; Shenyang Hu3; 1Tsinghua University, Pacific Northwest National Laboratory; 2Tsinghua University; 3Pacific Northwest National Laboratory
    For reactor pressure vessel steels, the addition of Cu, Mn, and Ni has a positive effect on mechanical, corrosion and radiation resistance properties. However, experiments show that irradiation-enhanced segregation and phase separation is one of important material property degradation processes. In this work, we will present a model which integrates the rate theory and phase-field approaches to predict radiation-enhanced segregation. The rate theory is used to describe the accumulation and clustering of radiation defects while the phase-field model is used to describe the effect of radiation defects on phase stability and microstructure evolution during phase transitions. The model allows the effect of temperatures and radiation damage rates on segregation and Cu rich phase nucleation to be systematically investigated. The predicted nucleus has a core-shell composition profile, i.e., Cu rich at center and Mn rich at the interface, which are in good agreement with the theoretical calculation and experimental observation.

G-46: A Study on Irradiation Induced Microstructure Dependent Thermal Conductivity Change of Zircaloy using Nanomechanical Raman Spectroscopy: Hao Wang1; Vikas Tomar1; 1Purdue University
    This work focuses on the study of microstructure dependent thermal property of irradiated Zircaloy-4 using a combination of nanomechanical Raman spectroscopy with phase field simulations. Phase field method is used to understand the influence of cavities, voids, and dislocations. Vacancy-dislocation interaction is modeled to account for the thermal property change due to microstructure evolution. The local thermal conductivity value at each grid point is determined based on the phase field order parameter, which represents the local microstructure feature. The effective conductivities of the whole domain are analyzed with respect to dislocation density change. Nanomechanical Raman spectroscopy is used to measure the local thermal conductivity distribution in analyzed microstructures. A specific heat transfer model is derived for samples with thin coating layer.

G-47: Oxide Texture as Cause and Effect in the Corrosion of Zirconium Fuel Cladding - An Atomistic Simulation Study: Maria Yankova1; Christopher Race1; 1Materials Performance Centre, University of Manchester
    The structural and electronic properties of oxide grain boundaries strongly affect the transport of species through the oxide layer. Hence, changes in oxide texture under irradiation can significantly alter the oxidation kinetics of zirconium alloy fuel cladding in light water reactors. We have performed density functional theory (DFT) calculations of the stiffness of key lattice planes in bulk tetragonal and monoclinic zirconia to explore the effect of transformational stresses on the growth of certain grain orientations strongly represented in experimental texture maps. We have performed benchmark DFT calculations of structural and electronic properties of representative grain boundaries and inter-phase boundaries in the oxide microstructure and how these are altered by the presence of irradiation-induced defects and alloying elements. We have further used the results of our calculations to test a range of empirical potentials for the Zr-O system to establish their suitability for computational modelling of zirconia at microstructural length-scales.

G-48: The Effect of Niobium on the Irradiation Induced Growth Properties of Zr-Nb Binary Alloys Used forNnuclear Applications: Rebecca Jones1; Elisabeth Francis1; Philipp Frankel1; Aidan Cole-Baker2; 1University of Manchester; 2Rolls Royce Plc.
    Zirconium alloys are currently the material choice for fuel cladding in nuclear reactors. Irradiation induced growth is a common problem for these components and has been correlated with the formation of dislocation structures. Commercial alloys containing niobium (Nb) provide improved growth performance. A mechanistic understanding of the effect Nb has on the microstructure is therefore required. The binary Zr-Nb system is studied where the Nb content in solution and as secondary phase is varied. Techniques such as high-resolution transmission electron microscopy (TEM) and synchrotron x-ray diffraction (SXRD) are used to characterise the alloys before and after proton irradiation, to monitor SPP formation/dissolution and to observe dislocation structure progression. A correlation between increased dislocation formation with respect to irradiation damage is noted. The work allows the study of evolution of SPPs and dislocation structures with respect to both Nb content and proton dose.

G-49: Ex-situ and In-situ Investigation of Heavy Ion Irradiation Damage in Ti-6Al-4V: Aida Amroussia1; Carl Boehlert1; Florent Durantel2; Clara Grygiel2; Wolfgang Mittig3; Isabelle Monnet2; Frederique Pellemoine4; 1Michigan State University; 2CIMAP CEA/CNRS/ENSICAEN/UCN; 3National Superconducting Cyclotron Laboratory- Michigan State University; 4Facility for Rare Isotope Beams-Michigan State University
    Due to its high specific strength, good mechanical properties and corrosion resistance Ti-6Al-4V is considered as a structural material for the beam dump drum for the Facility for Rare Isotope Beams. Ti-6Al-4V samples were irradiated at the CIMAP-GANIL Facility (1 MeV/u) and Notre Dame University (0.1 MeV/u) to investigate the changes in microstructure and mechanical properties at 350şC and 25 şC. Nano-indentation results for samples at lower does (~1 dpa), indicated a low sensitivity to high electronic excitation (~7.5 keV/nm) and that the radiation damage was affected mainly by a dual dose and temperature dependence. Dose rate effect was also investigated (13 dpa/h and 0.8 dpa/h). In addition, an in-situ TEM investigation of irradiation damage was performed at the IVEM-Tandem Facility at Argonne National Laboratory. Three different Ti-6Al-4V TEM samples, with different processed microstructures, were irradiated with 1 MeV Kr2+ at 350şC up to a dose of 24 dpa.

G-50: Quantification of Dislocation Densities in Zirconium Hydride by X-ray Line Profile Analysis: Miguel Vicente Alvarez1; Javier Santisteban1; Pablo Vizcaino2; Gábor Ribárik3; Tamás Ungár3; 1Centro Atómico Bariloche; 2Centro Atómico Ezeiza, Argentina; 3Eötvös University Budapest
    Zirconium-based components in nuclear power plants are embrittled by delta-ZrH precipitates accompanied by forming fcc Zr sublattice. The hydride has a complex distorted internal structure. Specimens were investigated with different H content: (i) components charged in the laboratory with H250 wt-ppm, (ii) laboratory-produced hydride blisters in Zr2.5%Nb pressure tubes, (iii) Zircaloy-4 from cooling channels of Atucha I nuclear power plant after 10 years' service with~140 wt ppm equivalent H content, exposed to~10^22 neutrons/cm^2. Diffraction patterns in a hydride blister were scanned parallel and perpendicular to the pressure tube surface in the 1-ID synchrotron beamline of APS, Argonne. Dislocation and stacking fault densities were determined by the CMWP procedure in delta-hydrides as 5-20x10^15 cm^-2 and up to 2 %, respectively. In the matrix the dislocation density was about an order of magnitude smaller. Dislocations densities are proportional to the volume fraction of hydrides accounting for matrix hardness in the precipitate structure.

G-51: Microstructural Effects on Helium Plasma-materials Interaction in Tungsten: Kun Wang1; Chad Parish1; Mark Bannister1; 1Oak Ridge National Laboratory, UT-Battelle
    Tungsten is the leading candidate material for tokamak PFM applications in future fusion reactors. When exposed to the harsh fusion environment, tungsten's microstructure will degrade with service, possibly to include recrystallization. We have exposed hot-worked and recrystallized tungsten to 80 eV helium ions at 900°C to fluences of 2 or 20×1023 He/m2. SEM and TEM were employed to investigate the evolution of surface morphology. The results indicated that an incipient nanofuzz structure in the case of low fluence or short but well-developed nanofuzz structure with high fluence helium ion exposure were observed. Evident difference of penetration depth of bubbles between hot-worked and recrystallized tungsten were found, indicating crystallographic defects might influence bubble nucleation and growth. Last, the bubble distribution in the unique Σ3 CSL grain boundaries in recrystallized W were compared to random boundaries to assess the effects of GB energy on helium bubble distribution.

G-52: Enhanced Radiation Tolerance and Thermal Fatigue Properties of Nanochannel W Films: Feng Ren1; Wenjing Qin1; 1Wuhan University
    Plasma facing components for fusion reactor need to face harsh environments. The accumulation of large gas bubbles is one of main reasons for the formation of fuzz on surface of W. In this work, we propose a new strategy to increase the radiation tolerance. Nanochannel W films with a high surface-to-volume ratio were exposed to 40 keV He ions irradiation and ELMs-like thermal shock loads conditions. TEM and SEM micrographs show that high density bubbles and large cracks appear in W bulk, while low density bubbles and no crack are found in the nanochannel W films because the rich free surfaces act as efficient sinks to trap and quickly release irradiation-induced defects along nanochannels to achieve “self-healing”. The absorbed power density related cracking threshold for the nanochannel film is located between 0.28 and 0.43 GW/m2 under 100 cycles ELMs-like thermal shock loads, which significantly higher than that of bulk W.

G-53: Impact of Low Dose ion Irradiation on Raman Spectra and Thermal Conductivity in Beta-SiC: Vinay Chauhan1; M Faisal Riyad1; Xinpeng Du1; Changdong Wei1; Beata Tyburska-Püschel2; Ji-Cheng Zhao1; Marat Khafizov1; 1The Ohio State University; 2University of Wisconsin
    Silicon carbide (SiC), a wide band gap semiconductor and hard ceramic having stable chemical and mechanical properties is important for both nuclear and electronics industry. The effect of low temperature and low dose irradiation induced defects on degradation of thermal and micro-structural properties of 3C-SiC were investigated. Raman spectroscopy was used to probe the vibrational modes and time domain thermo-reflectance (TDTR) was used to measure the thermal conductivity. It was observed that Raman intensity of longitudinal optical (A1 LO) peak decreased with irradiation dose in addition to its broadening and peak shift. In contrast to this, the transverse optical (E1 TO) peak did not exhibit any noticeable change caused by irradiation. A decreasing trend in peak ratio of A1 LO and E1 TO peaks was seen with respect to increasing irradiation dose in correlation with reduction in thermal conductivity is attributed to charged vacancy defects.

G-54: Microstructural Response of Si3N4 and AlN to Swift Heavy Ion Irradiation: Arno Janse van Vuuren1; Vladimir Skuratov2; Alexey Volkov3; Maxim Zdorovets4; 1Nelson Mandela Metropolitan University; 2Joint Institute for Nuclear Research; 3Nazarbayev University; 4National Nuclear Centre
    Si3N4 and AlN are not only important semiconductor materials but are also under consideration for use as candidate-inert-matrix-fuel-hosts (IM) for the burn-up of plutonium and minor actinides. The physical properties of these materials make them well suited to reactor conditions. However, in order to prove the viability of these materials for nuclear applications their radiation stability must be tested. In this investigation swift heavy ions are therefore used to simulate the effects of fission fragments on microstructure of these nitride ceramics. To this end these materials were irradiated with Xe and Bi ions, with energies ranging from 167 to 1030 MeV and temperatures ranging from LNT to 700 °C. In addition, these irradiation parameters allow for the determination of electronic energy deposition effects within the aforementioned materials.

G-56: Temperature and Se Dependence of Latent Track Morphology in TiO2 and Al2O3: Jacques O'Connell1; Vladimir Skuratov2; 1CHRTEM; 2JINR
    The study of latent track morphology in oxides irradiated with swift heavy ions is an important aspect in the ongoing quest to understand the mechanisms responsible for their creation. We have previously shown that the track morphology in pre-thinned specimens and in the near surface volume of bulk specimens vary greatly to the track morphology within bulk specimens. This observation is of great importance for the correct interpretation of indirectly obtained track parameters. We now continue our investigation into the latent track morphologies of these oxides by considering the effect of varying irradiation temperatures and electronic stopping powers on the structure of Xe, Ar and Kr induced latent tracks. Irradiation was performed at liquid nitrogen temperature, room temperature, 300 °C, 500 °C and 700 °C with ion energies of 220 MeV for Xe, 46 MeV for Ar and 40 - 107 MeV for Kr.

G-58: Ion Beam Induced Nanocrystal Formation with High Volume Fraction: Daryush Ila1; 1FSU
    Ion irradiation users have used ion implantation followed by thermal annealing in order to produce nanocrystals, while trying to control the size, distribution, and volume fraction of such nanocrystals. We avoided annealing and used ion radiation induced crystallization to produce nanocrystals, in order to have fine control on the size, distribution, and volume fraction of such nanocrystals. As of today, only two groups, Hubler et al and ILA et al have reached such high volume fraction of nanocrystals in order to infer the creation of pseudo-quantum dot lattices which have been shown in a series of research works initiated during the past decade. During this lecture we will review the results from past decades and present our most recent findings which resulted in production of thermally high insulating but electrically high conductive materials with high Seebeck coefficients produced by IBAD and produced by post irradiation by MeV ion beam.

G-59: Comparison of Microstructures in Neutron and Ion Irradiated Zr-1.0Nb-0.1Fe Cladding Alloys: Jing Hu1; 1Argonne National Laboratory
    A detailed study using a range of advanced microscopy techniques has been done on RXA Zr-1.0Nb-0.1Fe alloys under neutron and ion irradiation conditions. Neutron irradiated samples were received after 540 days and 5 dpa exposure in Vogtle reactor. In situ ion irradiation experiment on the autoclave oxidized sample has been carried out using Kr++ ion irradiation at elevated temperatures (320°C) up to 5 dpa at the Intermediate Voltage Electron Microscope-Tandem user facility at the Argonne National Laboratory. Neutron irradiation seems to have little effect on promoting fast oxidation or dissolution of â-Nb precipitates, but encourages dissolution of Fe from Zr-Fe-Nb precipitates. The dislocation loops are main defects produced during the irradiation. Size distribution, interaction with alloy elements and second phase particles, defect motility are compared between neutron and ion irradiated samples.