The stored energy and hardness of nanotwinned (NT) Cu are related to interaction between
dislocations and {111}-twin boundaries (TBs) studied at atomic scales by high-angle annular darkfield
scanning transmission electron microscope. Lack of mobile dislocations at coherent TBs
(CTBs) provides as-deposited NT Cu a rare combination of stability and hardness. The
introduction of numerous incoherent TBs (ITBs) reduces both the stability and hardness. While
storing more energy in their ITBs than in the CTBs, deformed NT Cu also exhibits high dislocation
density and TB mobility and therefore has increased the driving force for recovery, coarsening, and
recrystallization.

Recent developments within the nuclear materials community have lead current researchers to hypothesize that composite materials can address concerns regarding materials stability at higher temperatures and radiation environments. Advanced microscopy allows for direct insight into the physical mechanisms responsible for the improved performance of these materials. To extend our understanding of radiation tolerance and material stability, and to connect macroscale properties with potential sub-Ångstrom changes in electronic structure and electrostatics at interfaces. Tailoring interfaces considers the variation of structural and chemical properties. In the least, differences in crystal structure are selected to regulate changes in interface structure and the net role of ionic structure to either degrade or improve upon the material stability, in particular the susceptibility to amorphorize under irradiation. Studying the structure, chemistry, electrostatics, and response before, during, and after irradiation thereby leads to studies aimed at both understanding and controlling radiation tolerance.
In this talk, we examine the effects of interface structure on the damage evolution of structured oxide thin films, including varying interface roughness and film orientation. In detail, we have performed aberration corrected transmission electron microscopy on a series of irradiated CeO₂/SrTiO₃ (STO) and STO/MgO oxide interfaces to study changes in material stability in connection with interface structure. Based on the results, we determine that changes in step density layer are the structural pinning sites for the onset of amorphization for the CeO₂/STO interface. In composite STO/MgO thin films, we find that the amorphization susceptibility of STO is heavily influenced on the orientation relationship with the MgO substrate, which we attribute to long-range electrostatic effects. These results provide critical insights into the properties of oxide interfaces, radiation damage evolution, and how potential emerging nanomaterials can be tailored to be more radiation tolerant. They further highlight the complex behavior that can arise in these systems.
This work was supported by the Center for Materials at Irradiation and Mechanical Extremes (CMIME), an Energy Frontier Research Center (EFRC) funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number 2008LANL1026. JAA acknowledges access to the ORNL's ShaRE User Facility, which is sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, U. S. Department of Energy, where part of the TEM work was performed and the LeRoy Eyring Center for Solid State Science at Arizona State University. JAA acknowledges support from the postdoctoral program at Los Alamos National Laboratory.

The cascades of moving atoms that result from the impact of energetic ions on a solid are typically a few nanometers in length and width. When nanostructures with similar dimensions are ion-irradiated it is therefore to be expected that the comparability of length scales may give rise to effects not observed in bulk materials. Facilities such as MIAMI at the University of Huddersfield and the new facility at Sandia National Laboratories permit the ion-irradiation of specimens in situ within a transmission electron microscope enabling the study the effects of ion irradiation on nanostructures, at the level of individual ion impacts.
This paper will firstly report on recent work at Huddersfield and Sandia in which we have studied the effects of single ion impacts on monocrystalline Au nanorods (approximately 100 nm in length and 20 nm in width), using two different ions and energies, 80 keV Xe and 1.7 MeV Au. Many of the lower energy Xe ions deposit all of their energy and stop within the nanorods whereas the higher energy Au ions deposit only a fraction of their energy and the ions exit the nanorods. Morphology changes due to individual impacts, enhanced sputter yields and significant effects due to channeling will be presented and discussed and compared to results from molecular dynamics modeling.
From our work on Si nanostructures, we report on the amorphisation of polycrystalline Si nanowires by irradiation at room temperature with Ar and Xe ions in the energy range 40-80 keV. The nanowires have widths in the range 20-100 nm and lengths from 1-20 microns. A particular focus of the work is to compare both amorphisation and subsequent recrystallisation of the Si, during annealing, with behavior in bulk materials. In the first instance, the aim is to establish a phenomenological model of these processes in nanostructures.

Although radiation effects in bulk and thin film materials and carbon nanostructures have been studied very extensively, there are quite few theoretical studies of radiation effects in non-carbon nanowires and nanorods. Using molecular dynamics computer simulations, we have recently examined the effects of ion irradiation on Si, GaN and Au nanowires irradiated from the side (corresponding to a wire lying on a substrate) and from the top (standing on top of a surface). Both kinds of irradiation conditions have been realized experimentally. In this talk, I will overview our MD simulation results on primary damage production and sputtering in nanowires. The results show that although the basic energy dependence can be understood from a simple nuclear energy deposition picture, there are major finite-size effect enhancements in the damage production in nanowires compared to bulk. In particular, I will show that the damage production is strongly enhanced near the surface due to a combination of lowered threshold displacement energies and reduced recombination, both for lying and standing nanowires. I will also show that sputtering can be strongly enhanced in nanowires.

Nanotwinned metals have opened up exciting avenues for the design of high-strength, high-ductility materials owing to the extraordinary properties of twin boundaries. The recent advances in the fabrication of nanostructured materials with twin lamella on the order of a mere few atomic layers call for a closer examination of the stability of these structural motifs, especially at high temperatures. This paper presents a study of the entropic interaction between fluctuating twin boundaries by way of atomistic simulations and statistical mechanics based analysis. The simulations reveal that for all twin boundary spacings d, the interaction force between coherent twin boundaries varies as 1/d. For such crystalline interfaces that exhibit small fluctuations, this long range interaction is rather intriguing albeit consistent with the recent work by Freund [L. B. Freund, Entropic pressure between biomembranes in a periodic stack due to thermal fluctuations, PNAS 110, 2047 (2013)] on the steric repulsion between fluid membranes. However, the continuum analysis of the thermal fluctuations of twin boundaries shows that the statistical average of the entropic force between twin boundaries is zero and hence cannot be captured by molecular dynamics simulations. The 1/d dependence observed in simulations is then attributed to the inhomogeneity in the thermal expansion coefficient due to the interfacial regions. Thus, the interactions between crystalline interfaces such as twin and low angle grain boundaries, whose thermal fluctuations are dominated by the elastic fields in the surrounding medium, cannot be captured purely by membrane theory. These observations warrant further investigation through computations and possible experiments.

Ion beam mixing offers the unique opportunity to induce disorder at the atomic scale in materials leading to steady states associated with the appearance of new phases unexpected by the classical phase diagram¹. Even if the Molecular Dynamics technique is extensively used to predict the appearance of complex and simple defects at least in metals and alloys, this technique is limited to short times (few picoseconds) and avoids a correct description of the steady states associated with diffusion time scales (few micro seconds)². To overcome this difficulty, we propose in this work to combine results from MD simulation at the atomic scale with the Phase Fields method describing the microstructure over few tens of nanometers at at diffusion time³. This multi scale modelling has been applied on AgCu compounds⁴, a textbook example of alloys, to predict the steady states induced by irradiation. From the analysis of simulations, it appears that irradiation is able to stop the usual coarsening of Ag rich domains associated with a spinodal decomposition out of irradiation. Simulations clearly shows that the size distribution of precipitates as weell as their spatial distribution can be tuned by the choice of the nature and the energy of incident beam.
References:
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[2] H Emmerich, H Lowen,et al., Advances in Physics, 92, 1-95 (2012)
[3] D. Simeone, G. Demange, L. Lunéville, Phys. Rev. E 88 (2013) 3.
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We discuss the modifications to the Young-Laplace capillarity equation needed to describe nanoscale gas bubbles embedded in metals, scale at which the finite width of the interface region cannot be neglected. We focus in particular on the case of He in Fe. Using both, the concept of Tolman's length that provides a curvature dependence for the interface energy, and a new equation of state for He at the nanoscale that accounts for interface effects (see A. Caro et al. Appl. Phys. Lett. 103, 213115 (2013), we derive expressions to predict pressure, density and the amount of He in nanoscale bubbles. We find that conditions for equilibrium are found for values of pressure or density at variance by a factor of ~2 with respect to the traditional approach found in the literature for this problem (i. e. use of capillarity equation and bulk He EOS).

The interaction of ions with solids results in energy loss to both atomic nuclei and electrons in the solid. At low energies, nuclear energy loss dominates, leading to damage production via ballistic collision processes. At high energies typical of fission products and swift heavy ions, electronic energy loss dominates, leading to intense local ionization that can cause damage production [1], track formation [2] or damage recovery [3]. At intermediate ion energies, nuclear and electronic energy losses are of similar magnitude and can lead to simple additive effects on damage production [1,4], competitive damage recovery processes [4-6], and even synergistic effects that produce more damage than the sum of separate effects from nuclear and electronic energy loss. This intermediate ion energy regime is relevant to energies of primary knock-on atoms created by fission and fusion neutrons, energies of ions used to mimic fast neutron damage processes in materials, and ion energies used to modify materials by producing novel defects, nanostructures and phase transformations in order to tailor chemical and physical properties or create unique functionalities.
We have integrated experimental and computational approaches to investigate the separate and combined effects of nuclear and electronic energy loss on the response of ceramics to ion irradiation over a range of energies. Experimentally, ion mass and energy are used to control the ratio of electronic to nuclear energy loss; whereas, large scale molecular dynamics simulations, which include ballistic collision processes and the inelastic thermal spike from ionization, are used to model these effects. Using these approaches, additive effects of nuclear and electronic energy loss on damage production have been demonstrated in crystalline LiNbO₃ and grain growth in CeO₂ and ZrO₂. Similarly, the competitive effects of nuclear and electronic energy loss are confirmed in studies on ionization-induced damage recovery in SiC. The efficiency of such damage recovery processes decreases with decreasing electronic energy loss, and a threshold for defect recovery in SiC has been determined. Moreover, a highly synergistic effect of electronic and nuclear energy loss has been identified in SrTiO₃ and confirmed with molecular dynamics. These results have significant implications for ion beam modification of materials, non-thermal recovery of ion implantation damage, and the response of materials to extreme radiation environments.
This work was supported by the U.S. Department of Energy, Office of Basic Energy Sciences, Materials Sciences and Engineering Division.
References
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Swift heavy ion (SHI) irradiation-induced phase transformations are critical to understanding the response of materials in a wide range of radiation environments. However, there are few facilities in the world that provide the necessary energies for such ion beam irradiations. We have investigated the use of ultrafast laser irradiation as an efficient, low cost surrogate to SHI irradiation. A proof of concept experiment was performed where cubic Gd₂O₃ and monoclinic ZrO₂ were irradiated by a Ti:Sapphire laser with a pulse duration of 150 fs and lambda;=780 nm. The radiation-induced phase transformations were characterized by GIXRD and Raman spectroscopy. Results for Gd₂O₃ showed a cubic-to-monoclinic phase change as evidenced by the emergence of the (40macr;2), (003), (310), and (11macr;2) peaks in the GIXRD data and of the 482 cm^-1 peak and a broad peak around 260 cm^-1 in the Raman data. Results for ZrO₂ showed a monoclinic-to-tetragonal phase change as evidenced by the emergence of the (101) peak in the GIXRD data. These crystalline-to-crystalline phase transformations are consistent with those caused by SHI irradiation. These results will be discussed in terms of their similar damage mechanisms including thermal spike, bond weakening, and shock wave generation.

The continuing renaissance in nuclear materials development highlights the uncertainty in the use of ceramics as high burn up fuel matrices, or as first wall barrier materials within a fusion core. The key effects in next generation fission reactors are radiation damage, gas bubble formation and transmutation. These factors are difficult to predict at high burn ups, therefore model systems are studied for damage resistance and the effects of He bubble formation, predominantly to test existing models of damage accumulation and recovery. As part of a larger ongoing project examining such effects, results are presented for a selection of model ceramics, with both long and short range ordering, and the effects of radiation damage and gas bubble formation, for example delta phase oxides, perovskites, and novel garnet compositions. Radiation damage was induced in these materials by energetic ion implantation. Crystallographic structure and defect morphologies were examined by Grazing Incidence X-ray diffraction and cross-sectional Transmission Electron Microscopy, respectively.

The development of techniques like spin and dip coating, physical or chemical vapor deposition during the past decades offers now the opportunity for tailoring materials with numerous interfaces like multi layers or mesoporous materials. Glancing incidence X ray diffraction as well as the X ray reflectometry now offer the opportunity to probe the microstructure and the structure of materials at different depth, varying from few nanometers to few microns[1]. In this work, we firstly discuss the opportunity to use these techniques to follow the evolution of the microstructure and the structure as a function of the damage profile for new advanced nuclear materials. In the second part of this talk, we will point out the interest of the X ray reflectometry for studying ion beam mixing and then the stability of multi layers under irradiation. We lastly point out the interest of diffraction techniques to validate large scale simulations predicting the microstructural evolution of solids under irradiation [3].
References:
[1] D. Simeone, G. Baldinozzi, D. Gosset, G. Zalczer, and J. F. Berar. J. Appl. Crystallography, 44 :1205-1210, 2011
[2] G. Baldinozzi, G. Muller, C. Laberty-Robert, D. Gosset, D. Simeone, and C. Sanchez. J. Phys. Chem.C, 116(14) :7658-7663, 2012.
[3] D. Simeone, G. Demange, L. Luneville, Phys Rev E 88(3), 032116, 2013.

The goal of employing silicon carbide in fusion blanket designs requires a basic understanding of defect formation and evolution. We demonstrate that the most commonly used interatomic potentials are inconsistent with ab initio calculations of defect energetics. Both the Tersoff potential used in this work and an alternate modified embedded atom method potential reveal a barrier to recombination which is much higher than the DFT results. The barrier obtained with a newer potential by Gao and Weber is closer to the DFT result but the overall energy landscape is significantly different. This difference results in a significant difference in the cascade production of points defects. We have completed both 10 keV and 50 keV pka energy cascade simulations in SiC systems containing 27 M atoms at temperatures equal to 300, 600, 900 and 1200 K. Results were obtained for the number of stable point defects using the Tersoff/ZBL and Gao-Weber/ZBL (GW) interatomic potentials. In a contrast to Tersoff potential, the GW potential produces almost twice as many C vacancies and interstitials at the time of maximum disorder (~0.2 ps) but only about 25% more stable defects at the end of the simulation. Only about 20% of the carbon defects produced with the Tersoff potential recombine during the in-cascade annealing phase, while about 60% recombine with the GW potential. The GW potential appears to give a more realistic description of cascade dynamics in SiC, but still has some shortcomings when the defect migration barriers are compared to the ab initio results.
This work was supported by the US Department of Energy Office of Fusion Energy Sciences. Additional computational resources have been used for this work through collaboration with JAEA.

Silicon carbide (SiC) is a useful material for widely varying applications. This contribution will mainly focus on single crystalline electronic applications, but the material is also useful in poly crystalline and amorphous forms, for instance, as construction material in nuclear applications. These two application areas both utilize the radiation hardness and the high temperature properties of SiC. The aim here is to understand the damage processes in highly damaged material starting from isolated point defects in crystalline SiC.
Defect formation during the early stage of damage build-up in electronic materials is conveniently studied by electronic and optical spectroscopies, such as deep level transient spectroscopy (DLTS) and photo luminescence (PL). Only a handful of the intrinsic point defects in SiC has been unambiguously identified. One important defect in 4H-SiC was recently categorized as a carbon vacancy, V_C, with two electronic levels corresponding to different charge states in the upper half of the bandgap. This defect has a very strong influence on the charge carrier lifetime and the control of this defect poses a major challenge to the realization of high voltage bipolar SiC devices. For higher radiation fluence isolated point defects tend to form clusters and more extended defects, which generate stress and dislocations and eventually the material is no longer crystalline. The crystalline to amorphous transition can be studied by, for instance, Rutherford backscattering spectroscopy (RBS) and this presentation will attempt to bridge the “gap” between the low and high fluence damage regions.

Introduction of nanostructures is one of the possible ways to enhance radiation tolerance of materials. We have recently found that SiC with nanolayers of planar defects, typically stacking faults and twins, shows that complete amorphization is suppressed until a radiation dose of ~3 dpa (displacement of atom) at room temperature [1]. This value is much larger than that for amorphization of single crystalline SiC (ge;0.3 dpa) under similar irradiation conditions. This is attributed to the large annihilation rate of radiation defects in the nanostructured SiC. The defect annihilation depends strongly on irradiation temperature. In the present study, we have examined radiation tolerance of nanostructured SiC under cryogenic temperature. A nanostructured SiC thin film was deposited on a Si substrate. The specimens were irradiated at 100 K with 2 MeV Si ions to fluences ranging from 5x10¹⁵ to 1x10¹⁶ cm^-2, and they were characterized by means of transmission electron microscopy (TEM). TEM observations revealed that amorphous region increases with ion fluence, but crystallites remain at the surface up to the fluence of 6x10¹⁵ cm^-2. In combination with SRIM calculations, it was found that the amorphization is suppressed up to ~0.6 dpa which is much larger than 0.2 dpa for amorphization of bulk SiC at cryogenic temperature. We also examine the structural changes of nanostructured SiC under electron-beam irradiation [2]. [1] Y. Zhang, M. Ishimaru, T. Varga, T. Oda, C. Hardiman, H. Xue, Y. Katoh, S. Shannon, and W. J. Weber, Phys. Chem. Chem. Phys. 14, 13429 (2012). [2] M. Ishimaru, Y. Zhang, W. J. Weber, and S. Shannon, Appl. Phys. Lett. 103, 033104 (2013).

Graphite is currently pursued as a neutron moderator for the next generation, high temperature reactors. It has a lamellar structure, where the basal planes - constituted by strong, sp²-hybridized carbon atoms arranged in a hexagonal lattice - are held together through weak dispersion forces. The mechanisms of defect formation in graphite during the early stages of irradiation are still not completely resolved.
We report results from a molecular dynamics investigation on early radiation damage mechanisms in graphite using AIREBO interatomic potential that has been benchmarked to several defect and thermo-mechanical properties. Using non-equilibrium radiation cascade simulations, we show that topological defects, which maintain the sp² connectivity, are generated on the basal planes in addition to three dimensional sp³ defects. Interestingly, a primary knock-on atom that is parallel to the basal plane generates most number of defects. Our analysis also shows a logarithmic relationship between the temporal displacement of the primary knock-on atom and its energy indicating the formation of a dynamical cage facilitated by the layered structure in graphite. We apply our results for interpreting the defect or D peak in Raman spectra of irradiated graphite.

Progress in fusion and advanced fission power generation technology depends critically on the development of new high temperature materials. In metal alloys, including tungsten alloys, microstructural evolution occurring under irradiation is strongly dependent on the diffusion properties of point defects, such as interstitial atoms and vacancies, and also on the properties of extended defects, such as dislocations and surfaces. The mechanism of voids formation implies firstly the formation of smaller clusters as di-vacancy. Di-vacancy in tungsten, as in all BCC metals of the group VI B, has an unusual energy landscape. In tungsten, the first nearest neighbour is slightly repulsive or attractive depending of DFT calculations or exchange-correlation functional while the second nearest neighbour configuration is strongly repulsive [1]. The same tendency is observed for all elements of VI B group and is not the case for V B metals and iron for which the most stable configuration of di-vacancy is the second nearest neighbour configuration. However, in experiments, in the high temperature limit vacancy clusters are directly observed in tungsten [2].
The goal of this work is to shed light on the temperature impact over the energy landscape of point defects of tungsten based alloys under irradiation. Moreover, this work provides input data for kinetic models for microstructural evolution. The formation/migration free energies of mono- and di-vacancies / interstitials are computed using Density Functional Theory (DFT) calculations. The vibrational part of the free energy is computed in the frame transition state theory (TST) using harmonic approximation. The anharmonic part is evaluated by the recently developed method based on adiabatic reweighting algorithm for computing the free energy along an external parameter from adaptive molecular dynamics simulations [3]. The electronic entropy contribution to the free energy is also taken into account. The defect binding free energies and defect diffusion coefficients deduced from these calculations can be used to perform simulations of isochronal resistivity recovery experiments. We show that the temperature deeply impact the energy landscape of di-vacancy at higher temperature and can even change the relative stability of various configurations.
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Alloying allows the properties of a material to be tuned to a specific application, and advanced materials for a wide variety of applications (e.g., structural metals and ceramics, optoelectronic semiconductors, and functional oxides) are typically alloys. Radiation-induced point defects can modify and degrade material properties. Defect diffusion controls the rate of defect reactions and annealing, and thus quantitative models of defect diffusion are an essential component when predicting the time-dependent properties of a material exposed to radiation. In an alloy, defect energies and characteristics are sensitive to the occupations of nearby atomic sites and thus vary with location in the alloy. A defect can become trapped in energetically favorable regions of the alloy with profound consequence for the rate of defect diffusion. We have recently developed a model of defect diffusion in alloys that combines Kohn-Sham Density Functional Theory calculations, the Cluster Expansion approach, and kinetic Monte-Carlo simulations. We will describe this model and demonstrate its application to diffusion of intrinsic point defects in the compound semiconductor alloy InGaAs.
Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under Contract DE-AC04-94AL85000.

The irradiation behavior of a Ni-Mo-Cr-Fe alloy, of the type currently being considered for use in future molten salt cooled reactors, has been investigated in situ using 1 MeV Kr ions at temperatures of 723 and 973 K. When irradiated to 5 dpa, experimental observations reveal the instantaneous formation and annihilation of point defect clusters, with such processes attributed to the long range elastic interactions that occur between defects through multiple intra-cascade overlap. Corresponding differences in the defect cluster density and size distribution suggest that changes to the microstructure were dependent upon temperature and dose, affecting the growth, accumulation and mobility of irradiation-induced defect clusters under these conditions.

We describe new features and procedures at the IVEM-Tandem national user facility at Argonne National Laboratory, recently reopened to 100% operating time with full support from DOE-NE. To illustrate one of the unique applications of this TEM with in situ ion irradiation, we summarize recent experiments to measure interstitial defect formation and clustering in thin film molybdenum at 80 C during 1 MeV Kr ion irradiation. Using this in situ ion irradiation data we describe the construction of a model for interstitial cluster formation and dynamics that can then be applied to predict results for neutron irradiation of bulk Mo. The latter model prediction is compared to experimental results for neutron irradiation of the same pure Mo material at similar temperature. Careful attention is given to the physics of both ion and neutron damage mechanisms. The loss of mobile interstitial defects to free surfaces is uniquely measured in 3D by electron diffraction tomography and well modeled. Future directions for this and similar experiments using this unique user facility will be suggested.

Uranium dioxide, the standard nuclear fuel in pressurized water reactors, motivates intensive research to get further insight into the link between radiation damage and microstructure evolution of the material. Cerium dioxide is often considered as a non-radioactive model material for uranium dioxide, for which the experimental study of radiation damage can be performed more easily. Using first-principles calculations based on the density functional theory (DFT) and its DFT+U variant, we compare these two oxides in terms of point defect formation and fission gas location and migration.
First, after recalling the calculated bulk properties of both UO₂ and CeO₂ crystals, the charge states of point defects are discussed as a function of the stoichiometry of the materials, as well as the propensity of the single point defects to form aggregates. The comparison of the nature of defects in UO₂ and CeO₂ contributes to validate the use of CeO₂ as a model material for the study of radiation damage in UO₂.
Second, the interplay between point defects and fission gas behavior is studied. The most favorable sites for Kr and Xe atom incorporation in vacancies or vacancy clusters in UO₂ and CeO₂ are determined and some elementary migration mechanisms for Kr are proposed.

The conventional, crystallographic description of radiation damage posits a sequence of events based on the assumption that the radiation creates point defects. The vast majority of these self anneal by recombination, but the ones that persist act independently to diminish first the long range and ultimately the mid-range order, transforming the original crystalline material to an amorphous one. Local structure measurements provide a different description that invokes more of a "chemistry" type of approach instead of a mean field approach in which the different elements and sites act in a very similar way derived based on their averaged characteristics. The response of the various sites to the radiation field is therefore determined by their chemical speciation that is largely retained within the material. The first consequence is that, in a chemically heterogeneous compound, the energy from the radiation is at least somewhat mobile and will first attack the weaker bonds, giving radiation effects some specificity in chemically heterogeneous systems. Corollary to this effect is that internal stress may be relaxed by changes in local bonding so that radiation-induced defects may be stabilized and retained if they can interact with the material in this way. At the extreme, one could envision novel types of non-equilibrium structures within materials that are based on the incorporation of defects. We will show examples of these types of scenarios that have been identified by pair distribution function and XAFS measurements in various materials and especially in delta stabilized Pu-Ga alloys, whose initial stress from the presence of the smaller Ga atoms and internal radiation produce several different types of unique ordered structures in domains around the diffraction limit in size. We will also show how shock waves in Zr and Cu(Pb) give analogous structures, presumably with similar origins albeit on much shorter time scales. Retained shock damage, however, appears to be more oriented towards rotations of individual sites rather than stucural transformations and may be preferentially located near interfaces because of the interactions of relfected shock waves.

Uranium dioxide is currently the most widely used fuel material in fission reactors. During reactor operation, the fission of uranium atoms causes among others the formation of numerous defects, such as vacancies and vacancy clusters, which induce a significant evolution of the fuel physical properties. The vacancies can trap insoluble fission products, in particular fission gases and it is of great importance to understand their role in the early stages of the formation of gas bubbles in UO₂.
One of the experimental techniques permitting the investigation of vacancy-type defects is positron annihilation lifetime spectroscopy (PALS), which consists in recording the radiation emitted at the beginning and end of life of positrons in a material and deducing the positron lifetime and the properties of the electrons with which they have annihilated. Vacancies can trap positrons and can therefore be detected by monitoring e.g. changes in the lifetime of positrons in the material. To identify the types of defects present in the examined materials, however, comparison with calculated positron lifetimes or with results of other characterization techniques is required.
We present positron annihilation lifetime spectroscopy measurements on uranium dioxide irradiated with 45 MeV α particles. These samples have been already studied by PALS at room temperature in a previous study [1]. In the present study we measured the positron lifetime as a function of the temperature in the 15-600 K range. The experimental results were combined with electronic structure calculations of the most stable charge states [2,3] and positron lifetimes of vacancies and vacancy clusters in UO₂. Neutral and charged defects consisting of from one to six vacancies were studied computationally using the DFT+U method to take into account strong correlations between the 5f electrons of uranium. The two-component density functional theory with fully self-consistent schemes was used to calculate positron lifetimes. All defects were relaxed taking into account the forces due to the creation of defects and the positron localized in the vacancy. The interpretation of the experimental observations in the light of the DFT+U results indicates that neutral V_U+2V_O trivacancies (bound Schottky defects) are the predominant vacancy type defects detected in the 45 MeV α irradiated UO₂ samples.
[1] M.-F. Barthe, H. Labrim, A.Gentils et al., Phys. Status Solidi C 4, 3627 (2007).
[2] J. Wiktor, E. Vathonne, M. Freyss et al., in MRS Proceedings, Vol. 1645, pp. Mrsf13-1645, Cambridge Univ Press (2014).
[3] E. Vathonne, J. Wiktor, M. Freyss et al., in press, J. Phys.: Condens. Matter (2014).

Minor actinides like americium or curium are produced during the irradiation of the most commonly used nuclear fuels, namely uranium dioxide or the mixed uranium-plutonium oxide (MOX), plutonium itself being also produced by neutron irradiation of uranium.
In open nuclear fuel cycles, characterized by a period of surface storage followed by disposal of spent fuel in a geological repository facility, plutonium and the minor actinides, together with some long lived fission products, will be responsible for the long term radiotoxicity of spent nuclear fuel, and for the medium term heat loading of the fuel. These transuranic elements are mainly alpha-emitters and will generate increasing amounts of alpha-damage and helium during spent fuel storage and after disposal.
In closed fuel cycles, the plutonium and minor actinides separated during spent fuel reprocessing are intended to be incorporated in new fuel for transmutation in fast neutron reactors as MABB (Minor Actinide Bearing Blanket), with the aim to reduce the inventory of radiotoxic elements and the heat loading of the final high level nuclear waste forms.
The decay heat produced by some of the short-lived actinides has been used in radio-isotopic thermal generators (RTG) to power space probes traveling across the solar system.
Alpha-decay generates defects mostly through elastic energy losses from the recoil of daughter nuclei. In addition, when the alpha-particle comes to rest it generates a helium atom that can alter the microstructure of the material, e.g. by forming microscopic bubbles.
All actinide dioxides have the fluorite structure, which is known as being a radiation damage resistant material. However, high alpha-dose can have detrimental effects on the long term stability of materials envisaged as fuels, MABB or within RTGs.
In this paper we report on experimental observations of alpha-damage effects in ²³⁸PuO₂, ²⁴¹AmO₂, (U, ²⁴¹Am)O₂, (U, ²³⁸Pu)O₂ by transmission electron microscopy (TEM), X-ray diffraction (XRD), thermal helium desorption spectrometry (THDS).
An attempt is made to identify the compounds most stable against damage build-up and to determine the specific features responsible for the observed differences.

The migration of uranium in uranium dioxide is an important factor in determining the mobility and release of fission gases from fuel pellets, as well as being a key parameter in the recovery from radiation damage. The effect of cluster configuration on vacancy mediated uranium migration is investigated for stoichiometric UO₂ and hyper-stoichiometric UO_2+x. It has been shown that the minimum enthalpy pathway involves reconfiguration of the most stable charge neutral cluster to a metastable configuration, in stoichiometric and hyper-stoichiometric systems. The reduction in migration enthalpy due to cluster rearrangement was 0.19 eV in UO₂ and 0.64 eV in UO_2+x. Additionally, a much larger number of alternative metastable configurations were identified for UO_2+x compared to UO₂. As such, a large contribution to the Arrhenius pre-exponential term with hyper-stoichiometry is expected. This will be an important mechanism for migration in other systems where defect clustering is significant and this work underpins the approach that will be taken for further studies.

Advanced nuclear fuel cycles offer considerable promise for improvements in safety, performance, reduced radiotoxicity of waste materials, and provide opportunities for associated methods of energy production (e.g., hydrogen based systems). In general, the proposed reactor systems require new materials capable of performing under the extreme conditions imposed by temperature, radiation fields, and corrosive media. Here we present a summary of results obtained via atomistic simulations and laboratory observations using ion irradiation and other methods. Atomistic simulations are conducted using density functional theory and molecular dynamics and are employed to 1) investigate the fundamental aspects of defect formation and migration on picosecond time scales, 2) predict the phase stability within chemical systems, and 3) examine aspects of the thermal behavior (primarily in the area of fuel performance). Most of the experimental radiation damage studies discussed here have been conducted using either in situ irradiation of thin TEM samples or irradiation of centimeter-size bulk ceramic samples. The results generally set out the groups of potential actinide (and other radionuclide) host phases in terms of those with intrinsic radiation tolerance due to recovery of damage on picosecond time scales (e.g., fluorite), those with favorable kinetics for longer term damage recovery (e.g., monazite), and many others with unfavorable kinetics. Together with information from the literature, these data illustrate some of the recent directions being taken in order to understand radiation effects in nuclear materials.

A complex Ce bearing oxide, Gd₂Ce₂O₇ was synthesized in order to simulate Pu in a fluorite derivative oxide. X-ray diffraction (XRD) was collected using a lab diffractometer at room temperature and Rietveld refinements of the XRD pattern was carried out using GSAS and XND programs. The diffraction pattern could be indexed with a C-type cubic bixbyite crystal structure with a cell parameter of 10.86413(8) Å. Several peaks showed peak asymmetry and line broadening and could not be fitted with conventional Rietveld refinement procedures. A full pattern refinement, assuming a possible formation of micro domains of a disordered phase with short range ordering and a structure correlated with the host C-type cubic lattice, gave a good fit with the experimental data. Understanding the structure of host lattice and its relation with the micro domains can have a profound effect on the radiation stability of this material. The diffraction data analysis and computational results will be presented and correlated with the complementary data obtained from transmission electron microscopy and Raman spectroscopy analysis.

Complex oxides, that is, oxides with multiple cation constituents, often form crystal structures based on simple layered atomic stacking arrangements, but with special atomic patterns within these layers. These atom patterns can be ordered or disordered, depending on stoichiometry or alternatively, depending on physical effects such as thermal or radiation-induced entropic disorder.
In this presentation, we will consider a layered structure model to describe atomic arrangements in a variety of oxides, from corundum to spinel to fluorite to pyrochlore. We will examine how atomic order varies as a function of compound stoichiometry and how atomic disorder is accommodated within these structures. We will also relate these layered atom arrangements to the radiation damage response of certain model oxide compounds. In particular, we will consider radiation-induced point defect accumulation in spinel and fluorite derivative compounds.

Effective modelling of dislocations demands large supercell molecular dynamics - it is therefore not surprising that this endeavour is a relatively new destination for simulators of nuclear ceramics. We will review recent studies that consider the relative stability (line energies) of different dislocations (edge and screw), what the models tell us about core sizes and elastic regions (where the elastic models start to work) and enhanced oxygen ion transport along dislocation cores (and the influence of compressive v&’s tensile regions). There are results too for single Xe atom migration along dislocations and the formation of clusters which are immobile and that pin the dislocations. We will also consider the models - which potentials are useful. To date there have only been a few studies of dislocations in UO₂ and none in non-stoichiometric materials or MOX. The interaction of dislocations with surfaces and grain boundaries also remains unexplored. Those studies will come and we will gradually gain a better understanding of the place that dislocations have in controlling the engineering performance of nuclear fuels.

Noble metal particles in the Mo-Pd-Rh-Ru-Tc system have been simulated on the atomic scale using density functional theory techniques for the first time. These noble metal fission product alloys form in nuclear fuels and by understanding their behaviour, improvements to performance can be made which will improve fuel efficiency and safety. Configurational entropy effects were considered to predict the stability of the alloys with increasing temperature in a similar manner to recent work on high entropy alloys. The variation of Mo content was modelled to understand the change in alloy structure and behaviour with fuel burn-up (Mo content decreases as burn-up increases in fast reactor fuels). The predicted structures compare extremely well with experimentally ascertained values where an increase in Mo content was found to increase the lattice constants and also increase the c/a ratio of the hexagonal close packed structure. Vacancy formation energies were all predicted to be positive over the composition ranges considered; a good indication of the alloys&’ structural stability, while the majority of fission products considered were found to be insoluble, as expected, iodine and tin were calculated to have negtive accommodation and solution enthalpies into selected noble metal compositions.

As the burn-up of a nuclear fuel is increased there is a marked increase in both the dislocation density and the concentration of fission products in the matrix. While it has been shown that fission products may segregate to the dislocations [1,2] there is currently very little understanding of what happens to the fission products once they reach the dislocation core. Shea has proposed that dislocations may provide a pathway for enhanced diffusion of gaseous fission products out of the fuel grain, which may explain the enhanced gas release observed during transients at high temperatures[3].
Here we perform molecular dynamics simulations, employing empirical pair potentials to investigate the interactions between individual fission gas atoms and bubbles with dislocations. We explore both the influence of the dislocation on fission products by examining the diffusion of Xe in the core region of the dislocation as well as studying how the presence of the gas atoms/bubbles can affect dislocation motion. The simulations show that there is enhanced diffusion of Xe in the vicinity of the dislocation, however, this leads to the formation of small nanobubbles which become immobile thereby reducing the influence of the dislocations on Xe mobility. We also observe a significant increase in the critically resolved shear stress as the concentration of Xe around the dislocations is increased.
[1] P. V. Nerikar, D. C. Parfitt, L. A. Casillas Trujillo, D. A. Andersson, C. Unal, S. B. Sinnott, R. W. Grimes and C. R. Stanek, “Segregation of xenon to dislocations and grain boundaries in uranium dioxide,” Phys. Rev. B, 84, 174105 (2011).
[2] A. Guyal, T. Rudzik, B. Deng, M. Hong, A. Chernatynskiy, S. B. Sinnott and S. Phillpot, “Segregation of ruthenium to edge dislocations in uranium dioxide,” J. Nucl. Mater., 441, 96 (2012).
[2] J. Shea, “An extension to fission gas release modeling at high temperatures,” EHPG Storefjell (2013).

We discuss the importance of equilibrium oxygen partial pressure and how this intensive parameter affects material properties essential to all aspects of nuclear oxide fuels. Its effect on point defect concentrations hence on a wide range of crucial properties such as electrical conductivity, anion and cation self- or chemical diffusion, trace diffusion of xenon or creep are reviewed. Conversely, using UO_2+x or MOX fuels as models for other oxide fuels, we show that characterizing the dependence of these properties upon oxygen partial pressure, temperature and impurity content enables the determination of associated atomic processes and defect characteristics (defect charge, atomic migration mechanisms, defect formation and migration energies). A method is suggested for assessing such experimental data against first principles calculations. Finally, we tentatively discuss the extent to which these concepts apply to the practical case of irradiated fuels and make suggestions that could help model these extremely complex systems more predictively.

Tungsten laminate materials constructed from rolled sheet are promising candidates for structural applications in future fast neutron fusion and fission nuclear systems. However their response under irradiation has not been studied. High-energy neutron damage will cause both cascade damage and also composition change through transmutation, producing up to 10% Re after in reactor service. In this study, W-5%Re was used as an analogue to such effects, and was ion-implanted to assess effects of radiation damage.
Rolled tungsten 5 wt% rhenium sheet was studied in two microstructural variants: (a) as received with a high dislocation density (mean value of 1.4×10¹⁴lines/m²), measured using HR-EBSD, and pancake shaped grains with a thickness ofasymp;200nm and (b) annealed at 1400^oC for 24 hours to produce equiaxed grains with average grain size of asymp;90 µm and low dislocation density (with a mean value of 4.8×10¹³ lines/m²). Both materials were ion implanted with 2MeV W+ ions at 300^oC to damage levels of 0.07, 0.4, 1.2, 13 and 33 displacements per atom (dpa). Nanoindentation was used to measure the change in hardness after implantations. Irradiation induced hardening saturated in the as-received material at an increase of 0.4dpa from the unimplanted hardness of 8GPa at 0.4dpa. In the annealed material saturation does not occur by 13dpa and the hardness change of 1.3GPa from the unimplanted hardness of 6.2GPa was over four times higher. At 33dpa both material types showed a further increase in hardening of 2.1GPa (as-received) and 3.2GPa (annealed). Atom probe tomography showed no clustering of Re at 13dpa and below. At 33dpa clusters of asymp;4nm diameter with a rhenium concentration of asymp;11% were seen in both material types. In both cases the number density and volume fraction are similar at asymp;3100 x1000/µm³ and volume fraction of asymp;13%.
These differences in radiation response are likely to be due to the high damage sink density in the as-received microstructure in the form of dislocation networks, as even in the as-received material the average grain size is too large to provide sufficient sinks. Initially this provides a large sink network for radiation damage resulting in less hardening in the rolled material. However at 33dpa the formation of rhenium clusters occurs at similar levels in both material conditions. These dominate the hardening mechanisms and result in secondary hardening at high damage levels. This work demonstrates the advantage of using such nanostructured tungsten sheet in composite materials for structural applications as they will have improved radiation resistance as compared to bulk tungsten products, at low damage levels. At higher damage levels changes in chemistry may result in radiation induced segregation dominating and extreme hardening and the possibility of embrittlement occurring. This study also highlights the danger of using idealized annealed microstructures for radiation damage studies due overestimations of hardening responses.

Nanocrystalline oxides are of high interests for a wide range of applications due to their exceptional size-dependent materials properties. The effective use of nanostructured oxides in nuclear related applications requires better understanding of these materials performance in harsh environments, including high temperature and extreme radiation. In particular, nanostructured oxides are considered as potential candidates to meet the demand for advanced fuels and cladding materials that can withstand extreme radiation environments with improved accident tolerance over a long period of time, and with improved performance in advanced nuclear energy systems. Cubic ceria (CeO₂) and zirconia (ZrO₂) are well known ionic conductors that are also isostructural with urania, plutonia, and thoria-based nuclear fuels. Understanding irradiation response of ionic-covalent materials is important for advanced nuclear energy systems.
Ion beam processing is an effective approach to tailor size-dependent material properties of oxide-based nanomaterials. Grain growth of nanocrystalline materials is generally thermally activated, but can also be driven by irradiation at much lower temperature. Under ion irradiation, defect production and ionization effect lead to effective modification of interface volume in nanocrystalline ceria and zirconia. Experimental results have shown that both high electronic energy loss and nuclear energy loss lead to disorder and radiation-induced growth of the crystallite size is a function of total energy deposited. Atomistic simulations by adding high levels of disorder in the simulation cell have revealed fast grain boundary (GB) movements due to the present of high-level disorder in the close proximity to GBs, and the results is in good agreement with our the experimental results. The coupling of energy deposition to the electronic and lattice structures should both be taken into consideration when engineering nanostructural materials.
This work was supported by the Materials Science of Actinides, an Energy Research Frontier Center supported by the U.S. Department of Energy, Basic Energy Sciences.

Grain boundaries have been shown to be effective sinks for radiation-induced point defects and associated dislocation loops. This observation suggests that nanocrystalline materials are good candidates to resisting mechanical property degradation in irradiation environments. However, nanometer-scale grains tend to be highly unstable with respect to thermal annealing. Recently, several alloy materials have proven to be resistant to grain growth at temperatures up to 400 0C. Are these materials also resistant to irradiation induced creep? As long as temperature is held below that at which diffusional creep initiates, this strategy may prove effective if radiation-enhanced diffusion enables point defect to reach the grain boundaries. In this preliminary study, we explore electroplated NiW films of 100 nm thickness with grain sizes as small as 3 nm. Both cathodic and pulsed plating methods on electropolished Cu substrates are used to incorporate W into the Ni films and control grain size. An anneal above 125 0C tends to relax the grain boundaries, however, resulting in higher grain boundary ordering. Therefore, we self-irradiate films with Ni ions that have and have not been annealed to determine their relative effectiveness in gettering defects. Analysis is conducted by high resolution transmission electron microscopy.

Oxide dispersion strengthened ferritic alloys have superior radiation tolerance and thus become appealing candidates as fuel cladding materials for next generation nuclear reactors. In this study we constructed a model system, Fe/Y₂O₃ nanolayers with individual layer thicknesses of 10 and 50 nm, in order to understand their radiation response and corresponding damage mitigation mechanisms. These nanolayers were subjected to in situ Kr ion irradiation at room temperature up to ~ 8 displacements-per-atom. As-deposited Y₂O₃ layers had primarily amorphous structure. Radiation induced prominent nanocrystallization and grain growth in 50 nm thick Y₂O₃ layers. Conversely, little crystallization occurred in 10 nm thick Y₂O₃ layers implying size dependent enhancement of radiation tolerance. In situ video also captured grain growth in both Fe and Y₂O₃ and outstanding morphological stability of layer interfaces against Kr ion irradiation. This research is supported by NSF-DMR-Metallic Materials and Nanostructures program.

Abstract
Grain boundaries and interfaces are known to promote radiation tolerance in materials in a nuclear reactor. We offer a novel strategy that can trap radiation-induced point defects between grain boundaries and inhibit them from evolving into extended defects. With optimized superlattice structures, we show using molecular dynamics (MD) simulations that spatial control of defects can be enforced in β-SiC in a radiation environment. We have conducted a series of non-equilibrium radiation cascade simulations both on single crystal SiC and superlattices comprising of several grains boundaries using a modified Tersoff interatomic potential that has been benchmarked to a host of thermo-mechanical and defect properties. While the layered superlattice structures do not reduce the number of defects appreciably, the spatial spread of the defects is significantly altered. In a single crystal, the defects are predominantly distributed in the longitudinal direction - in the direction of the knock - while they are distributed in the transverse direction - perpendicular to the direction of the knock - for the layered structures. Thus we show that layered superlattice structures can be advantageously employed to control the spatial defect distribution in a radiation environment.

Swift heavy ions (SHIs) induce a cylindrical region of structural transformation known as ion track. Ion tracks in SiO₂ can be used to change its refractive index or as a mean to induce anisotropy for etching [1]. Furthermore, it was recently found out that ion irradiation can be used to induce a shape transformation in metal nanocrystals (NCs) that are embedded in silica. Spherical NCs elongate along the ion beam direction and are shaped into nanorods or prolate spheroids. The phenomeon can be exploited to produce large arrays of equally aligned nanocrystals, which is difficult to achieve otherwise. The mechanism by which this transformation occurs is unclear.
The effects of SHI impacts in materials are poorly understood on the atomic scale and a predictive theory is essential for the controllability of the SHI-induced structural changes. We model these using the so-called Two-Temperature Molecular Dynamics method [2-3]. The simulations help us to explain and provide an insight into the fundamentals of ion-solid interactions.
[1] B. Afra et al., J. Phys.: Condens. Matter 25, 045006 (2013)
[2] A.A. Leino, O.H Pakarinen, F. Djurabekova, K. Nordlund, P. Kluth and M. C. Rigway, Materials Research Letters 2, 37 (2014)
[3] A. A. Leino, S. Daraszewicz, O H. Pakarinen, F. Djurabekova, K. Nordlund, B. Afra and P. Kluth, Nucl. Instrum. Methods Phys. Res. B 326 289 (2014)

Crystalline metals in the nuclear energy system often suffer from the radiation-induced embrittlement, resulted from the interaction between of dislocation motions and the microstructural damages by irradiation. This transition to the brittle failure may lower the mechanical stability of the irradiated components, posing the significant threats to the operation and security of the whole systems. In contrast to the crystalline metals, amorphous alloys, often called metallic glasses, have the different plasticity mechanism without involving the dislocation motion and hence are expected to exhibit the different mechanical behavior under the high energetic flux. In this study, we investigated the changes in mechanical properties of the Zr-based metallic glass caused by the proton irradiation at 100, 150 and 200 keV. The experimental results were corroborated by the molecular dynamic simulation to fully understand the atomic level mechanisms of the irradiation effects in the metallic glasses.

A beamline for real-time and in-situ studies of Materials in a Radiation Environment (MRE), to study radioactive materials and radiation effects is being developed at the National Synchrotron Light Source II (NSLS II). The beamline will consist of two endstations. The first endstation will focus on in-situ and time resolved diffraction and spectroscopy studies of radiation-induced processes by combining the NSLS-II capabilities with a suite of ion beam accelerators and ultrafast detectors. The second endstation will be dedicated to examining the structural damage in previously irradiated materials using X-ray diffraction and tomography techniques. This beamline will provide synchrotron X-ray techniques for investigating advanced nuclear structural materials, materials of interest to the nuclear forensics, national security, and fabrication of materials using ion implantation. A satellite building adjacent to NSLS II will be built to house the ion accelerators and more easily accommodate active materials. In addition, a sample preparation laboratory with the capability to load more highly activated samples into containment sample holders and materials storage area will be made available at another location at BNL to prevent delays and limitations due to the total activity present at the beamline.

It has long been postulated that interfaces in immiscible metallic multilayers are defect sinks for radiation induced defects. However there is a lack of in situ evidence to validate such a hypothesis. Recently we have reported that immiscible Ag/Ni multilayers have drastically different responses to He ion and proton radiation [KY Yu et al, Journal of Nuclear Materials, 440 (2013) 310]. Here monolithic Ag and Ni films, and Ag/Ni multilayers with individual layer thickness of 5 and 50 nm were subjected to in situ Kr ion irradiation at room temperature to 1 displacements-per-atom [KY Yu et al, Philosophical Magazine, 93 (2013) 3547]. Monolithic Ag has high density of small loops (4 nm in diameter), whereas Ni has low density of much greater loops (exceeding 20 nm). In comparison dislocation loops, ~ 4 nm in diameter, were the major defects in the irradiated Ag/Ni 50 nm film, while loops were barely observed in the Ag/Ni 5 nm film. At 0.2 dpa, defect density in both monolithic Ag and Ni saturated at 1.6 and 0.2 × 10²³/m³, compared with 0.8 × 10²³/m³ in Ag/Ni 50 nm multilayer at a saturation fluence of ~ 1 dpa. Direct observations of frequent loop absorption by layer interfaces suggest that these interfaces are indeed highly efficient defect sinks. Ag/Ni 5 nm multilayer showed superior morphological stability against Kr ion radiation compared to Ag/Ni 50 nm film. This research is funded by NSF-DMR-Metallic Materials and Nanostructures Program.

Monoclinic LaPO₄ (monazite) is a chemically stable compound with a high melting temperature, high radiation damage resistivity, and very low solubility in water; it has various applications arising from its unique structure. One such is immobilization of some radioactive elements by substituting them for La³⁺ in the LaPO₄ lattice. Low temperature solution synthesis results in formation of a hexagonal hydrate, LaPO₄.1/2H₂O (rhabdophane). The presence of rhabdophane in the compound results in formation of a liquid phase at high temperatures due to availability of excess phosphorus, thus making monazite less applicable as a potential material used in advanced nuclear technologies. Further heat treatment is needed to eliminate rhabdophane from the sample and attain pure monoclinic LaPO₄.
In this study, different batches of LaPO₄ were synthesised at different temperatures using the direct precipitation synthesis method. The synthesis technique in this study coupled with washing and ball-milling the samples yields monoclinic monazite without the formation of rhabdophane; negligible liquid phase is formed during sintering at higher temperatures. The discovered optimal temperature range for direct precipitation synthesis of pure monoclinic monazite is confirmed using XRD. The morphologies and the possible presence of excess phosphorus are studied by SEM imaging method.

Study on the irradiation effects on ferroelectric materials is important to use them in devices exposed to high-energy radiation such as in space, nuclear reactors, and military applications. Most of previous studies on irradiation effects were focused on the Pb(Zr_xTi_1-x)O₃ (PZT), which is the most widely used ferroelectric materials in commercialized devices at present. Recently, as growing interest in lead-free materials owing to the toxicity of lead, (K_0.5Na_0.5)(Mn_0.005Nb_0.995)O₃ (KNMN), which is expected to have comparable properties with the PZT, has been considered as a highly promising candidate for replacing the lead-based ferroelectric materials. Nevertheless, the studies on the irradiation effects of KNMN have never been reported yet. In this work, we investigated on the gamma-ray (γ-ray) irradiation effects on the morphology and electrical properties of the KNMN thin films. The KNMN thin films were prepared using a chemical solution deposition method through a spin-coating process, which were subject to γ-ray with various total irradiation doses from 0-3000 kGy. The grain size of the KNMN films visibly changed with an increase in total dose, while their crystalline quality did not vary significantly. It was observed that the remnant polarization value of the films decreased by ~10%, but their ferroelectric property maintained even after irradiation up to 3000 kGy. Our results suggest that the KNMN films have much high radioactivity resistance, which is comparable with lead-based ferroelectric thin films.

High entropy alloys (HEAs), a new type of material system, have potential application as a structural material in advanced nuclear energy systems due to their impressive mechanical properties. HEAs consist of four or more elements with nearly equimolar ratios, giving rise to high configurational entropy. HEAs can be either face centered cubic (f.c.c) or body centered cubic (b.c.c) phases without formation of brittle phases that are frequently observed in conventional simple intermetallic alloys. It is hypothesized that the high configurational entropy might modify point defect (solute and radiation defects) diffusion processes and thereby promote enhanced recombination of radiation-produced point defects. The goal of the present investigation is to evaluate the microstructure changes of an ion-irradiated f.c.c HEA and compare to changes in conventional FeCrNi and FeCrMn solid solution f.c.c alloys. In this study, a novel 27%Fe-27%Mn-28%Ni-18%Cr single phase f.c.c alloy has been synthesized. The tensile strength of the pre-irradiated samples are similar to commercial austenitic steels, with ultimate tensile strength of ~600 MPa at room temperature, and ~275 MPa at 600 #730;C. The unirradiated samples remain ductile even at liquid nitrogen (~45% elongation). Preliminary characterization has shown Cr enrichment and Ni depletion at grain boundaries following ion irradiation at elevated temperatures, which is opposite of conventional austenitic steels. Four HEA samples have been irradiated by 3 MeV Ni ions to dose levels of 1 dpa and 10 dpa, at 500 #730;C and 600 #730;C. Ab initio modeling has also been conducted to calculate the diffusional parameters within an FeMnNiCr HEA, and these fundamental properties will be discussed in relationship to experimental analytical electron microscopy characterization of grain boundary segregation before and after irradiation.
Research sponsored by the Office of Fusion Energy Sciences, U.S. Department of Energy.

Colloidal silica particles are being intensively studied due to their potential applications in catalysis, intelligent materials, optoelectronic devices, photonic bandgap crystals, masks for lithographic nanopatterning, etc. Moreover, in nanoscale electronic, photonic and plasmonic devices, feature dimensions shrink towards a critical limit, and new experimental approaches have to be explored in lithographic patterning. For this work, spherical submicrometer-sized silica particles were prepared by the sol-gel technique and deposited as a self-assembled monolayer onto high-purity silica glass plates by means of a spin coater system. This monolayer is then used as a mask to create regular arrays of nanoscale features in the sample by MeV Ag ion implantation. On the other hand, previously to the ion implantation, the masks can be modified by MeV Si ion irradiation to tailor the size and arrangement of these embedded features as a function of the ion fluence. Indeed, amorphous glassy materials like silicon dioxide can undergo extreme deformations under exposure to high-energy beams, which induce damage and structural changes in solids due to energy losses of MeV heavy ions via ionization events and atomic collisions. Some of the samples were irradiated at room temperature with Si ions at 4 and 6 MeV and fluences up to 5×10¹⁵ Si/cm², under an angle of 90° with respect to the sample surface. After the irradiation the silica particles turned into oblate particles, as a result of the increase of the particle dimension perpendicular to the ion beam and the decrease in the parallel direction. The size, size distribution and shape of both the array of silica particles were determined by scanning electron microscopy as a function of the sample preparation and ion irradiation parameters.

A growing population and industrial facilities will require more electricity supply which is inexpensive and stable service. Nuclear power plants have played a significant role in this area. For meeting expected future demand, nuclear industry should look for way to produce electricity in more efficient and economical way. Since extension of fuel service life is directly related to the economic feasibility, more accurate prediction of fuel behavior have received attention. Unlike previous fuel performance formalism based on empirical model, a new formalism based on multi - scale is expected to predict fuel behavior with high fidelity. Besides, it can be applied to fuel material as well as any irradiated nuclear materials. In this formalism, atomic scale model of nuclear materials, especially nuclear fuels, is not established yet. Thus, we bring to focus on one of the atomic scale phenomenon in nuclear related materials: primary damage state from displacement cascades in grain boundary of uranium dioxide.
When material is irradiated by high energetic small particles such as neutron, one atom in the material lattice is collide with the small particle. And the atom also has kinetic energy in the order of keV. It is called primary knock-on atom (PKA). This PKA present very complex collision with neighbor atoms and make atoms displaced. Some of them remain as defects as time goes on. It have been typically described by molecular dynamics simulation. Particularly, we concentrate on the situation in which PKA is placed on grain boundary region. Since grain boundary is known as sink for defects, cascade morphology and defect properties show different behavior compared to collision in bulk region. Also, it is expected that ionic materials like uranium dioxide can show novel feature due to long range Coulomb effect between ions. And evolution of thermal boundary resistance after collision will be calculated from Green - Kubo formula. Besides, as application for grain boundary, we plan to investigate nanocrystal in radiation condition. So, finding optimum grain size can suggest new form of nuclear fuel nanocrystalline materials design.

Low energy ion implantation has the potential to modify the interfacial properties of SiC without the use of heavy impurity doping and high temperatures [1]. In this paper, we examine the effect of prior implantation with P or C ions at low energy on the electrical characteristics of Ni/Ti/Au contacts to n-SiC. The substrates were comprised of 3H n-SiC (1 x 10¹⁸ cm^-3) as grown epitaxially on Si. The SiC was implanted with either P or C ions at an energy of 5.0 keV using doses in the range 10¹³- 10¹⁵ ions/cm². Arrays of circular transmission line patterns (CTLM) [2] were then fabricated on the surface of the SiC to enable the measurement of sheet resistance, R_s and specific contact resistance, ρ_c. In these results, the surfaces of SiC previously implanted with P have shown lower values of R_s (300 Omega;/square) and ρ_c (~1 x 10^-6 Omega;cm²) than for C implants. Both R_s and ρ_c have increased with the dose and reduced with annealing temperature. These new trends have been discussed in relation to the effect of the implant ion species on the depth profile of the near surface damage. 1. F Lui, CH Li, AP Pisano, C Carraro and R Maboudian, J.Vac.Sci.Technol., A28(5) 1259 (2010). 2. Y Pan, , GK Reeves, PW Leech, and AS Holland, IEEE Trans on Electron Dev, 60(3) 1202 (2013).

Lithium aluminate (LiAlO₂) is a candidate material for radiation detection applications involving optically stimulated luminescence and thermoluminescence. The presence of ⁶Li nuclei with a large cross-section for thermal neutron absorption suggests that this material may be especially well-suited for neutron detection applications (enrichment well above 7.5% natural abundance of ⁶Li is easily achieved). In our study, commercially available LiAlO₂ crystals are irradiated at room temperature with x rays (60 kV). The primary defects are characterized using electron paramagnetic resonance (EPR), photoluminescence (PL), optical absorption (OA), and thermoluminescence (TL). Lithium vacancies are the dominant hole trap. At room temperature, radiation-induced holes are localized on oxygen ions adjacent to lithium vacancies and form paramagnetic defects with S = 1/2. The EPR spectrum of these trapped-hole centers has an eleven-line hyperfine pattern due to equal interactions with two ²⁷Al nuclei (I = 5/2, 100% abundant). The g matrix determined from the angular dependence of the EPR spectrum verifies that the hole is in a p orbital on the oxygen ion. A broad OA band in the visible is correlated with this EPR spectrum. These trapped-hole centers thermally decay when the crystal is heated above 100 °C. The dominant electron traps are associated with Fe impurities. An intense EPR spectrum from Fe⁺ ions (3d⁷ with an effective S of 1/2) is observed after a room-temperature x-ray irradiation. These Fe⁺ ions are formed when electrons become trapped at Fe²⁺ ions initially present on lithium sites. EPR spectra from Fe³⁺, Cr³⁺, and Cu²⁺ are also observed after room-temperature x-ray irradiations. Ultraviolet excitation and emission peaks from Cu⁺ ions are seen in PL spectra taken at room temperature. An intense TL peak appears near 100 °C. This TL peak correlates with the thermal decay of the holes trapped adjacent to the lithium vacancies and the decay of the radiation-induced Fe⁺ ions on lithium sites (the Fe⁺ ions return to the Fe²⁺ state). Additional less intense TL peaks appear between 100 and 300 °C. Thus far, EPR signals from oxygen vacancies have not been detected before or after an x-ray irradiation at room temperature.

During service in a light-water reactor, the chemistry of uranium dioxide (UO₂) fuel evolves to higher oxidation states of uranium with burn up. In addition, fission products, including rare earth elements, are simultaneously incorporated into the fuel matrix. At high burnups, this chemical evolution results in the formation of complex (multi component) oxides that are encompassed within the ternary oxide system, UO₂-UO₃-Ln2O₃ (Where Ln stand for a Lanthanide, i.e. an RE³⁺ cation). Oxidation of UO₂ is important for understanding fuel variations during reactor operation, and for predicting the chemistry of spent fuels. In the present work, we have studied the structure of complex oxides in the UO₂-UO₃-Ln₂O₃ ternary oxide system using density functional theory (DFT). In particular, we have examined compounds along the line that connects UO₂ with 50% Ln₂O₃ in the ternary phase diagram. We have investigated two model compounds: ULa₂O₆ and ULaO₄. These compounds maintain a 1:2 cation to anion stoichiometric ratio and possess a crystal structure similar to fluorite. To model and analyze the structure of these complex oxides we have used a unit cell based on a layered atomic model. In the layer model, the unit cell is composed of alternating planes of anions and cations. We have considered single species cation layers versus mixed species cation layers. This approach simplifies the task of identifying the ordering arrangements within the planes of cations. Due to the highly correlated nature of the f-electrons in uranium, DFT+U has been employed. We report here on the structural stability of the ULa₂O₆ and ULaO₄ compounds in the context of the cation layer assumptions described above.

Finding a silicon-based light emitting material is one strategy used to explore advancements in optoelectronic devices. Efficient light emission from silicon is challenging due to its indirect band gap. Ion Implantation is a material-processing technique that can be used to modify nonstoichiometric composition and electronic properties of a material. In this study carbon ions were implanted into p-type silicon at ion fluences of 3, 5, 7, and 10 x 10¹⁶ C⁺/cm². The ion implantation was carried out at both ambient (ion heating only) and high temperature (HT) pre-heated substrate at 400 °C to examine the effect of in situ post-annealing on Schottky-type light emitting diodes (LEDs) fabricated from the C⁺ implanted silicon. Ion-implanted samples were post-annealed at 1000 °C for 1 h in flowing nitrogen. Aluminum and semitransparent gold contacts were applied to the C⁺ implanted silicon using a physical vapor deposition, and the contacts were improved by firing the devices at 400 °C in flowing nitrogen for 1 h. The LEDs had turn-on voltages of ~2 V and showed red-orange luminescence, due to damaged silicon and oxide defects. LEDs that were fabricated from the HT implanted silicon showed higher luminescence intensity than the devices fabricated from ambient implanted silicon.

The in-service and off-normal behaviour of structural materials is largely determined by the kinetics of thermally activated atomic-scale processes. Monte Carlo models are widely used for the study of microstructural and microchemical evolution of these materials under irradiation. However, they often link explicitly the relevant activation energies to the energy difference between local equilibrium states. By combination of Monte Carlo and Molecular Dynamics techniques we show that the explicit link between two energies may lead to the misinterpreting of the actual path the object chooses for its motion. We provide a simple example (di-vacancy migration in iron) in which we use a rigorous activation energy calculation by means of both empirical interatomic potentials (drag method by applying MD simulations) and density functional theory methods (nudged elastic band method, implemented in VASP). Both cases clearly show that such a link is not granted, revealing a migration mechanism that a thermodynamics-linked activation energy model cannot predict. Such a mechanism is, however, fully consistent with thermodynamics. This example emphasizes the importance of basing Monte Carlo methods on models where the activation energies are rigorously calculated, rather than deduced from widespread heuristic equations.
By applying similar approach we also compare di-vacancy migration mechanisms in iron and in tungsten.

Austenitic face-centered cubic (fcc) alloys exhibit notable properties such as corrosion and creep resistance, and are being developed for future fusion and advanced fission reactors. Studies of point defect density in austenitic and Fe-Cr-Ni ternary alloys have typically been performed on large grained samples resulting in limited knowledge of the effect of grain size and individual microstructural features (e.g. grain boundaries) on radiation damage. Materials with grain diameters ranging from a few to ~100 nm, termed nanocrystalline materials, have measured less radiation damage than the same samples with coarser grain sizes (of the order of a few microns) exposed to the same conditions. An investigation of the effects of irradiation on nanocrystalline nickel-chromium (NiCr) is conducted as a discernible yet comprehensive study of how grain boundaries affect radiation damage in fcc alloys. This work presents a study of point defect cluster formation coupled with grain size, grain boundary character, and radiation dose in freestanding nanocrystalline NiCr films. Films were irradiated ex-situ on 3 mm TEM grids to approximately 2, 5, and 12 dpa at 5000C with 5 MeV Ni⁴⁺ ions at Sandia National Laboratories Ion Beam Facility then characterized using the TEM and electron diffraction orientation mapping. The results were corroborated with large-grained bulk NiCr samples irradiated under the same conditions.

Recently the laser sintering process has attracted researchers&’ attention due to advantages such as the possibility of using very high heating and cooling hates, and the possibility of sintering of materials with high melting points. In this work, laser sintering technique was used to produce dense, homogeneous and crack-free ceramics bodies of different systems. In this method a CO₂ laser is used as the main heating source during sintering process, keeping the laser beam at the center of the sample throughout the sintering. Before the irradiation, the samples were heated up to 300 0C following a heating rate of 50 0C/min and kept in this temperature throughout the process. After this pre-heating, the laser power is raised at a linear rate up to maximum power, and kept at this value for few seconds. After the first face sample irradiation, this process is repeated on the other sample side. Using these conditions, the total laser sintering time of CaCu₃Ti₄O₁₂, BaTiO₃, SrAl₂O₄, CaAl₂O₄, BaAl₂O₄ and Y₃A_l5O₁₂ was smaller than 5 min. The laser sintered ceramics presented distinct properties from those observed in conventionally sintered samples. For example, CaCu₃Ti₄O₁₂ ceramics presented dielectric constant values in the range of 2000 (with dielectric loss of ~ 0.06) when measured at 10 kHz at room temperature. This dielectric loss value is sensibly lower than observed in Literature for ceramics sintered by conventional methods. In MAl₂O₄: Eu, Dy (M = Ca, Sr, Ba) ceramics was observed the reduction of Eu³⁺ to Eu²⁺ without the need to use a reducing atmosphere, that is, the laser sintering is performed in open atmosphere. As consequence of Eu reduction, these samples showed the persistent luminescence phenomenon with intensity visible to the naked eye for at least 4h. The other compositions also presented distinct properties that will be presented during the Meeting.

One of the most important observations from Literature review of Ru based diffusion barriers is that its barrier failure is primarely due to its nano-crystalline structure. We undertook innovative strategies for preparing high performance ultra-thin Ru based diffusion barrier layers that are amorphous and highly resistant to crystallization at functional processing temperatures. Systematic studies of 3 strategies and associated controlled experiments were undertaken : (1) Ion irradiation induced amorphization of the Ru layer, (2) N2 plasma modification of near surface Ru layer, and (3) Al doping induced amorphization of nitrided Ru diffusion barrier layer. All these studies were done on 5 nm Ru thin films. The above comparative studies show the crystallization limitations of pure Ru layers and a marked improvement in crystallization inhibition and delay for the nitrided Ru thin films with Al doping. The Al doping was shown to stabilize the nitrogen in the 5 nm Ru thin film barrier thus inhibiting its crystallization and leading to an effective diffusion barrier for copper beyond 450 ^oC with minimal diffusion at 500 ^oC annealing temperature. This represents a gain in processing temperatures of 150 ^oC -200 ^oC over the performance of as prepared pure Ru thin film barriers.

The cerium-doped lanthanum halide crystals have gained special interest due to their high density and atomic number, which results in excellent scintillation properties and higher detection efficiencies in comparison to NaI(Tl). A new analytical method of efficiency calibration is proposed for cylindrical lanthanum bromide (LaBr₃:Ce) scintillation detectors. This method depends on the accurate analytical calculation of two important factors; the path length d, the photon traverses within the active volume of a gamma detector, and the geometrical solid angle, subtended by the source to the detector at the point of entrance. In addition, the attenuation of photons by the detector housing materials is also treated by calculating the photon path length through these materials. The comparisons with the experimental and Monte Carlo method data reported in the literature indicate that the present method is useful in the efficiency calibration of the cylindrical lanthanum bromide (LaBr₃:Ce) scintillation detectors.

It was recently shown that the energy resolution of Ce-doped LaBr₃ scintillator radiation detectors can be crucially improved by co-doping with Sr, Ca, or Ba. Here we outline a mechanism for this enhancement on the basis of electronic structure calculations. We show that Sr dopants create and bind to Br vacancies, resulting in stable neutral complexes. The association with Sr causes the deep vacancy level to move toward the conduction band edge. This is essential for reducing the effective carrier density available for Auger quenching during thermalization of hot carriers. Subsequent de-trapping of electrons from the complexes can activate Ce dopants that have previously captured a hole leading to luminescence. This mechanism implies an overall reduction of Auger quenching of free carriers, which is expected to improve the linearity of the photon light yield with respect to the energy of incident electron or photon. Optical properties of the Ce-Sr-vacancy triple complex are discussed and compared to experiment. Prepared by LLNL under Contract DE-AC52-07NA27344.

BiI₃ is a wide band-gap compound semiconductor with a high effective atomic number that is anticipated to exhibit higher detection efficiency than other compound semiconductors such as HgI_2, PbI₂, and CdZnTe. This makes BiI₃ of particular interest for moderate and high energy gamma-ray detection applications. However, the low resistivity of BiI₃ results in high leakage currents and degrades the electrical properties and detecting performance of the detectors. Here we show that the main reason for the low resistivity in BiI₃ is due to the high volatility of iodine and the high concentration of intrinsic Schottky defects. Furthermore, we will discuss novel defect engineering strategies that successfully mitigate the obstacles associated with iodine vacancies in the material. The electrical properties and radiation response of both undoped and Sb-doped BiI₃ (SBI) single crystals grown via the vertical Bridgman growth technique will be presented. Most importantly, we will demonstrate the first ever recorded gamma-ray spectrum using BiI₃-based detectors.