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The influence of radiation-induced (1 MeV energy H+ to ∼0.1 displacements per atom (dpa) at 450 °C), nonequilibrium point defect populations on mass transport is studied with an integrated campaign...

The stability of structural materials in extreme nuclear reactor environments—with high temperature, high radiation, and corrosive media—directly affects the lifespan of the reactor. In such extreme environments, an oxide layer on the metal surface acts as a passive layer protecting the metal underneath from corrosion. To predict the irradiation effect on the metal layer in these metal/oxide bilayers, nondestructive depth-resolved positron annihilation lifetime spectroscopy (PALS) and complementary transmission electron microscopy (TEM) were used to investigate small-scale defects created by ion irradiation in an epitaxially grown (100) Fe film capped with a 50 nm Fe2O3 oxide layer. In this study, the evolution of induced vacancies was monitored, from individual vacancy formation at low doses—10−5 dpa—to larger vacancy cluster formation at increasing doses, showing the sensitivity of positron annihilation spectroscopy technique. Furthermore, PALS measurements reveal how the presence of a metal–oxide interface modifies the distribution of point defects induced by irradiation. TEM measurements show that irradiation induced dislocations at the interface is the mechanism behind the redistribution of point defects causing their accumulation close to the interface. This work demonstrates that the passive oxide layers formed during corrosion impact the distribution and accumulation of radiation induced defects in the metal underneath and emphasizes that the synergistic impact of radiation and corrosion will differ from their individual impacts.

A cadmium germanium arsenide compound, Cd3Ge2As4, was synthesized using a double-containment fused quartz ampoule method within a rocking furnace and a melt-quench technique. The crystal structure was determined from single-crystal X-ray diffraction (SC-XRD), scanning and transmission electron microscopies (i.e. SEM, STEM, and TEM), and selected area diffraction (SAD) and confirmed with electron backscatter diffraction (EBSD). The chemistry was verified with electron energy loss spectroscopy (EELS).

Understanding the structural evolution and reduction-oxidation behavior of nuclear fuel and cladding during operation is essential for predicting performance during and after service in light water reactors. Using TEM/STEM imaging of cross-sections of the fuel-cladding oxide interface region of high burnup BWR fuel, fission recoil radiation was demonstrated to not only stabilize the tetragonal phase of ZrO2 at temperatures well below the equilibrium temperature, but also to cause grain growth proportional to the fission recoil radiation damage. The tetragonal phase ZrO2 was exclusively present (no monoclinic phase) only in the region where fission product metal particles were found (∼6 μm depth).

Positron annihilation spectroscopy and transmission electron microscopy yield previously unknown insights on radiation damage., The nondestructive investigation of single vacancies and vacancy clusters in ion-irradiated samples requires a depth-resolved probe with atomic sensitivity to defects. The recent development of short-pulsed positron beams provides such a probe. Here, we combine depth-resolved Doppler broadening and positron annihilation lifetime spectroscopies to identify vacancy clusters in ion-irradiated Fe and measure their density as a function of depth. Despite large concentrations of dislocations and voids in the pristine samples, positron annihilation measurements uncovered the structure of vacancy clusters and the change in their size and density with irradiation dose. When combined with transmission electron microscopy measurements, the study demonstrates an association between the increase in the density of small vacancy clusters with irradiation and a remarkable reduction in the size of large voids. This, previously unknown, mechanism for the interaction of cascade damage with voids in ion-irradiated materials is a consequence of the high porosity of the initial microstructure.

A high-intensity proton beam exposure with 181 MeV energy has been conducted at Brookhaven Linac Isotope Producer facility on various material specimens for accelerator targetry applications, including titanium alloys as a beam window material. The radiation damage level of the analyzed capsule was 0.25 dpa at beam center region with an irradiation temperature around 120 degree C. Tensile tests showed increased hardness and a large decrease in ductility for the dual alpha+beta-phase Ti-6Al-4V Grade-5 and Grade-23 extra low interstitial alloys, with the near alpha-phase Ti-3Al-2.5V Grade-9 alloy still exhibiting uniform elongation of a few % after irradiation. Transmission Electron Microscope analyses on Ti-6Al-4V indicated clear evidence of a high-density of defect clusters with size less than 2 nm in each alpha-phase grain. The beta-phase grains did not contain any visible defects such as loops or black dots, while the diffraction patterns clearly indicated omega-phase precipitation in an advanced formation stage. The radiation-induced omega-phase transformation in the beta-phase could lead to greater loss of ductility in Ti-6Al-4V alloys in comparison with Ti-3Al-2.5V alloy with less beta-phase., 23 pages, 9 figures, In Press

A high-strength dual alpha+beta phase titanium alloy Ti-6Al-4V is utilized as a material for beam windows in several accelerator target facilities. However, relatively little is known about how material properties of this alloy are affected by high-intensity proton beam irradiation. With plans to upgrade neutrino facilities at J-PARC and Fermilab to over 1 MW beam power, the radiation damage in the window material will reach a few displacements per atom (dpa) per year, significantly above the ~0.3 dpa level of existing data. The RaDIATE collaboration has conducted a high intensity proton beam irradiation of various target and window material specimens at BLIP facility, including a variety of titanium alloys. Post-Irradiation Examination of the specimens in the 1st capsule, irradiated at up to 0.25 dpa, is in progress. Tensile tests in a hot cell at PNNL exhibited a clear signature of radiation hardening and loss of ductility for Ti-6Al-4V, while Ti-3Al-2.5V, with less beta phase, exhibited less severe hardening. Microstructural investigations will follow to study the cause of the difference in tensile behavior between these alloys. High-cycle fatigue (HCF) performance is critical to the lifetime estimation of beam windows exposed to a periodic thermal stress from a pulsed proton beam. The 1st HCF data on irradiated titanium alloys are to be obtained by a conventional bend fatigue test at Fermilab and by an ultrasonic mesoscale fatigue test at Culham Laboratory. Specimens in the 2nd capsule, irradiated at up to ~1 dpa, cover typical titanium alloy grades, including possible radiation-resistant candidates. These systematic studies on the effects of radiation damage of titanium alloys are intended to enable us to predict realistic lifetimes of current beam windows made of Ti-6Al-4V and to extend the lifetime by choosing a more radiation and thermal shock tolerant alloy., 10 pages, 7 figures, In Proc. of IWSMT-14, 14th International Workshop on Spallation Materials Technology, 11th-17th Nov. 2018 at Fukushima, Japan

Tungsten heavy alloys (WHAs) are a type of ductile phase toughened alloy that are becoming increasingly interesting as an alternative to polycrystalline tungsten for fusion reactor plasma facing material components due to their balanced strength and ductility. To justify their use in the extremely harsh environment of a fusion reactor, understanding detailed microstructural features of WHAs associated with their mechanical property changes is necessary. A 90W-7Ni-3Fe WHA alloy has been chosen to investigate the effect of thermomechanical treatment and microstructural manipulation on the overall effectiveness of deformation accommodation in these bi-phase metallic composites. Both in-situ tensile testing and 3D microstructural analysis of the samples reveal a predominance of microcracking at tungsten grain boundaries that are blunted and arrested by the ductile phase, while there remains little to no instances of interfacial debonding. Thermomechanical treatment of this alloy is found to alter the spherical brittle phase domains into elongated plates, drastically reducing the ductile phase connectivity, and changing the nature of material deformation. Characterization of the ductile phase toughening mechanisms in these materials has provided deeper insight into the underlying physics governing material behavior in these alloys; revealing a surprising interfacial strength between the different phases.

A foil of a metastable β Titanium alloy Ti-15V-3Cr-3Sn-3Al was irradiated at the J-PARC neutrino experimental facility with 1.4 × 1020 30 GeV protons at low temperature (100–130 °C at most), and microstructural characterization and hardness testing were conducted as an initial study on the radiation damage effects of Titanium alloy by the high energy proton beam exposure. Expected radiation damage at the beam center is about 0.06–0.12 displacement per atom. A high density (> 1023 m−3) of a nanometer-sized precipitate was observed by TEM studies, which would be identified as martensite α-phase and athermal ω-phase formed during the solution-treatment process to fabricate metastable β alloy. They did not appear to change substantially after irradiation with protons. In the irradiated specimen, we could not identify an obvious signature of radiation damage distributed along the proton beam profile. Very small, nanometer-scale black dots were present at a low density in the most highly irradiated region, and may be small dislocation loops formed during irradiation. The micro-indentation test indicated that the radiation exposure led to tiny increase in Vickers micro-hardness of ΔHV = 20 at beam center. Atom probe tomography reveals compositional fluctuations that reach a maximum amplitude of 10 at% Ti within a space of

Single crystal cubic phase silicon carbide (3C-SiC) films on Si were implanted to 9.6 × 1016 25Mg+/cm2 at 673 K and annealed at 1073 and 1573 K for 2, 6, and 12 h in an Ar environment. The data from scanning transmission election microscopy (STEM) and electron energy loss spectroscopy (EELS) mapping suggest a possible formation of unidirectionally aligned tetrahedral precipitates of core (MgC2)-shell (Mg2Si) in the implanted sample annealed at 1573 K for 12 h. There are also small spherical voids near the surface and larger faceted voids around the region of maximum vacancy concentration. Atom probe tomography confirms 25Mg segregation dominated by small atomic clusters with local 25Mg concentrations up to 85 at.%. The resulting precipitate size and number density are found to decrease and increase, respectively, probably as a result of the thermal annealing that decomposes the 25Mg-bearing precipitates at the elevated temperatures and subsequent nucleation and growth below 1073 K during the cooling stage. The results from this study provide data needed to fully understand the property degradation of SiC in a high-flux fast neutron environment.

Despite their excellent strength-to-weight ratio, wider use of magnesium (Mg) alloys for light-weight applications is limited by their poor corrosion resistance, especially in chloride-containing environments. The present study shows improved corrosion resistance imparted by laser surface processing (LSP) of a commercial AZ31 (Mg-3Al-1Zn) alloy. Nanosecond laser processing at three different power settings was carried out on the surface of a 1-mm-thick rolled AZ31 sheet. Electrochemical studies and salt spray testing (ASTM B117) indicate substantial enhancement of corrosion resistance in LSP-treated AZ31. The underlying reasons behind improved corrosion resistance of the LSP-AZ31 surface have been studied through detailed microstructural characterization and chemical analysis by SEM, TEM, and XPS. Formation of a ∼0.5 μm thick mixed metal (Mg, Al) oxide surface film, together with refinement in the size and number density of Al-Mn intermetallic particles, are shown to play a major role toward improved corrosion resistance after LSP treatment of the AZ31 alloy.

This study reports on dose rate effects in self-ion irradiated monocrystalline tungsten (mono-W) and polycrystalline tungsten (poly-W). Both mono-W and poly-W were irradiated at 900 K to 1 dpa at two dose rates, 10−4 and 10−3 dpa/s. An additional mono-W disk was irradiated at 300 K up to 44 dpa. The irradiated samples were analyzed using Rutherford backscattering spectrometry in channeling geometry (RBS/C) and transmission electron microscopy. The RBS/C data suggest that self-interstitial atoms in mono-W are extremely mobile at 300 K and higher; the disordering rate in mono-W irradiated at 900 K is ~6 times lower at 10−4 than 10−3 dpa/s at the depth of the damage peak. There is a uniform distribution of voids in both mono-W and poly-W irradiated at 900 K. The observed void diameters were corrected, which ranged from 1.5 to 3.4 nm. The void number density is on the order of 2 × 1017 voids/cm3 in mono-W irradiated at 900 K to 1 dpa at 10−3 dpa/s. The growth rate of the voids in W irradiated at 900 K is higher at 10−4 than 10−3 dpa/s. At the same dose rate, voids grow faster in mono-W than poly-W. A simplified three-dimensional model is proposed to assess the dose rate effects based on the diffusion and interaction of migrating point defects from two consecutive damage cascades.

Radiation induced segregation and precipitation phenomena in an in-service boiling water reactor 304 stainless steel component were investigated using directly correlated 3D-atom probe tomography and scanning transmission electron microscopy. Significant quantitative differences in measured segregation at grain boundaries were found between the atom probe and energy dispersive spectroscopy measurements of the exact same locations. In particular, a much stronger Si segregation (~10 atomic% via atom probe versus ~4 atomic% via electron microscopy) and different Cr profile shapes were detected that are critical to models of radiation induced segregation and stress corrosion cracking behavior. These quantitative differences highlight the need for comparative microscopy and critical evaluation of limitations in each analytical method. Elemental segregation to dislocations and conjoined-clusters were also highlighted by atom probe; confirming and expanding upon what has been observed in test reactor neutron and accelerator-based ion irradiations.

In-situ He + ion irradiation is performed under a helium ion microscope to study nanostructural evolution in polycrystalline γ-LiAlO 2 pellets. Various locations within a grain, across grain boundaries and at a cavity are selected. The results exhibit He bubble formation, grain-boundary cracking, nanoparticle agglomeration, increasing surface brightness with dose, and material loss from the surface. Similar brightening effects at grain boundaries are also observed under a scanning electron microscope. Li diffusion and loss from polycrystalline γ-LiAlO 2 is faster than its monocrystalline counterpart during H 2 + ion implantation at elevated temperatures. There is also more significant H diffusion and release from polycrystalline pellets during thermal annealing of 300 K implanted samples. Grain boundaries and cavities could provide a faster pathway for H and Li diffusion. H release is slightly faster from the 573 K implanted monocrystalline γ-LiAlO 2 during annealing at 773 K. Metal hydrides could be formed preferentially along the grain boundaries to immobilize hydrogen.

An investigation of the influence of helium on damage evolution under neutron irradiation of an 11 at% Al, 19 at% Cr ODS ferritic PM2000 alloy was carried out in the High Flux Isotope Reactor (HFIR) using a novel in situ helium injection (ISHI) technique. Helium was injected into adjacent TEM discs from thermal neutron 58Ni(nth,γ) 59Ni(nth,α) reactions in a thin NiAl layer. The PM2000 undergoes concurrent displacement damage from the high-energy neutrons. The ISHI technique allows direct comparisons of regions with and without high concentrations of helium since only the side coated with the NiAl experiences helium injection. The corresponding microstructural and microchemical evolutions were characterized using both conventional and scanning transmission electron microscopy techniques. The evolutions observed include formation of dislocation loops and associated helium bubbles, precipitation of a variety of phases, amorphization of the Al2YO3 oxides (which also variously contained internal voids), and several manifestations of solute segregation. Notably, high concentrations of helium had a significant effect on many of these diverse phenomena. These results on PM2000 are compared and contrasted to the evolution of so-called nanostructured ferritic alloys (NFA).

Polycrystalline 3C-SiC was sequentially irradiated at 400 and 750 °C with 120 keV He 2+ and 4 MeV Kr 15+ ions to 10 17 and 4 × 10 16 cm −2 , respectively. The Kr 15+ ions penetrated the entire depth region of the He 2+ ion implantation. Three areas of He 2+ , Kr 15+ and He 2+ + Kr 15+ ion implanted SiC were created through masked overlapping irradiation. The sample was subsequently annealed at 1600 °C in vacuum and characterized using cross-sectional transmission electron microscopy and energy-dispersive X-ray spectroscopy. Compared to the He 2+ ion only implanted SiC, He cavities show a smaller size and higher density in the co-implanted SiC. At 25 dpa, presence of He in the co-implanted 3C-SiC significantly promotes cavity growth; much smaller voids are formed in the Kr 15+ ion only irradiated SiC at the same dose. In addition, local Kr migration and trapping at cavities occurs, but long-range Kr diffusion in SiC is not observed up to 1600 °C.

Crystalline materials exhibit long-range ordered lattice unit, within which resides nonperiodic structural features called defects. These crystallographic defects play a vital role in determining the physical and mechanical properties of a wide range of material systems. While computer vision has demonstrated success in recognizing feature patterns in images with well-defined contrast, automated identification of nanometer scale crystallographic defects in electron micrographs governed by complex contrast mechanisms is still a challenging task. Here, building upon an advanced defect imaging mode that offers high feature clarity, we introduce DefectSegNet - a new convolutional neural network (CNN) architecture that performs semantic segmentation of three common crystallographic defects in structural alloys: dislocation lines, precipitates and voids. Results from supervised training on a small set of high-quality defect images of steels show high pixel-wise accuracy across all three types of defects: 91.60 ± 1.77% on dislocations, 93.39 ± 1.00% on precipitates, and 98.85 ± 0.56% on voids. We discuss the sources of uncertainties in CNN prediction and the training data in terms of feature density, representation and homogeneity and their effects on deep learning performance. Further defect quantification using DefectSegNet prediction outperforms human expert average, presenting a promising new workflow for fast and statistically meaningful quantification of materials defects.

A cadmium germanium arsenide compound, Cd3Ge2As4, was synthesized using a double-containment fused quartz ampoule method within a rocking furnace and a melt-quench technique. The crystal structure was determined from single-crystal X-ray diffraction, scanning and transmission electron microscopies, and selected area diffraction and confirmed with electron backscatter diffraction. The chemistry was verified with electron energy loss spectroscopy., A cadmium germanium arsenide compound, Cd3Ge2As4, was synthesized using a double-containment fused quartz ampoule method within a rocking furnace and a melt-quench technique. The crystal structure was determined from single-crystal X-ray diffraction (SC-XRD), scanning and transmission electron microscopies (i.e. SEM, STEM, and TEM), and selected area diffraction (SAD) and confirmed with electron backscatter diffraction (EBSD). The chemistry was verified with electron energy loss spectroscopy (EELS).

The NT-02 neutrino physics target made of the isotropic graphite grade produced neutrinos for the MINOS and MINERVA high-energy physics experiments. The segmented, 95-cm-long NT-02 target was bombarded with a 340 kW, Gaussian 1.1 mm sigma beam of 120 GeV protons reaching $6.516\ifmmode\times\else\texttimes\fi{}{10}^{20}$ protons on target and a peak fluence of $8.6\ifmmode\times\else\texttimes\fi{}{10}^{21}\text{ }\text{ }\mathrm{protons}/{\mathrm{cm}}^{2}$. Reductions in detected neutrino events during the experiment were attributed to radiation-induced damage on the target material leading to the NT-02 target replacement. With future neutrino physics targets aiming at the multimegawatt power regime, identifying life expectancy or fluence thresholds of target materials is of paramount importance, and, therefore, pinpointing the exact cause and target failure mode triggering the neutrino yield reduction is critical. To help unravel the effects of the 120 GeV beam on the isotropic graphite structure at the microstructural or lattice level, x-ray beams from National Synchrotron Light Source II were utilized to study failed in-beam as well as intact NT-02 target segments. The primary objective was to arrive at a scientifically sound explanation of the processes responsible for the target failure by correlating macroscopic observations with microstructural analyses. Results from transmission electron microscopy studies were integrated in assessing the microstructural evolution. The x-ray diffraction study revealed (a) the diffused state reached by the graphite microstructure within the $1\ensuremath{\sigma}$ of the beam where the graphite lattice structure transforms into a nanocrystalline structure, a finding supported by electron microscopy examination, thus providing an indication of the fluence threshold, and (b) the dominant role of the irradiation temperature profile exhibiting a high gradient from the beam center to the heat sink and aggravating the damage induced in the microstructure by the high proton fluence. The effects of the 120 GeV protons on the isotropic graphite target structure are corroborated by observed damage induced by 160-MeV protons and by fast neutrons to comparative doses on similar graphite, an assessment that will aid the design of next-generation megawatt-class neutrino targets.

Deliberately designed nanostructured materials containing a high density of dopant stabilized interfaces provide energetically favorable sites for the annihilation of defects that form within the crystalline matrix due to irradiation. In this study, we explore the role of dopants on the mechanisms governing this behavior in a nanocrystalline W alloy doped with 20 at.% Ti using heavy ion irradiation experiments. Defect evolution is mapped in situ up to the saturation dose and bridged to extreme dose behavior using ex situ measurements with microstructural analysis focused on quantifying both the defect state and the impact of ion irradiation on the nanocrystalline grain structure. Compared with a nominally pure nanocrystalline W film, the W-20 at.% Ti alloy exhibits smaller defect loops and a delayed saturation dose with a period of irradiation induced grain growth occurring during transient damage accumulation. Microstructural evolution is modeled in the context of cascade-induced thermal spikes and reveals that the alloy evolves to a much finer nanocrystalline grain size relative to the predictions for pure W, indicative of Ti stabilizing the nanostructure against irradiation induced grain growth. In situ mapping of defect evolution during the growth of a single grain confirms the correlation between the global defect density trends and evolution of the microstructure, thus providing insights into the effect of Ti on the grain boundary sink strength through the transient defect accumulation and recovery stages.

For the advancement and development of nuclear systems used in heavy irradiation environments such as fusion DEMO reactors, fission reactors, fast reactors, and accelerator driven target systems, it is necessary to fully understand the changes of mechanical properties and the other properties of the materials induced by irradiation and to clarify the synergistic effect of displacement damage and helium generation. In this study the mechanical property changes and microstructural development induce by displacement damage and helium production have been mainly examined in austenitic stainless steels, 316FR and type304, and ferritic/martensitic steel, HCM12A, irradiated at around 550oC in JRR-3M reactor and/or JOYO fast reactor. At 550°C, 316FR steel was superior to 304 steel in terms of the amount of ductility and strength with respect to irradiation resistance. It is noteworthy that at 550 °C as well as room temperature, the higher fracture strength of the 316FR steels is a remarkable result. It is found that helium atoms strongly influenced on creep lifetime of the irradiated austenitic stainless steel, 316FR. It was found that the ratio of creep rupture time is slightly lower than the lower limit of previous study's Miyaji and co-workers in the region from 0.01 appm to 1 appm. It is also found that the lower limit of reduction ratio of creep rupture time (irradiation specimen to unirradiation one) does not decrease linearly with the helium production above 10 appm up to about 33 appm. dpa enhanced the reduction of creep lifetime. Recent R&D of high-energy accelerator driven target systems used under heavy irradiation environment is also introduced and discussed for high radiation resistance materials such as Ti alloys with very high number density of nano size precipitate, which have been studying under RaDIATE collaboration, from points of view of irradiation damage and materials development.

Nuclear reactor environments provide a unique scientific and engineering challenge wherein materials must tolerate prolonged exposure to concurrent irradiation, elevated temperatures, and corrosive media. However, uncontrolled variability in material composition and structure often prohibits truly single-variable experiments that can reveal basic aspects of environmental damage. Magnetron sputtering is used here to provide a more controlled model system for these fundamental studies, yielding reproducible single-crystal Fe and FeCr thin films containing 8 and 18 at.% Cr. Electron microscopy is used to determine the systematic correlations between growth conditions and the resulting film microstructure and surface morphology. It is found that the substrate temperature and applied radio frequency (RF) bias can be tuned to obtain consistent homogeneous and single crystal films with a minimal amount of Ar impurities from the RF bias process. Epitaxial, single-crystal Fe films are obtained on MgO substrates at 500 °C with 10 Watt (W) RF bias deposition. However, when Cr is alloyed with Fe, higher substrate temperatures (600 °C) and applied RF biases (15 W) are required to achieve a similar epitaxial single-crystal FeCr film. Accelerated molecular dynamics simulations reveal that Cr impedes surface transport, explaining the need for higher temperature and bias during the growth of the Cr-bearing films.

Hysteretic behavior was studied in a series of Fe thin films, grown by molecular beam epitaxy, having different grain sizes and grown on different substrates. Major and minor loops and first order reversal curves (FORCs) were collected to investigate magnetization mechanisms and domain behavior under different magnetic histories. The minor loop coefficient and major loop coercivity increase with decreasing grain size due to higher defect concentration resisting domain wall movement. First order reversal curves allowed estimation of the contribution of irreversible and reversible susceptibilities and switching field distribution. The differences in shape of the major loops and first order reversal curves are described using a classical Preisach model with distributions of hysterons of different switching fields, providing a powerful visualization tool to help understand the magnetization switching behavior of Fe films as manifested in various experimental magnetization measurements.

Dislocation imaging using transmission electron microscopy (TEM) has been an invaluable tool for characterizing crystallographic defects in metals. Compared to conventional TEM imaging, diffraction contrast imaging scanning transmission electron microscopy (DCI STEM) with appropriate setting can provide better defect contrast with almost negligible bend contour artifacts, enabling more effective analysis of dislocation structures. Here, we investigated why STEM can suppresses bend contour, and how dislocation contrast behaves along with different STEM imaging parameters. Using a body-centered cubic HT-9 ferritic/martensitic alloy as an example, a simple procedure and operational theory are described at the beginning to help set up DCI STEM experiments. Comparing with conventional TEM and the STEM strictly complying with the principle of reciprocity, we found that a pair of STEM convergence and collection semi-angles, αS and βS, a few milliradians in size is essential for bend-contour-free defect imaging. It works in concert such that the convergence STEM probe opens up the reciprocal space, and then a comparable collection region evens out the rocking-curve oscillation and alleviates bend contours from the reciprocal space. This fundamental advantage is unique in DCI STEM. Practical guidelines regarding STEM parameters and specimen orientation and thickness are then provided for DCI STEM dislocation imaging. Lastly, we show that coupling DCI STEM with spectrum images of electron energy loss spectroscopy and of energy-dispersive X-ray Spectroscopy offers a comprehensive characterization of crystallographic defects and chemical information of complex microstructures.

Single crystal Fe thin films (∼250 nm) were grown on MgO substrates and irradiated with 2.0 MeV Fe+ ions at 10 and 50 dpa, and the defect evolution was studied using high resolution Transmission Electron Microscopy (HR-TEM) and Doppler Broadening Positron Annihilation Spectroscopy (PAS). It was shown that irradiation induced or exacerbated a thin oxide layer at the outer interface and produced substantial Fe/Mg mixing at the film/substrate interface, particularly for the higher dose. Modeling of the PAS data allowed interpretation of the defect types at different distances from the Fe surface, and included several types of MgO substrate damage and annihilation condition changes, indicative of damage due to ballistic effects of the Fe atoms as well as chemical changes due to implantation and subsequent diffusion. This detailed PAS study compared with TEM and energy dispersive spectroscopy (EDS) provides significant insight into depth-dependent defect creation. These results will be useful for predicting defect creation in Fe-based materials under irradiation conditions, for extension to neutron irradiated structural materials.

Major pharmaceutical companies contribute important expertise to health research and development (R&D), particularly in their ability to develop and bring pharmaceuticals to market. The Access to Medicine Index evaluates how 20 of the world’s largest pharmaceutical companies are directing R&D efforts towards the needs of people living in low- and middle-income countries. In dissemination of its findings, the Index stimulates pharmaceutical companies to expand R&D activities in this direction. The Index methodology is reviewed every 2 years, most recently for the 2018 Index, to ensure their R&D activity is benchmarked against current health R&D priorities as defined by the global health community. The review is based on consensus-building processes involving global health stakeholders. In the latest review, two main changes to the methodology were made that will further deepen the Index’s analysis of (1) how far companies’ R&D activity aligns with global health priorities; and (2) whether companies make plans to ensure resulting innovations reach populations in need globally. These changes will be applied in the 2018 Access to Medicine Index. Importantly, the methodology review process highlighted the need for further prioritisation from the global health community, in particular to emphasise to innovators which product innovations are needed most critically to address the burden of non-communicable diseases in low- and middle-income countries. Should such prioritisations be developed, the Index can play an important role in tracking and stimulating company contributions towards them. Electronic supplementary material The online version of this article (10.1186/s12961-018-0332-y) contains supplementary material, which is available to authorized users.

Accelerator-based ion beam irradiation techniques have been used to study radiation effects in materials for decades. Although carbon contamination induced by ion beams in target materials is a well-known issue in some material systems, it has not been fully characterized nor quantified for studies in ferritic/martensitic (F/M) steels that are candidate materials for applications such as core structural components in advanced nuclear reactors. It is an especially important issue for this class of material because of the strong effect of carbon level on precipitate formation. In this paper, the ability to quantify carbon contamination using three common techniques, namely time-of-flight secondary ion mass spectroscopy (ToF-SIMS), atom probe tomography (APT), and transmission electron microscopy (TEM) is compared. Their effectiveness and shortcomings in determining carbon contamination are presented and discussed. The corresponding microstructural changes related to carbon contamination in ion irradiated F/M steels are also presented and briefly discussed.

Capacity and voltage fading of layered structured cathode based on lithium transition-metal oxide is closely related to the lattice position and migration behavior of the transition-metal ions. However, it is scarcely clear about the behavior of each of these transition-metal ions in this category of cathode material. We report direct atomic resolution visualization of interatomic layer mixing of transition metals (Ni, Co, Mn) and lithium ions in layered structured oxide cathodes for lithium-ion batteries. Using chemical imaging with an aberration-corrected scanning transmission electron microscope (STEM) and density function theory calculations, we discovered that, in the layered cathodes, Mn and Co tend to reside almost exclusively at the lattice site of transition-metal (TM) layer in the structure or little interlayer mixing with Li. In contrast, Ni shows a high degree of interlayer mixing with Li. The fraction of Ni ions resides in the Li layer followed by a near linear dependence on total Ni concentr...

Electroplating was used as a purification method and produced thick (3.2–12.2 mm) copper deposits of ultra-high radiopurity. Due to the extreme thickness of these electrodeposits compared to traditional electroplating, characterization is necessary to prevent costly failures and ensure device reliability. The deposition rate was carefully controlled to maintain a uniform growth front and required plating for a continuous 8 months in order to produce the 12.2-mm-thick copper specimen. Tensile testing shows the electroplated copper to exhibit significant strain hardening as would be expected with face-centered cubic materials, indicating that the material is free of significant defects and voids. Testing of eight tensile samples machined according to ASTM-E8 specifications exhibited yield strengths of 95 ± 4 MPa. Hardness was measured to be 79.8 ± 5.3 HV using a 200-gf load. Microstructure and deformation showed the grains to be highly aligned with respect to the growth direction, and electron backscatter diffraction showed the development of a (110) texture.

As a candidate material for fusion reactor applications, silicon carbide (SiC) undergoes transmutation reactions under high-energy neutron irradiation with magnesium as the major metallic transmutant; the others include aluminum, beryllium and phosphorus in addition to helium and hydrogen gaseous species. The impact of these transmutants on SiC structural stability is currently unknown. This study uses ion implantation to introduce Mg into SiC. Multiaxial ion-channeling analysis of the as-produced damage state suggests that there are preferred Si interstitial splits. The microstructure of the annealed sample was examined using high-resolution scanning transmission electron microscopy. The results show a high concentration of likely non-faulted tetrahedral voids and possible stacking fault tetrahedra near the damage peak. In addition to lattice distortion, dislocations and intrinsic and extrinsic stacking faults are also observed. Magnesium in 3C-SiC prefers to substitute for Si and it forms precipitates of cubic Mg2Si and tetragonal MgC2. The diffusion coefficient of Mg in 3C-SiC single crystal at 1573 K has been determined to be 3.8±0.4×10e-19 m2/sec.

There is increased interest in improved methods for in situ non-destructive interrogation of materials for nuclear reactors in order to ensure reactor safety and quantify material degradation (particularly embrittlement) prior to failure. Therefore, a prototypical ferritic/martensitic alloy, HT-9, of interest to the nuclear materials community was investigated to assess microstructure effects on micromagnetics measurements (Barkhausen noise emission, magnetic hysteresis measurements, and first order reversal curve analysis) for samples undergoing three different heat treatments. Microstructural and physical measurements consisted of high precision density, resonant ultrasound elastic constant, Vickers microhardness, grain size, and texture determination. These were varied in the HT-9 alloy samples and related to various magnetic signatures. In parallel, a mesoscale microstructure model was created for α-iron and the effects of polycrystallinity and the demagnetization factor were explored. It was observed that Barkhausen noise emission decreased with increasing hardness and decreasing grain size (lath spacing), while coercivity increased. The results are discussed in terms of the use of magnetic signatures for the non-destructive interrogation of radiation damage and other microstructural changes in ferritic/martensitic alloys.

In this study, we describe a methodology to examine dislocation loops in irradiated steels based on a combination of crystallographic information and g⋅b invisibility criteria. Dislocation loops in transmission electron microscope (TEM) images can be conveniently analyzed using this method. Through this analysis approach, dislocation loops in reduced activation ferritic/martensitic (RAFM) steels irradiated at 400 °C have been examined. The predominant types of loops found in irradiated RAFM steels were 〈1 0 0〉{2 0 0} and 1/2〈1 1 1〉{1 1 1}. The size, density, and density anisotropy of these two types of dislocation loops were quantified. It was observed that the 〈1 0 0〉{2 0 0} loop density is more than twice that of 1/2〈1 1 1〉{1 1 1} loops. A large density anisotropy of 〈1 0 0〉{2 0 0} loops was identified.

Formation of nanoscale helium (He) bubbles in reduced activation ferritic/martensitic steels may lead to degradation of mechanical properties of materials. Transmission electron microscopy (TEM) has commonly been used to image the Fresnel contrast of He bubbles, using an underfocus of 0.5-1 µm. This paper presents our study of multislice simulation of the size correlation between imaged Fresnel rings and the actual He bubbles. It was found that for bubbles equal to or >3 nm in diameter, the imaged bubble size, represented by its inner diameter of the first dark Fresnel ring (D(in)) in underfocused imaging conditions, increases with increasing electron-beam incoherency, but decreases with increasing underfocus. The electron-beam accelerating voltage, bubble size, bubble position and TEM sample thickness were found to have no significant influence on the deviation of D(in) from the actual bubble size (D(0)). However, for bubbles equal to or

A new capability has been developed for analyzing solid oxide fuel cells (SOFCs). This paper describes the initial results of in situ X-ray diffraction (XRD) of the cathode on an operating anode-supported solid oxide fuel cell. It has been demonstrated that XRD measurements of the cathode can be performed simultaneously with electrochemical measurements of cell performance or electrochemical impedance spectroscopy (EIS). While improvements to the technique are still to be made, the XRD pattern of a lanthanum strontium cobalt ferrite (LSCF) cathode with the composition La 0.6 Sr 0.4 Co 0.2 Fe 0.8 O 3− δ (LSCF-6428) was found to continually but gradually change over the course of more than 60 h of operation in air under typical SOFC operating conditions. It was determined that the most significant change was a gradual increase in the cubic lattice parameters of the LSCF from 3.92502 A (as determined from the integration of the first 20 h of XRD patterns) to 3.92650 A (from the integration of the last 20 h). This analysis also revealed that there were several peaks from unidentified minor phases that increased in intensity over this timeframe. After a temporary loss of airflow early in the test, the cell generated between 225 and 250 mW cm −2 for the remainder of the test. A large low frequency arc in the impedance spectra suggests the cell performance was gas diffusion limited and that there is room for improvement in air delivery to the cell.

Advanced characterization tools, such as electron backscatter diffraction and transmitted IR microscopy, are being applied to study critical microstructural features and orientation relations in as-grown CZT crystals to aid in understanding the relation between structure and properties in radiation detectors. Even carefully prepared single crystals of CZT contain regions of slight misorientation, Te-particles, and dislocation networks that must be understood for more accurate models of detector response. This paper describes initial research at PNNL into the hierarchy of microstructures observed in CZT grown via the vertical gradient freeze or vertical Bridgman method at PNNL and WSU.

High helium levels produced in fusion neutron spectra may lead to severe increases in the brittle fast fracture temperature, enhanced void swelling and degradation of creep rupture properties at lower, intermediate and higher irradiation temperatures, respectively. Thus it is important to develop structural alloys with stable microstructures that can manage helium based on understanding of its transport, fate and consequences. We report on the initial results of a study of helium in a nanostructured ferritic alloy (NFA), MA957, that is dispersion strengthened by an ultra-high density of nm-scale Y–Ti–O nanofeatures (NFs). An in situ helium implanter technique uniformly deposited ≈380 appm helium to ≈6 μm in MA957 irradiated in the High Flux Isotope Reactor (HFIR) to ≈9 dpa at 500 °C. Through focus transmission electron microscopy (TEM) showed that helium is in extremely fine bubbles that often appear to coincide with bright field features taken as a NF.

The interactions of antimony with the nickel–zirconia anode in solid oxide fuel cells (SOFCs) have been investigated. Tests with both anode-supported and electrolyte-supported button cells were performed at 700 and 800 °C in synthetic coal gas containing 100 ppb to 9 ppm antimony. Minor performance loss was observed immediately after Sb introduction to coal gas resulting in ca. 5% power output drop. While no further degradation was observed during the following several hundred hours of testing, cells abruptly and irreversibly failed after 800–1600 h depending on Sb concentration and test temperature. Antimony was found to interact strongly with nickel resulting in extensive alteration phase formation, consistent with expectations based on thermodynamic properties. Nickel antimonide phases, NiSb and Ni 5 Sb 2 , were partially coalesced into large grains and eventually affected electronic percolation through the anode support. Initial degradation was attributed to diffusion of antimony to the active anode/electrolyte interface to form an adsorption layer, while the late stage degradation was due the Ni–Sb phase formation. Assuming an average Sb concentration in coal gas of 0.07 ppmv, a 500 μm thick Ni/zirconia anode-supported cell is not expected to fail within 7 years when operated at a power output of 0.5 W cm −2 and fuel utilization above 50%.