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A common problem in analytical scanning electron microscopy (SEM) using electron backscatter diffraction (EBSD) is the differentiation of phases with distinct chemistry but the same or very similar crystal structure. X-ray energy dispersive spectroscopy (EDS) is useful to help differentiate these phases of similar crystal structures but different elemental makeups. However, open, automated, and unbiased methods of differentiating phases of similar EBSD responses based on their EDS response are lacking. This paper describes a simple data analytics-based method, using a combination of singular value decomposition and cluster analysis, to merge simultaneously acquired EDS + EBSD information and automatically determine phases from both their crystal and elemental data. I use hexagonal TiB2 ceramic contaminated with multiple crystallographically ambiguous but chemically distinct cubic phases to illustrate the method. Code, in the form of a Python 3 Jupyter Notebook, and the necessary data to replicate the analysis are provided as Supplementary material.

We report the dependence of mechanical properties and ionic conductivity of the Na1+xMnx/2Zr2–x/2(PO4)3 (designated as NMZP) ionic conductor on its composition (with x = 0.5, 1.0, 1.5, and 2.0). Lo...

Complex material systems in which microstructure and microchemistry are nonuniformly dispersed require three-dimensional (3D) rendering(s) to provide an accurate determination of the physio-chemical nature of the system. Current scanning transmission electron microscope (STEM)-based tomography techniques enable 3D visualization but can be time-consuming, so only select systems or regions are analyzed in this manner. Here, it is presented that through high-efficiency multidimensional STEM acquisition and reconstruction, complex point cloud-like microstructural features can quickly and effectively be reconstructed in 3D. The proposed set of techniques is demonstrated, analyzed, and verified for a high-chromium steel with heterogeneously situated features induced using high-energy neutron bombardment.

Statistically significant crystal structure and composition identification of nanocrystalline features such as nanoparticles/nanoprecipitates in materials chemistry and alloy designing using electron microscopy remains a grand challenge. In this paper, we reveal that differing crystallographic phases of nanoprecipitates in alloys can be mapped with unprecedented statistics using transmission Kikuchi diffraction (TKD), on typical carbon-based electron-transparent samples. Using a case of multiphase, multicomponent nanoprecipitates extracted from an improved version of 9% chromium Eurofer-97 reduced-activation ferritic-martensitic steel we show that TKD successfully identified more than thousand M23C6, MX, M7C3, and M2X (M=Fe, Cr, W, V, Ta; X = C, N) nanoprecipitates in a single scan, something that is currently unachievable using a transmission electron microscope (TEM) without incorporating a precision electron diffraction (PED) system. Precipitates as small as ∼20–25 nm were successfully phase identified by TKD. We verified the TKD phase identification using high-resolution transmission electron microscopy (HRTEM) and convergent beam electron diffraction (CBED) pattern analysis of a few precipitates that were identified by TKD on same sample. TKD study was combined with state-of-art analytical scanning transmission electron microscopy (STEM)-energy dispersive X-ray (EDX) spectroscopy and multivariate statistical analysis (MVSA) which provided the complete crystal structure and distinct chemistries of the precipitates in the steel in a high throughput automated way. This technique should be applicable to characterizing any multiphase crystalline nanoparticles or nanomaterials. The results highlight that combining phase identification by TKD with analytical STEM and modern data analytics may open new pathways in big data material characterization at nanoscale that may be highly beneficial for characterizing existing materials and in designing new materials.

Neutron irradiation in mixed spectrum reactors is an effective approach to investigate the synergistic effects of solid transmutations and atomic displacement damage in tungsten. In this article the findings and interpretations from a large-scale tungsten irradiation campaign conducted using the High Flux Isotope Reactor of Oak Ridge National Laboratory are summarized. The response of originally-unalloyed tungsten to mixed spectrum neutron irradiation is characterized by enormous hardening, degradation of modulus of toughness eventually leading to embrittlement, and decreased thermal conductivity. These property changes may be attributed primarily to production, segregation, and precipitation of rhenium and osmium and assisted by displacement damage. The stages describing the evolution of the microstructural development and property changes with the neutron dose and the accompanying accumulation of transmutation products are proposed. This manuscript has been co-authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan ).

Erosion characteristics of tungsten-alternative plasma-facing materials (PFMs) were tested under high heat flux conditions in the electrothermal plasma source facility at Oak Ridge National Laborat...

The United States and Japan have collaborated on fusion materials research in a series of agreements reaching back to 1981. The PHENIX collaboration is the latest U.S.-Japan project which s...

Tungsten is the material of choice as the plasma-facing material in future plasma-burning fusion reactors. During operation, plasma-facing materials will be simultaneously exposed to 14-MeV...

Plasma-facing materials in future large-scale fusion reactors must be designed to withstand high heat fluxes from extreme off-normal events such as edge localized modes and unmitigated plasma disruptions. The erosion rates of possible tungsten-alternative materials are tested under high heat flux conditions at the DIII-D National Fusion Facility. High-purity β-3C CVD silicon carbide was exposed alongside MAX phase ceramic Ti3SiC2 to both L- and H-mode plasma discharges in the DIII-D divertor. Samples survived average heat fluxes ranging from 2–10 MW/m2 over 16 s. A new micro-trench erosion measurement technique was successfully implemented and measured Ti3SiC2 and SiC erosion rates of 0–9 nm/s and 27–73 nm/s, respectively. Additionally, average ion impact angle estimates for an incident B-field angle of ∼1.5° from surface parallel were made using micro-trench impact patterns. Measurements ranged from θ = 24o–34o with respect to Bt and ϕ = 51.5o–55o below the surface normal.

To study the benefit of ~ 50-µm-thick air-plasma-sprayed (APS) ‘flash’ bond coatings, NiCoCrAlYHfSi high-velocity oxygen fuel (HVOF) bond coatings were deposited on alloy 247 disk substrates with APS yttria-stabilized zirconia topcoatings. Using 1-h cycles at 1100 °C in air with 10% H2O and HVOF-only bond coatings as a baseline, APS flash coatings extended the average coating lifetime by > 10% using NiCoCrAlYHfSi and > 70% using NiCoCrAlY powder. Coatings were characterized after 0, 100 and 300 cycles and after failure. Residual stress in the thermally grown alumina scale was mapped every 100 cycles using photo-stimulated luminescence piezospectroscopy. The flash coating created an interlocked metal–alumina layer that appeared to inhibit critical crack formation, and the underlying HVOF layer prevented further internal oxidation and appeared to supply the APS flash coating with Al. The addition of Hf and Y to the flash coating increased internal oxidation and accelerated Al consumption, thereby reducing the benefit of the flash coating.

Ultra-high temperature transition-metal ceramics are potential candidates for fusion reactor structural/plasma-facing components. We reveal the irradiation damage microstructural phenomena in Boron-11 enriched titanium diboride (TiB2) using mixed-spectrum neutron irradiations, combined with state-of-art characterization using transmission electron microscopy (TEM) and high resolution TEM (HRTEM). Irradiations were performed using High Flux Isotope Reactor at ∼220 and 620 °C up to 2.4 × 1025 n.m−2 (E > 0.1 MeV). Total dose including contribution from residual Boron-10 (10B) transmutation recoils, was ∼4.2 displacements per atom. TiB2 is susceptible to irradiation damage in terms of dislocation loop formation, cavities and anisotropic lattice parameter swelling induced micro-cracking. At both 220 and 620 °C, TEM revealed dislocation loops on basal and prism planes, with nearly two orders of magnitude higher number density of prism-plane loops. HRTEM, electron diffraction and relrod imaging revealed additional defects on { 10 1 ¯ 0 } prism planes, identified as faulted dislocation loops. High defect cluster density on prism planes explains anisotropic a-lattice parameter swelling of TiB2 reported in literature which caused grain boundary micro-cracking, the extent of which decreased with increasing irradiation temperature. Dominance of irradiation-induced defect clusters on prism planes in TiB2 is different than typical hexagonal ceramics where dislocation loops predominantly form on basal planes causing c-lattice parameter swelling, thereby revealing a potential role of c/a ratio on defect formation/aggregation. Helium generation and temperature rise from residual 10B transmutation caused matrix and grain boundary cavities for the irradiation at 620 °C. The study additionally signifies isotopic enrichment as a viable approach to produce transition-metal diborides for potential nuclear structural applications.

Transmutation-induced precipitation in neutron-irradiated tungsten is an important performance concern for its application as plasma facing material in fusion reactors. In this study, segregation and precipitation of transmutant elements in single crystal and polycrystal tungsten irradiated at 460–1100 °C to 0.02–2.4 displacements per atom (dpa) in the High Flux Isotope Reactor were investigated using transmission electron microscopy. The results indicated that nanoscale W-Re-Os clusters were identified in the low dose regime from 0.02 to 0.44 dpa with irradiation temperature lower than 800 °C while acicular-shape precipitates formed when irradiation dose is higher than 1.5 dpa. A tentative roadmap of the kinetics process of the transmutation-induced precipitation in neutron-irradiated tungsten is presented characterizing the defect features (i.e., W-Re-Os clusters and precipitates) consisting of transmutant elements in tungsten irradiated to various doses. All TEM-visible voids were associated with the acicular-shape precipitates. Voids were formed prior to the formation of acicular-shape precipitates and act as strong trapping sites for mobile species involved in the precipitation together with dislocations. Thermal stability of W-Re-Os clusters was assessed by performing a 2-h anneal at 1200 °C on tungsten irradiated to 0.44 dpa at 705 °C. The kinetics process of transmutant elements and radiation defects are discussed to reveal the underlying mechanisms controlling the formation of precipitates in tungsten.

The plasma-surface interactions from samples of high-conductivity graphitic foam biased to 120 V and placed in 6–8 eV deuterium plasmas with densities as high as 1019 m−3 were investigated at the PSI-2 linear plasma device in Jülich. Graphitic foam-plasma interactions were also studied at the Wendelstein 7-X (W7-X) stellarator in Greifswald by exposure to hydrogen and helium plasmas using the Jülich multi-purpose manipulator. The purpose was to explore the possibility of using the material in a plasma facing component, and initial results were encouraging. In W7-X, no measurable erosion or cracking was observed. The PSI-2 samples received a deuterium fluence of 5 × 1025 m−2 resulting in an average erosion of 43 µm or about 5 mg per sample. Residual gas analysis (RGA) data were acquired to monitor sample outgassing. Laser-induced Breakdown Spectroscopy (LIBS) was used to measure deuterium retention in the porous foam. After exposure, the surfaces were characterized with scanning electron microscopy, energy dispersive x-ray analysis and time-of-flight secondary ion mass spectrometry (ToF-SIMS). The graphitic foam has a thermal conductivity as high as 287 W/mK and is considered as a replacement to more exotic carbon fiber composites such as SepCarb NB31 or isotropic graphites like ATJ that are no longer produced but used in present-day tokamak experiments. Actively cooled monoblocks were made from the foam and underwent extensive materials characterization including infrared response studies at Oak Ridge National Laboratory. This material is under consideration for the proposed actively-cooled W7-X divertor scraper element. Keywords: Graphitic foam, Graphite, Plasma facing component, Plasma exposure, Monoblock

Thirteen austenitic stainless steels, nickel-base alloys, and ferritic alloys were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Comprehensive microstructural characterization was performed for irradiation-induced features, including dislocation loops, voids, precipitates, and radiation induced segregation (RIS). Dislocation loops formed in all alloys except 14YWT, while voids were observed in alloys 316 L, 310, C22, and 14YWT. Irradiation-induced formation of γ′ precipitates was observed in alloys 316 L, 310, 800, and 690; the irradiation-enhanced, long-range ordered Ni2Cr phase (Pt2Mo-type) was observed in alloys 690, C22, 625, 625Plus, 625DA, and 725; and G-phase was observed in alloy T92. No irradiation-induced precipitates were observed in alloys X750, 718 or 14YWT. Precipitation of the γ′ phase can be understood through segregation and clustering of Si, Al, and Ti. Overall, austenitic stainless steels are generally susceptible to irradiation damage in the form of loops, voids, precipitates, and RIS. Ni-base alloys have this same type of dislocation loops and RIS behaviors but are more resistant to void swelling. Ferritic alloys showed better resistance to loop formation, void swelling and irradiation-induced precipitation. From the degree of irradiation-induced microstructural change, alloy T92 was identified as the most radiation resistant among these alloys.

Tungsten is the primary candidate for plasma-facing materials (PFMs) in a magnetic fusion energy (MFE) devices such as ITER due to its high melting point, excellent erosion resistance, and low sputtering yield. However, tungsten will suffer from heat, neutron irradiation, large flux (1022-1024 He/m2s)- low energy (tens of eV to hundreds of eV) helium and hydrogen ion exposure, as well as microstructure evolution (such as recrystallization). We have exposed hot-rolled and recrystallized tungsten to 65 or 80 eV helium ions with a flux of 0.5 × 1022 or 5 × 1022 He/m2s to total fluence from 0.6 × 1024 to 4 × 1024 He/m2 at a temperature of ≈1100 K. The results show that recrystallized tungsten samples exhibit a roughened surface morphology in certain grains with orientations close to , while hot-rolled tungsten samples still maintain the smooth surface under all helium-ion exposure conditions. All the samples exhibit relatively shallow helium bubble penetration (

Modern electron microscopy techniques were used to conduct a thorough study of an irradiated urania fuel pellet microstructure to attempt at an understanding of high burnup structure formation in this material. The fuel was irradiated at low power to high burnups in a light water reactor, proving ideal for this purpose. Examination of grain size and orientation with strict spatial selectivity across the fuel pellet radius allowed for capturing the progression of the restructuring process, from its onset to full completion. Based on this information, the polygonization mechanism was shown to be responsible for restructuring, involving formation of low-angle grain boundaries with their initiation occurring at the original high-angle grain boundaries of the as-fabricated pellet and at the gas bubble-matrix interfaces. The low-angle character of boundaries between the subdivided grains disappeared in the fully developed high burnup structure, likely due to creep deformation in the pellet.

Tungsten is the primary candidate for plasma-facing materials in the critical regions of a magnetic fusion energy (MFE) device, such as a tokamak. However, tungsten (and many other metals) grow copious nanotendril "fuzz" when exposed to fusion-relevant He-bearing plasmas, and the impact of this fuzz on reactor operation is still unknown. Further, the mechanism of fuzz growth is also poorly understood at present. In order to experimentally probe the growth mechanisms, we developed a new sample preparation technique to examine the interface of individual nanotendril roots with the substrate (substrate/tendril interface) using electron microscopy and measured the grain boundary character distributions (GBCDs) of substrate/tendril interfaces, the GBs within individual tendrils (tendril/tendril interfaces), and GBs within the substrate (substrate/substrate interfaces). We find the substrate/tendril and tendril/tendril GBCDs are essentially indistinguishable, but very different from the substrate/substrate GBCDs, which implies that tendril growth periodically forms new GBs at the substrate/tendril interface, and these GBs are pushed upward to become tendril/tendril interfaces within the fuzz mat. These results will help guide future computational modelling studies to elucidate the details of the growth mechanisms.

Three types of microstructures, i.e., tempered-martensite (TM), ferrite (F), and dual-phase (TM + F), were developed in a castable nanostructured alloy that favors a high density of nanoprecipitates compared with the precipitates in current reduced-activation ferritic-martensitic steels. The effect of the distinct microstructures on tensile properties, Charpy impact toughness, and thermal helium desorption behavior was investigated with the full TM structure as a reference. The results indicated that the F domain in the TM + F structure governed the strength and slightly impaired the impact toughness. The full F structure exhibited the highest strength without compromising ductility, but it noticeably diminished impact toughness. All microstructures had a dominant helium desorption peak at ∼1070 °C. The higher density of nanoprecipitates and complex boundaries and dislocations in the TM + F structure enhanced the secondary helium desorption peak and extended the shoulder peak, in contrast to the full TM structure with an enlarged desorption peak associated with the ferrite-to-austenite transformation at ∼810–850 °C and the full F structure with a dominant desorption peak related to bubble migration at ∼1070 °C. These results suggest that components fabricated from functionally graded microstructures could be engineered to exploit the advantages of different microstructures for demanding application requirements.

Ultrasonic additive manufacturing (UAM) is a solid-state additive manufacturing process that uses fundamental principles of ultrasonic welding and sequential layering of tapes to fabricate complex three-dimensional (3-D) components. One of the factors limiting the use of this technology is the poor tensile strength along the z-axis. Recent work has demonstrated the improvement of the z-axis properties after post-processing treatments. The abnormally high stability of the grains at the interface during post-weld heat treatments is, however, not yet well understood. In this work we use multiscale characterization to understand the stability of the grains during post-weld heat treatments. Aluminum alloy (6061) builds, fabricated using ultrasonic additive manufacturing, were post-weld heat treated at 180, 330 and 580 °C. The grains close to the tape interfaces are stable during post-weld heat treatments at high temperatures (i.e., 580 °C). This is in contrast to rapid grain growth that takes place in the bulk. Transmission electron microscopy and atom-probe tomography display a significant enrichment of oxygen and magnesium near the stable interfaces. Based on the detailed characterization, two mechanisms are proposed and evaluated: nonequilibrium nano-dispersed oxides impeding the grain growth due to grain boundary pinning, or grain boundary segregation of magnesium and oxygen reducing the grain boundary energy.

Precipitates of the Laves and B2 phases were engineered in a newly-designed advanced ferritic alloy. Under creep test at 650 °C with 120 MPa, the material showed a steady-state minimum creep rate of 1 × 10−4 h−1, about one order of magnitude lower than T91. Microstructural characterization of the ferritic alloy revealed primarily ductile and partially brittle fractures after the creep test. Coarse Laves phase (~ 1 µm) was observed associating with the brittle fracture, resulting in reduced creep ductility. However, fine Laves phase precipitates (~ 100 nm) helped the dimple-ductile fracture and strengthened the material through impeding the motion of dislocations and boundaries. Unlike the B2 precipitates remained coherent exerting the classic Orowan bypassing mechanism at the brittle location, some of the B2 precipitates at the ductile location became incoherent and can develop an attractive interaction with dislocations. This coherency change of B2 precipitates, together with the nucleation of ultrafine (~ 40 nm) Laves phase precipitates during the creep test, would compensate for the coarsening-induced loss of Orowan strengthening of coherent B2 precipitates.

New candidate materials for GenIV or fusion nuclear energy systems, e.g., nanostructured ferritic alloys, are distinguished from older-generation nuclear materials by much smaller feature sizes and complex local nanochemistry and crystallography. Established and perspective nuclear materials, e.g. reactor pressure vessel steels or plasma-facing tungsten, also form small nanoscale structures under in-reactor service. Here, we discuss recent advances in materials characterization – high-efficiency X-ray mapping combined with datamining; transmission Kikuchi diffraction; and atom probe tomography – that make it possible to quantitatively characterize these nanoscale structures in unprecedented detail, which enables advances in understanding and modelling of radiation service and degradation.

Nano-scale Ni-Mn-Si-rich precipitates formed in a reactor pressure vessel steel under high neutron fluence have been characterized using highly complimentary atom probe tomography (APT) and scanning transmission electron microscopy with energy dispersive spectroscopy (STEM-EDS) combined with STEM-EDS modeling. Using these techniques in a synergistic manner to overcome the well-known trajectory aberrations in APT data, the average upper limit Fe concentration within the precipitates was found to be ∼6 at.%. Using this knowledge, accurate compositions of the precipitates was determined and it was found that the spread of precipitate compositions was large, but mostly centered around the Γ2-and G-phases. The use of STEM-EDS also allowed for larger areas to be examined, and segregation of minor solutes was observed to occur on grain boundaries, along with Ni-Mn-Si-rich precipitates that were smaller in size than those in the matrix. Solute segregation at the grain boundaries is proposed to occur through a radiation induced segregation or radiation enhanced diffusion mechanism due to the presence of a denuded zone about the grain boundary. It is also proposed that the reduced precipitate size at the grain boundaries is due to the structure of the grain boundary. The lack of Ni-Mn-Si precipitates observed in larger Mo-rich precipitates is also discussed, and the absence of the minor solutes required to form the Ni-Mn-Si precipitates results in the lack of nucleation. This is in contrast to cementite phases in which Ni-Mn-Si precipitates have been observed to have formed. It was also determined through this work that the exclusion of all the Fe ions during atom probe analysis is a reasonable approximation.

We report measurements of He-ion-beam induced tungsten nanofuzz formation for normal and non-normal incidence angles in the energy range 218 eV–10 keV. At 218 eV, the fuzz tendrils are fine and grow randomly away from the interface in the direction of the surface normal. Above 480 eV, the fuzz tendrils become increasingly coarser, and their growth direction is in the direction of the incident beam. This change is attributed to the ion-induced displacement damage which becomes effective once the displacement damage threshold energy is exceeded, and produces additional near-surface trapping sites in those portions of the surface that are in direct line of sight of the incident beam which can nucleate He clusters and initiate bubble growth. Once the surface morphology roughens sufficiently for shadowing to occur, the subsequent fuzz growth occurs preferentially toward the incident ion beam. Molecular dynamics (MD) simulations were carried out to determine the displacement damage threshold energies in the near-surface region along the three major crystallographic directions. It was found that the tungsten bulk values are established within the first 2–4 atomic layers below the tungsten surface. SRIM simulations based on the MD energy thresholds indicate that vacancy damage production in the near-surface region quickly dominates over sputtering in near-surface lattice modification effects as the energy above the damage threshold increases.

A compact electron cyclotron resonance plasma source has been utilized at Sandia National Laboratory to expose heated W samples (1270 K) to 50–75 eV He ions at fluxes on the order of 1019 m−2 s−1 a...

This work reports the first mechanical properties of Ti 3 AlC 2 -Ti 5 Al 2 C 3 materials neutron irradiated at ∼400, 630 and 700 °C at a fluence of 2 × 10 25 n m −2 (E > 0.1 MeV) or a displacement dose of ∼2 dpa. After irradiation at ∼400 °C, anisotropic swelling and loss of 90% flexural strength was observed. After irradiation at ∼630–700 °C, properties were unchanged. Microcracking and kinking-delamination had occurred during irradiation at ∼630–700 °C. Further examination showed no cavities in Ti 3 AlC 2 after irradiation at ∼630 °C, and MX and A lamellae were preserved. However, disturbance of (0004) reflections corresponding to M-A layers was observed, and the number density of line/planar defects was ∼10 23 m −3 of size 5–10 nm. HAADF identified these defects as antisite Ti Al atoms. Ti 3 AlC 2 -Ti 5 Al 2 C 3 shows abrupt dynamic recovery of A-layers from ∼630 °C, but a higher temperature appears necessary for full recovery.

Nano-infiltration and transient eutectic phase (NITE) sintering was developed for fabrication of nuclear grade SiC composites. We produced monolithic SiC ceramics using NITE sintering, as candidates for accident-tolerant fuels in light-water reactors (LWRs). In this work, we exposed three different NITE chemistries (yttria-alumina [YA], ceria-zirconia-alumina [CZA], and yttria-zirconia-alumina [YZA]) to autoclave conditions simulating LWR coolant loops. The YZA was most corrosion resistant, followed by CZA, with YA being worst. High-resolution elemental analysis using scanning transmission electron microscopy (STEM) X-ray mapping combined with multivariate statistical analysis (MVSA) datamining helped explain the differences in corrosion. YA-NITE lost all Al from the corroded region and the ytttria reformed into blocky precipitates. The CZA material lost all Al from the corroded area, and the YZA − which suffered the least corrosion −retained some Al in the corroded region. The results indicate that the YZA-NITE SiC is most resistant to hydrothermal corrosion in the LWR environment.

We observed that β-SiC, neutron irradiated to 9 dpa (displacements per atom) at ≈1440 °C, began transforming to α-SiC, with radiation-induced Frank dislocation loops serving as the apparent nucleation sites. 1440 °C is a far lower temperature than usual β → α phase transformations in SiC. SiC is considered for applications in advanced nuclear systems, as well as for electronic or spintronic applications requiring ion irradiation processing. β-SiC, preferred for nuclear applications, is metastable and undergoes a phase transformation at high temperatures (typically 2000 °C and above). Nuclear reactor concepts are not expected to reach the very high temperatures for thermal transformation. However, our results indicate incipient β → α phase transformation, in the form of small (~5–10 nm) pockets of α-SiC forming in the β matrix. In service transformation could degrade structural stability and fuel integrity for SiC-based materials operated in this regime. However, engineering this transformation deliberately using ion irradiation could enable new electronic applications.

In a magnetic fusion energy (MFE) device, the plasma-facing materials (PFMs) will be subjected to tremendous fluxes of ions, heat, and neutrons. The response of PFMs to the fusion environment is still not well defined. Tungsten metal is the present candidate of choice for PFM applications such as the divertor in ITER. However, tungsten's microstructure will evolve in service, possibly to include recrystallization. How tungsten's response to plasma exposure evolves with changes in microstructure is presently unknown. In this work, we have exposed hot-worked and recrystallized tungsten to an 80 eV helium ion beam at a temperature of 900 °C to fluences of 2 × 1023 or 20 × 1023 He/m2. This resulted in a faceted surface structure at the lower fluence or short but well-developed nanofuzz structure at the higher fluence. There was little difference in the hot-rolled or recrystallized material's near-surface (≤50 nm) bubbles at either fluence. At higher fluence and deeper depth, the bubble populations of the hot-rolled and recrystallized were different, the recrystallized being larger and deeper. This may explain previous high-fluence results showing pronounced differences in recrystallized material. The deeper penetration in recrystallized material also implies that grain boundaries are traps, rather than high-diffusivity paths.

A reduced activation ferritic/martensitic steel, Eurofer97, was neutron irradiated in the vicinity of 300 °C in the High Flux Isotope Reactor (HFIR) up to 72 dpa. Advanced analytical scanning transmission electron microscopy and conventional transmission electron microscopy were applied to investigate the radiation-induced segregation and phase instability behavior after neutron irradiation. Amorphization was observed in M23C6 carbides. Cr-rich clusters were seen within the matrix, near the lath boundaries and close to the M23C6 carbides. Cr enrichment and Fe depletion were detected at both prior austenite grain boundaries and lath boundaries, despite different segregation magnitude. In addition, the enrichment of Ni, the depletion of V, and tiny cavities (presumably helium bubbles) are also found at lath boundaries. This work interrogates the evolution of microstructures after neutron irradiation, which provides detailed understanding on the microstructural aspects controlling the mechanical integrity of Eurofer97 under high-dose neutron damage.

Graphite is the moderator material of several Generation IV nuclear reactor concepts, as well as the British Advanced Gas-cooled Reactors (AGR). Porosity can heavily influence the material properties, mechanical irradiation response, and neutron induced shrinkage or swelling of nuclear-grade graphite. Due to the sub-micron size of several types of pores found in graphite, only a high-resolution imaging technique such as electron tomography are capable of visualizing these features in three dimensions. In this research, we used electron tomography to characterize as-received and neutron irradiated samples of IG-110 nuclear-grade graphite to show for the first time the 3D structure of both native and irradiation-induced nano-cracks. This technique also reveals unique characteristics of graphite such as the structure that surrounds pores and could be used to inform molecular dynamic simulations of irradiated graphite and experimental techniques such as gas-absorption. This research also shows the utility of this technique for the study of other nuclear porous carbon-based materials.

Understanding the effects of helium on microstructures and mechanical properties of reduced-activation ferritic-martensitic steels is important to use of these steels in fusion reactor structures. The 9Cr-2WVTa steels were doped with 58Ni and 60Ni isotopes at 2 weight percent to control the rate of transmutation helium generation. The samples were irradiated in the High Flux Isotope Reactor to ~24 displacements per atom at nominal temperatures of 300, 400, and 500°C, producing 228 and 7 atomic parts-per-million helium in the 58Ni- and 60Ni-doped samples, respectively. Transmission electron microscopy revealed a variety of precipitates and the radiation-induced dislocation loops and cavities (voids or helium bubbles). Tensile tests of the irradiated samples at the irradiation temperatures showed radiation-induced hardening at 300°C and radiation-induced softening at 400°C. Analysis indicates that the hardening primarily originated from the loops and cavities. The 58Ni-doped samples had greater strengthening contributions from loops and cavities, leading to higher hardening with lower ductility than the 60Ni-doped samples. The greater helium production of 58Ni did not show pronounced reductions in ductility of the samples.

We grew nanotendril “fuzz” on tungsten via plasma exposure and performed transmission Kikuchi diffraction (tKD) in scanning electron microscopy of isolated nanotendrils. 900 °C, 10 23 He/m 2 sec, 4 × 10 26 He/m 2 exposure of tungsten produced a deep and fully developed nanotendril mat. tKD of isolated nanotendrils indicated that there was no preferred crystallographic direction oriented along the long axes of the tendrils, and the grain boundary character showed slightly preferential orientations. Tendril growth is sufficiently non-equilibrium to prevent any preference of growth direction to manifest measurably, and that new high-angle boundaries (with new grains and grain-growth axes) nucleate randomly along the tendrils during growth.

Deformation twinning plays a critical role on improving metals or alloys ductility, especially for hexagonal close-packed materials with low symmetry crystal structure. A rolled Mg alloy was selected as a model system to investigate the extension twinning behaviors and characteristics of parent-twin interactions by nondestructive in situ 3D synchrotron X-ray microbeam diffraction. Besides twinning-detwinning process, the “twinning-like” lattice reorientation process was captured within an individual grain inside a bulk material during the strain reversal. The distributions of parent, twin, and reorientated grains and sub-micron level strain variation across the twin boundary are revealed. A theoretical calculation of the lattice strain confirms that the internal strain distribution in parent and twinned grains correlates with the experimental setup, grain orientation of parent, twin, and surrounding grains, as well as the strain path changes. The study suggests a novel deformation mechanism within the hexagonal close-packed structure that cannot be determined from surface-based characterization methods.

Upsetting tests on two newly developed Mn and Cr-containing Al-Mg-Si-Cu alloys with various Mn contents were carried out at a speed of 15 mm/s under upsetting temperature of 450 °C after casting and subsequent homogenization heat treatment using a 300-Tone hydraulic press. STEM experiments revealed that Mn and Cr-containing α-Al (MnCrFe)Si dispersoids formed during homogenization showed a strong pinning effect on dislocations and grain boundaries, which could effectively inhibit recovery and recrystallization during hot deformation in the two alloys. Recrystallization fractions after solution heat treatment following hot deformation were measured by EBSD technique. It was found that the recrystallization fractions of the two alloys were less than 30%, giving rise to lower recrystallization fraction in the alloy with higher amount of Mn, which had higher number density of dispersoids. This implied that the finely distributed α-dispersoids were rather stable against coarsening and they stabilized the microstructure by inhibiting dislocation recovery and recrystallization during elevated temperature exposure. Increasing the content of Mn could increase the number density as well as the aspect ratio of the dispersoids, and more significantly, the effect of retardation on recrystallization were further enhanced.

Surface morphologies of tungsten surfaces, both polycrystalline and single-crystal [1 1 0], were investigated using SEM and FIB/SEM techniques after implantations at elevated surfaces temperatures (1200–1300 K) using well-characterized, mono-energetic He ion beams with a wide range of ion energies (218 eV–250 keV). Nanofuzz was observed on polycrystalline tungsten (PCW) following implantation of 100-keV He ions at a flux threshold of 0.9 × 1016 cm−2 s−1, but not following 200-keV implantations with similar fluxes. No nanofuzz formation was observed on single-crystal [1 1 0] tungsten (SCW), despite fluxes exceeding those demonstrated previously to produce nanofuzz on polycrystalline tungsten. Pre-damaging the single-crystal tungsten with implanted C impurity interstitials did not significantly affect the surface morphologies resulting from the high-flux He ion implantations. The main factor leading to the different observed surface structures for the pristine and C-implanted single-crystal W samples appeared to be the peak He ion flux characterizing the different exposures. It was speculated that nanofuzz formation was not observed for any SCW target exposures because of increased incubation fluences required for such targets.

Ultrasonic additive manufacturing (UAM) is a solid-state process, which uses ultrasonic vibrations at 20 kHz along with mechanized tape layering and intermittent milling operation, to build fully functional three-dimensional parts. In the literature, UAM builds made with low power (1.5 kW) exhibited poor tensile properties when loaded along the Z-direction, i.e., normal to the interfaces. This reduction in properties is often attributed to the lack of bonding at the interfaces. The generality of this conclusion is evaluated further in 6061 aluminum alloy builds made with very high power UAM (9 kW). Tensile deformation behavior along X and Z directions were evaluated with small-scale in-situ mechanical testing equipped with high-resolution digital image correlation, as well as, multi-scale characterization of builds. Interestingly, even with complete metallurgical bonding across the interfaces without any discernable voids, poor Z-direction properties were observed. This reduction is correlated to coalescence of pre-existing shear bands at interfaces into micro voids, leading to strain localization and spontaneous failure on tensile loading.