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Laser-assisted atom probe tomography (APT) was successfully applied to analyze the elemental distributions in a high-k/metal gate stack of a commercially available 45 nm node real device. APT revealed the multilayer structure of the high-k/metal gate stack with nearly atomic-scale resolution, and successfully detected small amounts of Zr in the thin layer of high-k HfO2 dielectrics and H in the Ti layer of the metal gate. The present results demonstrate the usefulness of APT as a tool of elemental analysis in nanoscale multilayer device structures. [ABSTRACT FROM AUTHOR]

Cu alloys with high durability and high thermal and electric conductivities are the key to achieving a more energetically efficient and eco-friendly future. Noble Zr-added oxide-dispersion-strengthened Cu alloys (ODS-Cu) with high strengths are promising materials for meeting all future service requirements. In this study, three-dimensional atom probe (3D-AP) and transmission electron microscopy (TEM) revealed that the majority of nano-oxide particles in the Zr-added ODS-Cu were novel yttria-stabilized zirconia (YSZ) particles with a diameter of less than 4 nm, which were completely different from Y 2 Zr 2 O 7 and Y 4 Zr 3 O 12 observed in other Zr-added ODS alloys. Multiscale observations including small-angle X-ray scattering (SAXS), 3D-AP, and TEM were performed to clarify the morphology of the oxide particles in the Zr-added ODS-Cu. Further strength enhancement through microstructural homogenization and specific incremental strength values were proposed. • Multiscale observation of oxide particles in Zr-added ODS-Cu was performed using small-angle X-ray scattering, 3D-AP and TEM. • Novel ultrafine YSZ nanoparticles with a diameter of less than 4 nm were confirmed to be mainly dispersed in the ODS-Cu. • Further strength enhancement and the specific incremental value of the ODS-Cu were predicted from the multiscale observation. [ABSTRACT FROM AUTHOR]

Three-dimensional distribution of arsenic (As) dopants at Σ3{111}, Σ9{221}, Σ9{114}, and Σ9{111}/{115} grain boundaries (GBs) in silicon (Si) is examined by correlative analytical methods using atom probe tomography (APT) combined with low-temperature focused ion beam (LT-FIB), scanning transmission electron microscopy, and ab initio calculations. Σ3{111} GBs, consisting of only 6-membered rings with small bond distortions, do not exhibit an apparent As segregation. Meanwhile, it is hypothesized that As atoms would segregate at 5-membered rings in the other GBs via anisotropic bond distortions spontaneously introduced so as to lower the donor level, as Jahn-Teller distortions. In addition, APT combined with LT-FIB suggests preferential As segregation around stretched 1 1 ‾ 0 bonds reconstructed in Σ9{114} and Σ9{111}/{115} GBs, that are inevitably introduced in the 1 1 ‾ 0 tilt GBs with the tilt angle larger than 70.5°. It is hypothesized that As atoms would form As dimers at stretched 1 1 ‾ 0 bonds and the adjacent ⟨ 111 ⟩ bonds, due to the tendency of As with five valence electrons to form a three-coordinated configuration, which is efficiently attained by an As dimer of a long length. This work provides important insights into As segregation at GBs; it is mainly determined by electronic interactions depending on the characteristics of valence electrons of As atoms, as well as on local bond distortions at GBs, via anisotropic bond distortions and dimerization. [ABSTRACT FROM AUTHOR]

We have investigated a post-annealing effect on microscopic structural changes of kink boundaries (KBs), which have been initially introduced by a hot-extrusion of an Mg-Zn-Y alloy containing long period stacking/order (LPSO) phase, based on scanning transmission electron microscopy (STEM) and atom-probe tomography (APT). STEM observations show that most of the KBs appear to be segmented into several submicron-scale, each of which is composed of arrays of extended basal (a -) dislocations that are geometrically necessary to account for the LPSO crystalline rotations. Atomic-scale STEM/APT observations reveal significant reconstructions of these extended dislocation-core structures, which occur at the solute-enriched face-centered cubic layers in the LPSO structures and forms local hexagonal close-packed Mg region with distinct solute depletions. These recovery-like features indeed enhance the thermal stability of the KBs, and the corresponding driving force can be reasonably described according to disclinations, whose elastic energy is defined by crystal rotations and the relevant dipole distances. • Microscopic structures of kink boundaries in unique long-period structured Mg-Zn-Y alloys have been investigated. • Microscopic segmentations of kink boundaries have been promoted during the post-annealing of the hot-extruded specimen. • Dislocation-core structures at kink boundaries have been significantly reconstructed with distinct solute redistributions. • Driving force to promote the observed recovery features can be reasonably described according to disclinations. [ABSTRACT FROM AUTHOR]

• Trace Si in RAFM steel contributes to suppressing neutron irradiation hardening. • Trace Si helps RAFM steel maintain a preferable elongation after irradiation. • Trace Si increases the area of ductile fracture in the neutron-irradiated steel. • Trace Si reduces the density of dislocation loops induced by neutron irradiation. • Retaining the trace Si is economical and sensible in RAFM steel design. The content of silicon (Si) element in reduced activation ferritic/martensitic (RAFM) steels varies significantly among different candidates of fusion reactor structural materials. Si is not intentionally added to the RAFM steel but rather remains as a residue during the deoxidation process in steel fabrication. To control and reduce the Si content, removing processes and special treatments are required. The additional refining processes significantly increase the cost of steel fabrication. Currently, the role of residual or trace Si in neutron irradiation of steel remains unclear. In the presented study, two 9Cr-2 W RAFM steels with Si contents of 0.1 wt.% and 0.01 wt.% were fabricated and subjected to a neutron irradiation at 563 K with a neutron fluence of 4.8 × 1023 m −2. The main results show that 0.1 wt.% Si can contribute to the suppression of irradiation hardening and embrittlement, as evidenced by the reduction in irradiation-increased hardness and yield strength, while maintaining preferable elongation. Microscopic analysis suggests that an appropriate content of Si helps refine the grains in RAFM steel, increases the ductile fracture region of the fracture surface, and effectively reduces the density of dislocation loops induced by neutron irradiation. The findings evidence that retaining a certain amount of trace Si should be an informed and rational choice in the design of RAFM steel. [ABSTRACT FROM AUTHOR]

The oxide dispersion strengthened (ODS)-Cu alloys produced by the combined process of mechanical alloying and hot isothermal pressing (MA-HIP) process have high strength properties, although they have low ductility. In this study, the MA-HIP process was modified based on the yttrium addition and oxygen addition methods to improve the ductility by controlling the microstructure of the ODS-Cu alloys. Namely, we replaced pure Y and Cu 2 O powders with a Cu-Y intermetallic compound and pre-heated Cu powders as the starting materials. The final product showed considerable microstructural differences: on the macro-scale, the frequency of interconnection of large/small Cu grain colonies, which was a main factor in brittle fracture, was reduced; while on the nanoscale, yttrium remained as yttria (Y 2 O 3) particles within the large size Cu grains. These multiscale microstructural changes resulted in the improved mechanical property: the newly-developed ODS-Cu bulk did not fracture at bending strains of 18% or greater, which can be compared with the previous value of 5% or less in a three-point bending test. Furthermore, the conductivities of our ODS-Cu alloys are 94–96% of pure Cu. This improvement in properties, therefore sheds light to the future application of the ODS-Cu alloys in high-temperature environments. • A newly elemental addition method for ODS-Cu alloys was used to improve the microstructure and mechanical properties. • The new ODS-Cu had improved ductility due to microstructural changes and did not fracture at bending strains above 18%. • The electric conductivities of a new ODS-Cu alloys are 94–96% of pure Cu. • The Y 2 O 3 particles were dispersed within the coarse Cu grains, increasing the maximum strength. [ABSTRACT FROM AUTHOR]

We investigated the fluence dependence of irradiation-induced solute cluster, dislocation loop, and very small defect to reveal the hardening mechanism in surveillance test specimens from a reactor pressure vessel steel with low-Cu content (0.04 wt%) using atom probe tomography (APT), weak-beam scanning transmission electron microscopy (WB-STEM), and positron annihilation spectroscopy. A high number density (>10 23  m −3 ) of solute clusters mainly composed of Ni, Mn, and Si atoms were found in highly neutron irradiated specimens (∼10 24 neutrons m −2 ( E  > 1 MeV)) by APT. These solute clusters were one of the main sources of hardening as reported previously. On the other hand, it was also revealed that dislocation loops were formed with a number density of ∼10 22  m −3 in the high-fluence specimens by WB-STEM. The estimated hardening due to dislocation loops was more than half of the actual hardening, showing that dislocation loops are also main source of irradiation hardening at high neutron fluence with the solid experimental evidences. Regarding specimens subjected to a low neutron fluence (∼10 23 neutrons m −2 ), very small defects, not detected by either WB-STEM or APT, were formed by positron annihilation spectroscopy. This result suggested that, at a low neutron fluence, the defects were the initial hardening source and they may grow the dislocation loops observed by WB-STEM at high fluence range. [ABSTRACT FROM AUTHOR]

To clarify the irradiation hardening behavior of neutron-irradiated tungsten(W), the formation behavior of rhenium(Re)-enriched cluster and W-Re precipitation with displacement damage in fission reactors environment was investigated by the atom probe tomography (APT) technique, and the results obtained by APT and microstructural observation by transmission electron microscopy (TEM) were compared. Neutron-irradiated pure W and W-10Re alloys were used for the analysis. These samples were irradiated to 0.96 dpa at 538 °C in a fast experimental reactor Joyo, and irradiated to 0.90 dpa at 500 °C in a mixed spectrum reactor HFIR. The chemical compositions of the matrix, Re clusters, and precipitates were determined by the AP analysis. The Joyo-irradiated samples contained lower levels of transmuted Re and Os, while the HFIR-irradiated samples contained higher levels of Re and Os. The Re-enriched clusters demonstrated characteristic spatial distribution related to defect cluster distribution in each irradiated specimen. [ABSTRACT FROM AUTHOR]

Abstract The composition evolution of γ′ nanoparticles as a function of aging duration in the Ti-doped CoFeCrNi high entropy alloy was studied by atom probe tomography (APT) and first-principle calculations. The APT results indicate that the evolution behaviors of Co, Fe, and Cr in the Ni 3 Ti-type γ′ nanoparticles are significantly different. Co saturates to a stable concentration in the γ′ nanoparticles rapidly in the initial stage of the precipitation, whereas Fe and Cr atoms are incorporated into the γ′ nanoparticles during the coarsening. The APT results are supported by the first-principle calculations from the perspective of the electronic state. Graphical abstract Unlabelled Image [ABSTRACT FROM AUTHOR]

[Display omitted] • The addition of Re and Cr suppressed the formation of vacancy clusters in W after high-temperature irradiation, while no effect was observed for W-5 at.% Ta and W-2.5 at.% Mo alloys. • Fine Cr-rich and Re-rich precipitates were observed in W-0.3 at.% Cr and W-5 at.% Re alloys, respectively, where the distributions of Mo and Ta in W-2.5 at.% Mo and W-5 at.% Ta alloys were uniform even after irradiation. • Irradiation hardening was observed for all materials but that of W-5 at.% Re alloy was significantly smaller than the hardening of W, W-2.5 at.% Mo and W-5 at.% Ta alloys. • The stability of mixed dumbbells plays an important role in the vacancy cluster formation, precipitation and irradiation hardening of W alloys. Irradiation responses of binary W alloys were investigated systematically from the perspective of the binding energy of an alloying element with a W self-interstitial atom (W-SIA). Plates of W, W-0.3 at.% Cr, W-5 at.% Re, W-2.5 at.% Mo and W-5 at.% Ta alloys were irradiated at 1073 K with 6.4 MeV Fe ions to 0.26 dpa at the damage peak, where the binding energy of alloying element with W-SIA is in order of Cr > Re > Mo > Ta. The formation of vacancy-type defects (vacancies and vacancy clusters) was studied by using positron lifetime measurement. The precipitation of alloying elements was studied by using atom probe tomography (APT) and the hardness changes in the irradiated volumes were measured by the nanoindentation technique. The formation of vacancy-type defects was strongly suppressed by the addition of Cr and Re, while Ta and Mo had no noticeable suppression effect. The APT measurements showed fine Cr- and Re-rich precipitates in W-0.3 at.% Cr and W-5 at.% Re alloys, respectively, where the density of precipitates in the latter was clearly lower than that in the former. The distributions of Mo and Ta were uniform even after irradiation. Irradiation hardening was observed for all materials but that of W-5 at.% Re alloy was significantly smaller than the hardening of W, W-2.5 at.% Mo and W-5 at.% Ta alloys. These observations suggest that the irradiation hardening of W, W-2.5 at.% Mo, and W-5 at.% Ta alloys were mainly caused by vacancy-type defects. It was concluded that an alloying element with moderate binding energy with a W-SIA effectively suppresses vacancy formation without significantly enhanced precipitation and consequently mitigates irradiation hardening. [ABSTRACT FROM AUTHOR]

• The role of Si content on irradiation embrittlement was clarified. • Solute atom clusters of highly neutron-irradiated Japanese pressurized water reactor surveillance test specimens were analyzed by atom probe tomography. • The cluster radius and number density decreased and increased, respectively, with increasing Si content, resulting in a constant volume fraction of SCs. • Increase in Si reduces the degree of irradiation embrittlement. This is consistent with the results of our previous study based on Bayesian statistical analysis. Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 × 1020 n/cm2 were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Mn, Ni, and Si atoms like a core-shell structure. In low-Cu bearing materials, Mn, Ni, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius (r) and number density (N d) decreased and increased with increasing nominal Si content, respectively. The shift in the reference temperature for nil-ductility transition (ΔRT NDT) showed a good correlation with the square root of volume fraction (V f) multiplied by ( V f × r). The negative relation between the nominal Si content and ΔRT NDT indicated that increasing of nominal Si content reduces the degree of embrittlement. [ABSTRACT FROM AUTHOR]

• The powders after MA consisted of large- and small-Cu grain layers due to inhomogeneous distribution of yttrium atoms. • Areal fraction of small Cu grain layer increases with increasing MA duration. • The powders after ball-milled have exothermic reaction at around 310°C, which may be recrystallization and grain growth of Cu grains. • During heat treatment above 300°C after MA, yttrium atoms are diffused into the outer peripheral region and forms a new small-Cu grain area. Oxide dispersion strengthened Cu alloys (ODS-Cu) having dispersed yttria (Y 2 O 3) fabricated using hot isostatic pressing (HIP) of mechanically alloyed (MA) Cu-Y powders. They exhibited a heterogeneous mixed grain structure, which were composed of a large and small Cu grain area. In the small Cu grain regions, Y₂O₃ particles were segregated along the grain boundaries, and acted as a crack propagation path, leading to brittle fracture. Therefore, we investigated the microstructural changes by the MA and heat treatment processes for the microstructure optimization. The powders of 4 to 32 MA hours formed a nested crust-like layered structure with large and small Cu grains. The layers were destroyed by heat treatment at 500°C, because recrystallization and grain growth of Cu occurred in the temperature range of 310 - 330°C. It was considered that Y 2 O 3 was moved out to the circumferential area of the powders. Based on these results, it was suggested that the structural changes inside MA powders during HIP are the essential factor of the Y 2 O 3 distribution within the Y 2 O 3 -added ODS-Cu alloys. [ABSTRACT FROM AUTHOR]