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Austenitic stainless steel D9 is a candidate for Generation IV nuclear reactor structural materials due to its enhanced irradiation tolerance and high-temperature creep strength compared to conventional 300-series stainless steels. But, like other austenitic steels, D9 is susceptible to irradiation-induced clustering of Ni and Si, the mechanism for which is not well understood. This study utilizes atom probe tomography (APT) to characterize the chemistry and morphology of Ni–Si nanoclusters in D9 following neutron or proton irradiation to doses ranging from 5–9 displacements per atom (dpa) and temperatures ranging from 430–683 °C. Nanoclusters form only after neutron irradiation and exhibit classical coarsening with increasing dose and temperature. The nanoclusters have Ni3Si stoichiometry in a Ni core–Si shell structure. This core–shell structure provides insight into a potentially unique nucleation and growth mechanism—nanocluster cores may nucleate through local, spinodal-like compositional fluctuations in Ni, with subsequent growth driven by rapid Si diffusion. This study underscores how APT can shed light on an unusual irradiation-induced nanocluster nucleation mechanism active in the ubiquitous class of austenitic stainless steels.

Growth of wurtzite ScxAl1−xN (x 0.18Al0.82N/GaN multi-quantum wells grown under these conditions display strong and narrow near-infrared intersubband absorption lines that confirm advantageous optical and electronic properties.

A series of Fe-3 to 18 at.% Cr binary ferritic alloys were neutron irradiated side by side, in the Advanced Test Reactor (ATR) at a temperature of ∼320 °C to a dose of ∼1.8 dpa. Three types of features that form under irradiation are relevant: (i) solute clusters; (ii) α′ precipitates; and, (iii) dislocation loops. The size and number density of the precipitates and loops were measured by transmission electron microscopy (TEM) and compared to previous atom probe tomography (APT) observations. The loop density systematically decreases with increasing Cr, while the number of α′ precipitates increase at 9Cr and above. A major objective of this work is to estimate the dispersed barrier obstacle strength factors (αj) for loops, clusters and α′ precipitates, based on the combined microstructural observations and corresponding irradiation hardening measurements. Standard dispersed barrier-hardening models and computationally derived superposition rules were least square fit to determine the αj. The optimized hardening predictions are in very good agreement with experiment if mixed linear sum and root square sum superposition rules are used. Five increments of 168 h isochronal anneals of the 6Cr alloy between 300 and 400 °C coarsened the loops, while 300 h anneals of the 18Cr at 500 and 600 °C coarsened and dissolved the α′ precipitates, respectively, consistent with the Fe-Cr phase diagram.

We have investigated proton irradiation induced Nb redistribution in Zr-xNb alloy (x = 0.4, 0.5, 1.0) by using Scanning Transmission Electron Microscopy (STEM) and Atom Probe Tomography (APT). Initially without an irradiation effect, the Zr matrix only contains βNb and native Laves phase native precipitates. However, after 2 MeV proton irradiation at 350 °C, STEM/EDS showed irradiation induced precipitation of Nb-rich needle-like particles (up to ∼100 nm in length) in the Zr matrix. APT also revealed that Fe- and Nb -enriched nano-clusters (less than 20 nm diameter) are present in the Zr matrix for Zr-0.5Nb and Zr-1.0Nb. Irradiation was found to reduce the matrix Nb concentration in the Zr solid solution, This reduction may contribute to the improved corrosion resistance of ZrNb alloys in the reactor environment at high burnup.

International audience; Atom Probe Tomography (APT) analysis of specimens embedded with metal oxide/metal leads to nonintuitive observations of a very thin layer of oxide at the interface due to oxygen migration under the influence of electric field in metal oxides. Detailed analyses of the FeO/Fe and ZrO2/ZrO interfaces are presented, explaining observation of the interfacial oxide layer with APT. These findings are relevant to the observation made for APT analysis of devices such as resistive switching, solar cells, oxides grown on metal/alloy during oxidation and corrosion wherein metal oxide is in interface with metallic layers. Because the APT technique is based on the application of an electric field on the oxide/metal interface, oxygen ions are driven toward the metal electrode and leads to a reaction with the metal and the formation of the additional interfacial oxide layer. Atomistic simulation performed on the FeO/Fe layer subjected to electric field confirms finding of oxygen migration from the oxide layer toward the oxide/metal interface.

In the framework of the SELENIUM project, ZrN coated U-Mo fuel kernels were irradiated in the Belgium Reactor 2 of SCK•CEN to a plate average burnup of 48% 235U and a local maximum burnup just below 70% 235U. The microstructure and chemical composition of ZrN coating before and after neutron irradiation have been analysed using scanning transmission electron microscopy (STEM) equipped with energy dispersive X-ray spectroscopy (EDS) and electron energy loss spectroscopy (EELS), and atom probe tomography (APT). The atomic ratio of N/Zr and fission product distribution determined by three techniques were compared and the combination of three techniques shows the advantages in characterization of chemical information for nuclear materials.

Nanostructured electrode materials represent a promising path forward to dramatically improving the performance of both Li-ion and beyond Li-ion battery systems; however, difficulties in characterizing the structural and electrochemical changes that take place in nanoscale systems, which are often poorly crystalline or amorphous, make it difficult to develop design rules for synthesizing new materials with optimal performance. Bilayered vanadium oxide-based materials (BL-V2O5) are an ideal platform for understanding the underlying physicochemical properties that determine capacity in nanomaterials, with electrochemically-synthesized V2O5 (EC-V2O5) exhibiting particularly high capacities. In this work we provide evidence that the source of high practical capacity in EC-V2O5 is the presence of “structural hydroxyl groups” that are an intrinsic feature of the electrochemical synthesis method. Using both in situ and ex situ characterization methods, we demonstrate that structural OH species are highly stable and persist in the structure during reversible cycling. We hypothesize three important roles for structural OH groups: they maintain a sufficient interlayer spacing to allow the physical diffusion of cations over multiple cycles; they maintain a consistent solvating environment in the bilayer regardless of structural H2O content; and they reduce the symmetry of vanadium polyhedra to favor electron transfer over pseudocapacitive adsorption, making it possible to access close to theoretical capacity. These insights have broad implications for understanding the performance of a variety of hydrated oxide systems, and indicate that the formation of covalently-bound hydroxyoxide species can lead to further improvements in the performance of nanoscale materials.

Electron Backscattered Diffraction (EBSD) is an effective technique for revealing many details about the microstructure of materials, e.g. crystallographic orientation, grain size, grain boundary properties, texture, intragranular misorientation, and subgrain formation. Since these details are of interest for improving the irradiation performance understanding of any irradiated nuclear fuel (e.g. swelling behavior), this technique has been applied successfully on U-7 wt% Mo before and after irradiation. This fuel is a high-density, low-enriched uranium fuel currently being developed for application in research and test reactors. Based on the results of this characterization, it was found that when as-fabricated U-7 wt% Mo is irradiated to around 5.3 × 1021 fissions/cm3 the original large grains (diameter ∼4 μm) are subdivided into much smaller grains (diameter ∼0.3 μm) and most of these subdivided grain boundaries are low angle boundaries. The EBSD analysis suggests that the grain subdivision in the irradiated U-7 wt% Mo was driven by polygonization, not recrystallization as defined by classic metallurgy as a result of heavy cold work followed by heat treatment.

Ferritic-martensitic steels are leading candidates for many nuclear energy applications. However, formation of nanoscale α′ precipitates during thermal aging at temperatures above 450 °C, or during neutron irradiation at lower temperatures, makes these Fe-Cr steels susceptible to embrittlement. To complement the existing literature, a series of Fe-9 to 18 Cr alloys were neutron-irradiated at temperatures between 320 and 455 °C up to doses of 20 dpa. In addition, post-irradiation annealing treatments at 500 and 600 °C were performed on a neutron-irradiated Fe-18 Cr alloy to validate the α-α′ phase boundary. The microstructures were characterized using atom probe tomography and the results were analyzed in light of the existing literature. Under neutron irradiation and thermal annealing, the measured α′ concentrations ranged from ∼81 to 96 at.% Cr, as influenced by temperature, precipitate size, technique artifacts, and, possibly, cascade ballistic mixing.

The heavy-ion irradiation behavior of bulk zirconium-niobium multilayered composites was investigated up to large doses. Multilayers with an average individual layer thicknesses ranging between 15 and 80 nm were synthesized by accumulative roll bonding technique. The irradiation was performed with a defocused 7 MeV Zr2+ ion beam at 500 °C. The maximum dose achieved was ∼145 dpa at the depth of ∼1.5 μm from the irradiated surface. Sub-surface microstructural damage and the chemical redistribution were characterized by transmission electron microscopy and energy dispersive spectroscopy, respectively. Irrespective of the layer thicknesses, the irradiation condition caused layer instability and the extent of damage varied with the dose levels. Doses lesser than ∼60 dpa caused layer fragmentation and greater than ∼60 dpa resulted in layer dissolution. The chemical mixing of layers occur to a depth of ∼1 μm, consuming multiple bi-layer periods. Despite the elevated irradiation temperature (500 °C) and a slightly positive heat of mixing (+6 kJ/mol), no phase separation was observed and single-phase was retained in the mixed region. The results demonstrate that chemical mixing was facilitated by the liquid phase miscibility of Zr and Nb, which propelled the interdiffusion within the thermal spikes towards mixing.

This work assesses the potential of proton irradiation to simulate the neutron damage to the matrix and laves phase Zr(Fe,Cr)2 precipitates in Zircaloy-4. Isothermal proton irradiation has been performed on Zircaloy-4 samples at irradiation temperatures ranging from 250 to 350 °C. Two-step proton irradiation was also performed to enhance the amorphization of and iron loss from the laves phase Zr(Fe,Cr)2 precipitates. The irradiated microstructures, including dislocation loops and rafts near SPPs, were observed in proton irradiated Zircaloy-4, which are consistent with neutron irradiated material at a similar damage level. The amount of irradiation-induced hardening after proton irradiation was similar to post neutron irradiated data. The significant amorphization of the SPPs and concurrent Fe redistribution observed on neutron irradiated materials can be effectively emulated using a two-step proton irradiation on Zircaloy-4. Hence, the neutron irradiation effect on Zircaloy-4 can be mostly captured using the two-step proton irradiation described in this study.

A comprehensive characterization study was carried out on 1 and 4 cycles X2®, aiming to reveal effect of irradiation-induced alloying element redistribution on the in-reactor corrosion kinetics. Using a combination of (scanning) transmission electron microscopy ((S)/TEM) and atom probe tomography (APT), the results strongly evidenced the existence of Nb-rich native precipitates and irradiation-induced platelets (IIPs)/nanoclusters in both metal and oxide. In addition, the suboxide may reject Nb content back into the metal matrix to achieve a thermodynamic equilibrium state. On the other hand, Fe-rich transition cracks were observed for the first time, along with Fe-rich oxide grain boundaries.

Understanding the microstructural and phase changes occurring during irradiation and their impact on metallic fuel behavior is integral to research and development of nuclear fuel programs. This paper reports systematic analysis of as-fabricated and irradiated low-enriched U-Mo (uranium-molybdenum metal alloy) fuel using atom probe tomography (APT). This study is carried out on U-7 wt.% Mo fuel particles coated with a ZrN layer contained within an Al matrix during irradiation. The dispersion fuel plates from which the fuel samples were extracted are irradiated at Belgian Nuclear Research Centre (SCK CEN) with burn-up of 52% and 66% in the framework of the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) project. The APT studies on U-Mo particles from as-fabricated fuel plates enriched to 19.8% revealed predominantly γ-phase U-Mo, along with a network of the cell boundary decorated with α-U, γ’-U2Mo, and UC precipitates along the grain boundaries. The corresponding APT characterization of irradiated fuel samples showed formation of fission gas bubbles enriched with solid fission products. The intermediate burnup sample showed a uniform distribution of the typical bubble superlattice with a radius of 2 nm arranged in a regular lattice, while the high burnup sample showed a non-uniform distribution of bubbles in grain-refined regions. There was no evidence of remnant α-U, γ’-U2Mo, and UC phases in the irradiated U-7 wt.% Mo samples.

Understanding and controlling phase transformations in metals and alloys under irradiation is vital to the design of irradiation-tolerant materials. The L12-ordered intermetallic phase in nickel-based superalloys provides superior high-temperature mechanical properties and corrosion resistance. Here, we studied the order-disorder transformation kinetics of an L12-structured gamma-prime phase in nickel-based superalloy Rene N4 under krypton ion irradiation and the effect of pre-implanted helium gas bubbles on the kinetics of phase transformation. The presence of pre-implanted helium gas bubbles significantly enhances the order-disorder transformation of L12-structured gamma-prime phase under irradiation at 300 °C. The evolution of pre-implanted helium gas bubbles under irradiation implies that helium gas bubbles could effectively trap irradiation-induced vacancies and impede the reordering process of anti-site defects in gamma-prime phase.

Engineered spinodal decomposition and spinodal precursors to precipitation hardening are effective approaches for tailoring material's properties and performance. These approaches leverage the spinodal and binodal nature of miscibility gaps. However, little is known about the chemical and crystallographic mechanisms controlling phase evolution across the spinodal curve and into the binodal regime. This study aims to reveal spinodal-binodal phase transformation in a binary U-50 wt.% Zr model alloy through coupled X-ray diffraction (XRD), in-situ transmission electron microscopy (TEM), and atom probe tomography (APT). The hexagonal ⍵-UZr2+x phase initially undergoes a spinodal decomposition into interconnected nanosized hexagonal Zr-rich and hexagonal U-rich domains at ~575 °C, near the hexagonal-bcc phase transformation. Between 600 and 800 °C, the microstructure evolution is dominated by the coarsening and chemical purification of decomposed bcc domains and a transition from spinodal to binodal nucleation. At 800 °C, the metastable Zr domains transfer to stable α-Zr phase and stay the same phase up to 1000 °C. On the contrary, metastable U domains transfer to orthorhombic α’-U firstly at 800 °C and then to tetragonal β-U at 1000 °C. Spinodal decomposed domain size is found to be linearly related to extrinsic specimen geometry. The effect of oxygen on spinodal decomposition is also discussed. In-situ TEM observations successfully captured the spinodal decomposition and the subsequent transition to binodal phase separation with temperature, providing critical evidence for an expanded miscibility gap, concerning both composition and temperature, on the U-Zr phase diagram.

Atom probe tomography (APT), a 3D microscopy technique, has great potential to reveal atomic scale compositional variations, such as those associated with irradiation damage. However, obtaining accurate compositional quantification by APT for high bandgap materials is a longstanding challenge, given the sensitivity to field evaporation parameters and inconsistent behaviors across different oxides. This study investigates the influence of APT laser energy and specimen base temperature on compositional accuracy in single crystal thoria (ThO2). ThO2 has a broad range of applications, including advanced nuclear fuels, sensors, lasers and scintillators, electrodes, catalysis, and photonics and optoelectronics. The expected stoichiometry of ThO2 is achieved at APT base temperature of 24 K and laser energy of 100 pJ. To overcome mass resolution limitations associated with significant thermal tails, Bayesian methods are applied to deconvolute ion identity within the mass spectra. This approach affirms that the parameters chosen are appropriate for APT analysis of ThO2.

The effect of proton irradiation on corrosion rate of α-annealed and β-quenched Zr-0.5Nb alloys is investigated. The major focuses of this study are to understand i) if the nucleation of irradiation-induced platelets (IIPs)/nanoclusters requires dissolution of Nb-rich native precipitates, ii) if the irradiated native precipitates and interlaths are stable in the oxide, and iii) how much Nb content in the solid solution is suitable to lower the corrosion rate for Zr-Nb alloys. To answer these questions, the major characterization techniques used in this study are APT and (S)TEM/EDS to study the microstructure and microchemistry evolution following irradiation and oxidation.

Tristructural isotropic (TRISO) coated particle fuel, as a key fuel concept for the high-temperature gas-cooled reactors and a candidate accident-tolerant fuel, has been investigated under the US-DOE Advanced Gas Reactor Fuel Development and Qualification Program. Extensive studies have been conducted to evaluate the fission-product release, diffusion of Ag, Pd, and Cd in the SiC layer and the TRISO coating system, and safety-test performance. However, to date there are limited reported results on the fuel kernels’ response to irradiation with or without post irradiation safety testing. To incrementally fill this knowledge gap, extensive studies using transmission electron microscopy (TEM) and atom probe tomography (APT) were conducted on a TRISO fuel-particle kernel with a 19.74% 235U enrichment, irradiated to 18.63% FIMA and subsequently subjected to safety testing at 1600 °C for 300 h. Microstructural characterizations, elemental analysis, and phase identification were conducted using conventional TEM and scanning TEM imaging, energy-dispersive X-ray spectroscopy, selected-area electron diffraction and APT. The following findings were made: (1) significant reconstructions and phase evolutions occurred in the irradiated and post safety-tested fuel kernel, and its microstructure consists of two primary phases, namely a higher-Z (atomic mass) UC phase and a lower-Z UO phase, (2) no fission-gas bubbles are identified within the fuel kernel, that can be attributed to the high-temperature post safety-testing, (3) fission products Zr, Nb, Mo, Ru, Tc and Rh were found to segregate preferentially into UC phase or to form metallic precipitates while the lanthanide fission products tend to stay in the solution of UO phase, and (4) Pd was detected in the rod-shaped precipitates.

The microstructure of EBR-II neutron-irradiated austenitic stainless steel 304 and nickel-base alloy X-750 was investigated. Both alloys were irradiated at low dose rates (∼2 × 10−8 dpa/s) to a neutron fluence of 6.9 × 1022 n/cm2 (E > 0.1 MeV) at 371–389 °C. Different types of defects, including Frank loops, cavities, and precipitates were characterized. The Frank loops in Type 304 stainless steel (SS) are larger in size (∼50 nm in diameter) and lower in number density (2.58 × 1021 m−3), compared to most previous higher dose rate neutron irradiation studies. The Frank loops in X-750 have an average size 26.0 nm of and a number density of 9.44 × 1021 m−3. In 304 SS and X-750, cavities are of ∼20 nm and ∼14 nm in diameter, respectively. The swelling of both alloys was found to be insignificant. In 304 SS, Ni and Si were found enriched at the cavity surfaces and Ni,Si-rich precipitates were also found. Multivariate statistical analysis using non-negative matrix factorization reveals that these Ni,Si-rich precipitates contain only ∼5.7 at.% Si, differing from the Ni3Si γ ’ precipitates found in several previous studies. In X-750, L12-structured γ ’ precipitates were found, and multivariate statistical analysis confirmed the 3:1 stoichiometry (Ni3(Ti,Al)) of the γ ’ precipitates and the superlattice reflections confirmed the stability of the crystal structure of these γ ’ precipitates, indicating higher-than-expected precipitate stability under high-dose neutron irradiation.

A novel approach is presented to determine the local burnup in irradiated fuels using isotopic quantification obtained by Atom Probe Tomography (APT). Considering the volume of sample used (

Microstructural changes in a Fe–15 at.%Cr model alloy neutron irradiated to 1.82 dpa at 290 °C were characterized by atom probe tomography. Homogenously distributed α′ precipitates as well as fewer clusters containing Si, P, Ni, and Cr, were observed in the matrix. Grain boundary analyses before and after irradiation revealed segregation of Cr, with W-shape concentration profiles developing in the vicinity of grain boundary carbide and nitride particles. After irradiation, impurities such as C, Si and P were segregated to the grain boundaries. Zones depleted of α′ clusters, and Si were found at the interfaces of carbide and nitride precipitates and along grain boundaries in the vicinity of these precipitates.

A series of model Fe–Cr alloys containing 3–18 at.% Cr was neutron irradiated at a nominal temperature of 563 K to 1.82 dpa. Solute distributions were analyzed by atom probe tomography, which revealed α′ precipitation for alloys containing more than 9 at.% Cr. Both the Cr concentration dependence of α′ precipitation and the measured matrix compositions are in agreement with the recently published Fe–Cr phase diagrams. An irradiation-accelerated precipitation process is strongly suggested.

Atom probe tomography has without any doubt become a routine technique to analyze the detailed three-dimensional chemistry of materials at the nanoscale. This article provides a general overview of what APT can reliably do today and what it might do tomorrow in terms of material characterization. The recent achievements in the analysis of new materials and new materials structures are first presented allowing some speculation on future possible developments. The ability to provide unique quantitative chemical information to link processing to device performance is then reviewed in the context of the recent nanowire and gate structures analyses. Finally examples of the systematic use of atom probe tomography to explore material behaviors and kinetic processes controlling microstructure evolution are presented.

International audience