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Active Tuning of Optical Constants in the Visible–UV: Praseodymium-Doped Ceria—a Model Mixed Ionic–Electronic Conductor
Mixed ionic-electronic conductors offer chemical and electrical means for active tuning of their optical constants, e.g. with variations in oxygen non-stoichiometry in Pr 0.1 Ce 0.9 O 2-δ , enabling implementation of adaptive thin film optical devices. Here, we demonstrate in-situ chemo-tuning of the extinction coefficient in Pr 0.1 Ce 0.9 O 2-δ at elevated temperatures and provide a tuning model that treats the interdependence of mobile oxygen vacancies and small polarons coupled to variations in optically active praseodymium ions. Further, a new means for electro-tuning of the optical constants of mixed ionic-electronic conductors was demonstrated experimentally and modeled for Pr 0.1 Ce 0.9 O 2-δ thin films deposited on grid-like electrode structures. Modeling of non-steady state optical transmittance modulations in the latter allows for estimation of oxygen vacancy mobility that determines the switching speed of the device. Quenched-in values of nr and k to room temperature become nonvolatile, providing a modulation range in the extinction coefficient of Δk ~0.1 (change of ~800%) and in the refractive index of Δn r ~0.1 (relative to initial n r of ~2.35). Key figures of merit including, transmission optical modulation of ~0.04 per 1 mV nm -1 , switching energy per area of 670 pJ μm -2 and switching times of seconds, were demonstrated, with further improvements possible.
Persistent and partially mobile oxygen vacancies in Li-rich layered oxides
Increasing the energy density of layered oxide battery electrodes is challenging as accessing high states of delithiation often triggers voltage degradation and oxygen release. Here we utilize transmission-based X-ray absorption spectromicroscopy and ptychography on mechanically cross-sectioned Li 1.18–x Ni 0.21 Mn 0.53 Co 0.08 O 2–δ electrodes to quantitatively profile the oxygen deficiency over cycling at the nanoscale. The oxygen deficiency penetrates into the bulk of individual primary particles (~200 nm) and is well-described by oxygen vacancy diffusion. Using an array of characterization techniques, we demonstrate that, surprisingly, bulk oxygen vacancies that persist within the native layered phase are indeed responsible for the observed spectroscopic changes. We additionally show that the arrangement of primary particles within secondary particles (~5 μm) causes considerable heterogeneity in the extent of oxygen release between primary particles. Finally, our work merges an ensemble of length-spanning characterization methods and informs promising approaches to mitigate the deleterious effects of oxygen release in lithium-ion battery electrodes.
Microstructural evolution of compositionally complex solid-solution alloys under in-situ dual-beam irradiation
Here, this work attempts to link the microstructural evolution of single-phase compositionally complex alloy (CCA) compositions under dual-beam irradiation to their Mn-content via the stacking-fault energy (SFE) and vacancy migration energies. Two alloys, Cr 18 Fe 27 Mn 27 Ni 28 and Cr 15 Fe 35 Mn 15 Ni 35 , along with less compositionally complex pure Ni and Fe 56 Ni 44 binary were irradiated at 500 and 600 °C under dual-beam 1 MeV Kr 2+ and 16 keV He + heavy-ions up to 7 displacements per atom (dpa) with a He/dpa ratio of 0.75 %/dpa using in situ transmission electron microscopy (TEM). Due to the bubble-stabilizing effect of implanted He, bubbles were observed in all irradiations, and populations of faulted interstitial loops were characterized in Cr 18 Fe 27 Mn 27 Ni 28 and Cr 15 Fe 35 Mn 15 Ni 35 . A reduction in swelling was observed in the two CCAs compared to pure Ni and Fe 56 Ni 44 . Although swelling increased from 500 to 600 °C in Fe 56 Ni 44 , Cr 15 Fe 35 Mn 15 Ni 35 and Cr 18 Fe 27 Mn 27 Ni 28 both swelled slightly more at 500 °C. This was attributed to the difference in vacancy mobility, stronger pinning effect of vacancies on He, and the sink strength of faulted dislocation loops. Faulted interstitial loops nucleated with a higher number density and dislocation line density in Cr 15 Fe 35 Mn 15 Ni 35 at both temperatures, and at 600 °C in both materials. The differences in faulted loop population and temperature effect on swelling are correlated to the Mn-content and the measured SFE (20.2 ± 6.7 mJ/m 2 for Cr 18 Fe 27 Mn 27 Ni 28 and 9.2 ± 3.4 mJ/m 2 for Cr 15 Fe 35 Mn 15 Ni 35 ).
Diffusion behavior of lanthanide fission products in bcc Fe cladding: A first-principles study
Fuel-cladding chemical interaction poses significant challenges in nuclear reactors, where fission products generated from nuclear fuel interact with Fe-based cladding materials, potentially compromising their structural integrity. This study investigates the diffusion behavior of lanthanide fission products, Lanthanum (La), Cerium (Ce), Praseodymium (Pr), and Neodymium (Nd), within body-centered cubic (bcc) Fe cladding using the density functional theory, nudged elastic band method, and self-consistent mean field theory. Our results reveal significant vacancy binding energies, particularly with the 1st and 2nd nearest neighbors, which diminish beyond the 5th nearest neighbor, with La exhibiting the strongest binding affinity, followed by Nd, Ce, and Pr. The nudged elastic band calculations indicate significant high barriers for the dissociation of 1st nearest neighbor vacancy-solute pairs for all fission products. The tracer diffusion coefficients of these fission products were derived in an Arrhenius form, with a magnetic correction that accounts for the high-temperature paramagnetic state. The significant trapping effect of vacancies caused by a very dilute concentration of fission products reduces vacancy mobility, potentially leading to modifications in point defect supersaturation, void nucleation, and swelling under irradiation. These represent critical challenges for irradiated cladding materials. The tracer diffusion coefficients indicate that Nd diffuses the fastest, followed by La, Ce, and Pr. Furthermore, this study provides essential insights for understanding fission product transport in cladding materials and informs future design strategies to mitigate fuel-cladding chemical interaction, ultimately enhancing nuclear reactor safety and performance.
Understanding effects of chemical complexity on helium bubble formation in Ni-based concentrated solid solution alloys based on elemental segregation measurements
Here, helium bubble formation and swelling were systematically studied in Ni-based concentrated solid solution alloys containing different numbers and types of elements. Our microscopy analysis showed that although increasing the alloy chemical complexity helps suppress bubble formation in general, there is no monotonic relationship between the bubble growth rate and the number of alloying elements. Certain elements (e.g., Fe and Pd) are more effective in suppressing bubble growth than others (e.g., Cr and Mn). Atom probe tomography was applied to accurately measure elemental segregation around bubbles, revealing unique effects of certain alloying elements on vacancy migration towards bubbles. More specifically, the high vacancy mobility via Cr sites leads to a large vacancy flux and an increased bubble size, while the high degree of atomic size mismatch introduced by Pd helps deflect vacancy flow away from bubbles and decrease the amount of swelling. The effects identified in this study provide new strategies to design concentrated solid solutions with superior resistance to swelling.
Atomistic modeling of lanthanide diffusion in refractory body-centered cubic molybdenum
Lanthanide fission products can strongly interact with candidate cladding alloys, but their transport properties in refractory metals remain poorly understood. Here, in this work, we investigate the atomic-scale diffusion behavior of La, Ce, Pr, and Nd in body-centered cubic (bcc) molybdenum, a potential candidate for advanced nuclear cladding. Self-consistent mean-field transport modeling is performed to evaluate the fission product transport and vacancy mobility, informed by first-principles and nudged elastic band calculations of vacancy formation energies, migration barriers, and solute–vacancy binding characteristics. Compared with bcc Fe, lanthanide solutes in bcc Mo exhibit slower tracer diffusion due to higher vacancy formation and migration energies. Furthermore, the calculations reveal that the influence of fission products on migration barriers in bcc Mo are not as extensive in range compared to bcc Fe. Among the studied lanthanides, La exhibits the strongest vacancy binding while also being the fastest diffuser in Mo. These findings highlight how refractory bcc alloys can reduce fission product infiltration, offering valuable insight into the development of durable cladding systems for advanced reactors.
The role of excess vacancies in stabilizing solute clusters in low-alloy steels
The formation of solute clusters in irradiated low-alloy steels, such as reactor pressure vessel steels, is a critical cause of radiation hardening and embrittlement. However, the chemical interactions among solute elements and excess vacancies remain to be fully understood. This study employs density functional theory, cluster expansion, and lattice-based Monte Carlo simulations to examine the stability and morphology of nano-size coherent solute clusters with excess vacancies. The findings reveal that excess vacancies are crucial in stabilizing and promoting the growth of Mn-Ni-Si-vacancy clusters. A minimum of seven vacancies is required for stable nucleation and growth of these clusters. These Mn-Ni-Si clusters act as defect sinks, effectively trapping and absorbing mobile vacancies generated under irradiation. Furthermore, phosphorus (P) preferentially dissolves in Mn-Ni-Si clusters due to chemical coupling with vacancies. In conclusion, this study offers new insights into solute-vacancy interactions, enhancing our understanding of solute-defect cluster formation and embrittlement in low-alloy steels.
Short Range Order in Disordered Spinel and the Impact on Cation Vacancy Transport
Spinels are important complex oxides for use in radiation damage environments and resulting from the corrosion of steels. It is known that, in these environments, normal spinels exist with some concentration of antisite cation pairs known as inversion. In this work we show that even in highly disordered states characterized by high levels of inversion, spinel still shows some short range order (SRO) that manifests itself in antisite chains. The propensity to form these antisite chains is confirmed through Monte Carlo simulations which find that the length of chains which can form depend on the spinel chemistry. We also consider the effect of antisite chains on the diffusivity of cation vacancies and find the effect strongly depends on the spinel chemistry. At the extremes, chains in FeCr2O4 significantly increase vacancy transport but in MgAl2O4 chains have the inverse effect of drastically reducing vacancy mobility. The explanation of these dramatically different effects results from the assessment of the thermodynamic stability of the antisite chains and contrasting attractive/repulsive interactions of vacancies with the chains.
Impact of Oxygen Non-Stoichiometry on Near-Ambient Temperature Ionic Mobility in Polaronic Mixed-Ionic-Electronic Conducting Thin Films
Enhanced ionic mobility in mixed ionic and electronic conducting solids contributes to improved performance of memristive memory, energy storage and conversion, and catalytic devices. Ionic mobility can be significantly depressed at reduced temperatures, for example, due to defect association and therefore needs to be monitored. Measurements of ionic transport in mixed conductors, however, proves to be difficult due to dominant electronic conductivity. This study examines the impact of different levels of quenched-in oxygen deficiency on the oxygen vacancy mobility near room temperature as measured by a novel dynamic current-voltage analysis. A Pr 0.1 Ce 0.9 O 2-δ film was grown by pulsed laser deposition and subsequently annealed, from 400-600° C, in various oxygen partial pressures to modify its oxygen vacancy concentration while minimizing microstructural growth and cation segregation. To monitor changes in film non-stoichiometry, we leverage the existence of an optical absorption center, related to the oxidation state of Pr ions in Pr0.1Ce0.9O2-δ. The oxygen vacancy migration enthalpy was found to exhibit a small increase from 0.73± 0.04 to 0.79 ± 0.02 eV with increasing oxygen deficiency, while the pre-factor was found to increase by a factor of 135. Here, a nearly 13-fold increase in ionic mobility at 60 °C for increases in oxygen non-stoichiometry from 0.032 ± 0.001 to 0.042± 0.001 was thereby detected. Raman spectroscopy was employed to rule out changes in strain as the primary cause for the significant change in mobility. Several factors potentially contributing to the large pre-factor changes are examined and discussed. Insights into how ionic defect concentration can markedly impact ionic mobility should help in elucidating the origins of variations seen in nanoionic devices.
Cluster Dynamics Simulations of Intra-Granular Fission Gas Bubble Size and Pressure Evolution in UO 2
Fission gases such as xenon (Xe) play a critical role in determining the behavior and response of nuclear fuel. Given that Xe has little solubility in UO 2 , it accumulates and forms bubbles, which significantly impact fuel performance. Intra- and inter-granular bubble nucleation and growth can lead to fuel swelling, and once bubbles interconnect at grain boundaries, fission gas can be released into the plenum. At low temperatures, limited uranium vacancy mobility can restrict swelling, therefore causing the bubbles to become highly pressurized. Consequently, this can induce micro-cracking, promote fission gas release (increasing the likelihood of cladding failure), and even lead to fuel pulverization under accident conditions such as a loss of coolant accident. As bubble evolution is strongly influenced by local temperature and fission rate, markedly different behavior occurs across the radial profile of the fuel pellet. Capturing the mechanisms that underpin bubble evolution is therefore important to predict these behaviors in the fuel. Previous models describing important mechanisms informed by lower length scale simulations have been developed under the NEAMS program. These can describe the evolution of a single bubble type (i.e., single value for radius and pressure) at each position in the pellet, for instance using the Centipede cluster dynamic code. However, in reality, a full distribution in bubble sizes and pressures exists within the microstructure at a given position in the pellet. To address this the cluster dynamics code Xolotl, which can predict Xe and vacancy phase space (i.e., bubble distributions) for intra-granular bubbles, has been used before. Prior work benchmarked the Xolotl code against the Centipede cluster dynamics code to ensure compatibility and to verify that mobile defect properties are adequately transferred between the two codes, along with some physics improvements. In this work, we go further by introducing a physics-based set of improvements that will allow us to accurately predict bubble size distributions and internal bubble pressures under representative UO 2 irradiation conditions. The improvements include (i) coupling bubble-defect reaction energies to a virial equation of state (EOS), (ii) including a bubble surface tension contribution, (iii) incorporating radiation-induced re-solution of Xe and vacancies, (iv) enabling pressure-driven dislocation loop punching through an effective emission of interstitial clusters informed by interstitial loop energetics, (v) accounting for radiation induced athermal diffusion of Xe, and (vi) implementing a Booth-type grain boundary sink representation for all mobile defects and defect clusters. After these modifications, we observe good agreement of Xolotl fission gas bubble size and concentration predictions with legacy experimental measurements. Additionally, it allows the distribution of Xe bubble pressures and radius to also be predicted and compared to data produced through the Advanced Fuels Campaign (AFC) program. Here, we have done this by running simulations under conditions similar to the AFC post-irradiation examination (PIE) samples irradiated at North Anna 2 light water reactor (LWR). Our results shows excellent agreement with these experimental measurements.
Diffusion Behavior of Oversized Fission Products in bcc Fe Cladding: A First-Principles Study
Fuel-Cladding Chemical Interaction (FCCI) poses significant challenges in nuclear reactors, where fission products from nuclear fuel interact with Fe-based cladding materials, potentially compromising their structural integrity. This study investigates the diffusion behavior of oversized fission products, Pr, Nd, Ce, and La, within bcc Fe cladding using density functional theory (DFT), nudged elastic band (NEB) method, and self-consistent mean field (SCMF) theory. Our results reveal significant long-range vacancy binding energies, particularly up to the 6th nearest neighbor, with La exhibiting the strongest binding affinity, followed by Nd, Ce, and Pr. The NEB calculations indicate significant high barriers for the dissociation of 1nn vacancy-solute pairs for all fission products. The tracer diffusion coefficients of these fission products was derived in Arrhenius form. The significant trapping effect of vacancies by a very dilute amount of fission products reduces vacancy mobility, leading to an oversaturation of point defects, void nucleation, and swelling. These are critical issues for irradiated cladding materials. The tracer diffusion coefficients indicate that Nd diffuses the fastest, followed by La, Ce, and Pr. This study provides essential insights for developing advanced cladding materials and design strategies to mitigate FCCI, ultimately enhancing nuclear reactor safety and performance.
In situ microstructural evolution in face-centered and body-centered cubic complex concentrated solid-solution alloys under heavy ion irradiation
This study characterizes the microstructural evolution of single-phase complex concentrated solid-solution alloy (CSA) compositions under heavy ion irradiation with the goal of evaluating mechanisms for CSA radiation tolerance in advanced fission systems. Three such alloys, Cr 18 Fe 27 Mn 27 Ni 28 , Cr 15 Fe 35 Mn 15 Ni 35 , and equimolar NbTaTiV, along with reference materials (pure Ni and E90 for the Cr-FeMnNi family and pure V for NbTaTiV) were irradiated at 50 K and 773 K with 1 MeV Kr ++ ions to various levels of displacements per atom (dpa) using in-situ transmission electron microscopy. Cryogenic irradiation resulted in small defect clusters and faulted dislocation loops as large as 12 nm in face-centered cubic (FCC) CSAs. With thermal diffusion suppressed at cryogenic temperatures, defect densities were lower in all CSAs than in their less compositionally complex reference materials indicating that point defect production is reduced during the displacement cascade stage. High temperature irradiation of the two FCC CSA resulted in the formation of interstitial dislocation loops which by 2 dpa grew to an average size of 27 nm in Cr 18 Fe 27 Mn 27 Ni 28 and 10 nm in Cr 15 Fe 35 Mn 15 Ni 35 . This difference in loop growth kinetics was attributed to the difference in Mn-content due to its effect on the nucleation rate by increasing vacancy mobility or reducing the stacking-fault energy.
Temperature-dependent cavity swelling in dual-ion irradiated Fe and Fe-Cr ferritic alloys
Fe-Cr ferritic-martensitic (FM) steels are promising structural material candidates for fusion and advanced fission reactors due to their attractive mechanical properties and volumetric swelling resistance. However, significant discrepancies exist regarding the effect of solutes and irradiation temperature on cavity swelling under ion versus neutron irradiation conditions. In this study, simultaneous dual ion irradiations (8 MeV Ni 3+ ions and energy-degraded 3.5 MeV He 2+ ions) were used to quantify the cavity swelling behavior in ultra-high purity Fe and Fe-Cr alloys (3-14 wt.% Cr), Fe-10 wt.% Cr-780 wt.ppm C, and Eurofer97 FM steel. The irradiations were conducted over a wide temperature range (400-550°C) with a mid-range dose of ~30 displacements per atom (dpa) and 0.1 appm/dpa He implantation rate. Here, using state-of-the-art transmission electron microscopy (TEM), we reveal that pure Fe has a ~50°C lower peak swelling temperature difference than Fe-Cr alloys, which is attributed to higher vacancy mobility in pure Fe. Chromium solute appears to strongly suppress cavity swelling in Fe-Cr alloys for temperatures below ~470°C, but seems to have little effect or slightly enhances swelling above ~470°C. Cavities were observed in all the irradiated samples between 400-550°C. This indicates that the narrow temperature range of observable cavities reported in prior ion irradiated Fe-Cr ferritic alloy studies is likely an artifact associated with the use of low ion energies (<5 MeV), which leads to pronounced near-surface and implanted ion effects that suppress cavity swelling even at midrange depths (particularly at high temperatures).
Thermal diffusivity recovery and defect annealing kinetics of self-ion implanted tungsten probed by insitu transient grating spectroscopy
Tungsten is a promising candidate material for plasma-facing armour components in future fusion reactors. A key concern is irradiation-induced degradation of its normally excellent thermal transport properties. In this comprehensive study, thermal diffusivity degradation in ion-implanted tungsten and its evolution from room temperature (RT) to 1073 K is considered. Five samples were exposed to 20 MeV self-ions at RT to achieve damage levels ranging from 3.2 x 10 -4 to 3.2 displacements per atom (dpa). Transient grating spectroscopy with in-situ heating was then used to study thermal diffusivity evolution as a function of temperature. Using a kinetic theory model, an equivalent point defect density is estimated from the measured thermal diffusivity. The results showed a prominent recovery of thermal diffusivity between 450 K and 650 K, which coincides with the onset of mono-vacancy mobility. After 1073 K annealing samples with initial damage of 3.2 x 10 -3 dpa or less recover close to the pristine value of thermal diffusivity. For doses of 3.2 x 10 -2 dpa or higher, on the other hand, a residual reduction in thermal diffusivity remains even after 1073 K annealing. Transmission electron microscopy reveals that this is associated with extended, irradiation-induced dislocation structures that are retained after annealing. A sensitivity analysis shows that thermal diffusivity provides an efficient tool for the assessing total defect content in tungsten up to 1000 K.
Behavior of helium cavities in ion-irradiated W-Ni-Fe ductile-phase toughened tungsten
This study reports on the distribution of helium (He) cavities in a hot-rolled W-Ni-Fe ductile-phase toughened tungsten (DPT W) composite irradiated to a dose and a helium concentration that are comparable to those in the material after 5-year irradiation in a conceptual fusion power plant. Here, the DPT W sample consists of W particles embedded in a ductile-phase NiFeW matrix with a nominal composition of 90W-7Ni-3Fe by weight. It was hot-rolled to a thickness reduction by 87% (87R DPT W). Sequential irradiation of the material with 1.2 MeV Ni + ions to a fluence of 2.15×10 16 Ni + /cm 2 and 90 keV He + ions to 6.5×10 15 He + /cm 2 was performed at 973 K. Larger He cavities with a lower number density are observed in NiFeW than W. Helium cavities are aggregated preferentially along the NiFeW/W interphase boundary. This behavior is not observed along the W/W grain boundary under the same irradiation conditions. A similar distribution of He cavities with comparable sizes is observed in 87R DPT W irradiated with He + ions only, suggesting that the He cavity distribution is not significantly affected by pre-irradiation of Ni + ions at 973 K. The data from this study does not provide any evidence for formation of visible voids or Ni precipitates in W phase irradiated with Ni + ions only. Diffusion and clustering of He atoms and mobile vacancies during ion irradiation at 973 K are responsible for the formation and growth of the He cavities.
The impact of chemistry on anion migration in bixbyite-structured lanthanide oxides
This manuscript describes atomistic calculations of oxygen vacancy and interstitial migration in bixbyite structured lanthanide oxides. We examine two types of compounds, one in which only one type of lanthanide cation is present and a second class in which two lanthanides are present in a 3:1 ratio as dictated by the symmetry of the bixbyite crystal structure. Using temperature accelerated dynamics and the nudged elastic band method, we quantify the role of chemistry on the energy barriers for the most important pathways for both vacancy and interstitial migration. We then analyze the impact of these variations on the overall diffusivity of each defect, quantifying the contribution of each pathway using the theory of kinosons. We find that vacancy mobility can vary by as much as three orders of magnitude through changes in chemistry at 500 K. Changes in interstitial mobility are more modest but can still vary by an order of magnitude. This points to the ability to tune the mass transport characteristics of these compounds through appropriate choices in chemistry. We have also included supplementary information containing the atomic structures of the relevant pathways.
Effects of transition metal carbide dispersoids on helium bubble formation in dispersion-strengthened tungsten
Abstract The formation of helium bubbles and subsequent property degradation poses a significant challenge to tungsten as a plasma-facing material in future long-pulse plasma-burning fusion reactors. In this study, we investigated helium bubble formation in dispersion-strengthened tungsten doped with transition metal carbides, including TaC, ZrC, and TiC. Of the three dispersoids, TaC exhibited the highest resistance to helium bubble formation, possibly due to the low vacancy mobility in the Group VB metal carbide and oxide phases. Under identical irradiation conditions, large helium bubbles formed at grain boundaries in tungsten, while no bubbles were observed at the interfaces between the carbide dispersoid and tungsten matrix. Moreover, our results showed the interfaces could suppress helium bubble formation in the nearby tungsten matrix, suggesting that the interfaces are more effective in trapping helium as tiny clusters. Our research provided new insights into optimizing the microstructure of dispersion-strengthened tungsten alloys to enhance their performance.