Hydrogen (H2) separation membranes have attracted considerable interest in regard to H2 production industries, membrane reactors, and hydrogen isotope separation in fusion energy systems. BaCe0.5Fe0.5O3−δ (BCF), a mixed protonic and electronic conductor, is a promising candidate for hydrogen permeation. However, conventional dense BCF membranes require thickness >0.3 mm to ensure mechanical integrity, resulting in long proton transport distances and limited permeation flux. In this study, we developed a thin BCF membrane supported on a porous yttria-stabilized zirconia–nickel (YSZ–Ni) substrate to overcome these limitations. A dense BCF layer with a thickness of approximately 8 µm was co-sintered with a YSZ–NiO support and then subsequently reduced to form a porous YSZ–Ni substrate, providing mechanical support and facilitating gas diffusion. Owing to excellent sintering compatibility between BCF and YSZ–NiO, a crack-free dense membrane was achieved. Hydrogen permeation tests between 500 and 700 °C showed a substantial flux enhancement compared to thick BCF membranes. At 700 °C and 40% H2 feeding, the BCF||YSZ–Ni membrane reached 2.26 × 10−2 mol m−2 s−1, nearly five times higher than that of a dense BCF pellet under identical conditions. Electrochemical impedance spectroscopy further revealed that BCF exhibits sufficient catalytic activity for hydrogen surface reactions, eliminating the need for an additional catalyst layer. These results demonstrate that YSZ–Ni porous-supported thin BCF membranes offer an effective and scalable strategy for improving hydrogen permeation performance in ceramic-based separation systems.

Protonic ceramic cells (PCCs) have emerged as a promising technology for power generation, energy storage, and value-added chemical synthesis, offering benefits such as fuel flexibility, low emissions, and efficient operation at intermediate temperatures (300–600 ​°C). Recently, significant breakthroughs in materials and manufacturing methods have markedly enhanced the performance of PCCs. However, establishing a fundamental understanding of their electrocatalytic reactions has gained less attention. As a fast and cost-effective method for physicochemical fingerprinting, diffuse reflectance infrared Fourier transform spectroscopy (DRIFTS) has proven to be a surface-sensitive analytical tool for structural and functional studies. This review critically examines the most up-to-date applications of DRIFTS for characterizing key components of PCCs, including oxygen electrodes, protonic electrolytes, and hydrogen electrodes for different applications, with a focus on revealing hydration properties and catalytic reactions, and guiding rational material design. The challenges for advancing DRIFTS, including quantitative capabilities and operando applications for PCC investigations, are highlighted and strategies to tackle these challenges are discussed. Ultimately, this review underscores the critical role of DRIFTS in accelerating the development of high-performance and durable PCCs for next-generation energy solutions, offering methodologies and insights broadly applicable to a wide range of electrochemical energy conversion and storage technologies.

Protonic ceramic electrochemical cells (PCECs) represent a promising class of solid-state energy conversion devices capable of high-efficiency hydrogen production and power generation. However, the practical deployment of planar PCECs is fundamentally constrained by severe structural deformation and mechanical failure during fabrication, stemming from asymmetric shrinkage between the thin electrolyte and the thick NiO-based support layer. In this work, a functionally integrated, symmetry-engineered double-sided electrolyte (DE) design is unveiled, which not only suppresses thermally induced curvature but also unlocks significant gains in electrochemical performance and stability. This architecture intrinsically balances shrinkage dynamics across the cell bilaterally, enabling the fabrication of ultra-flat 5 x 5 cm2 cells with sub-100 mu m thickness variation. A numerical solid mechanics simulation is introduced to investigate and interpret this achievement. Beyond structural advantages, the DE configuration enhances the cell operational stability, delivering a low open-circuit voltage degradation of 9.5 mV/100 h across 80 thermal cycles. This work establishes a compelling paradigm wherein architectural symmetry directly translates to both mechanical fidelity and functional enhancement, offering a promising route toward PCECs scale-up.

Metallic fuels hold numerous advantages over conventional uranium dioxide fuels and are a key component of several liquid metal-cooled advanced reactor concepts including sodium fast reactors. These fuels undergo rapid swelling during early burnup; consequently, they spend most of their reactor lifetime in a porous state. The presence of this porosity alters many of the mechanical properties of the fuel including creep impacting fuel deformation during axial swelling. This work investigates the creep behavior of the porous fuel using a spark plasma sintering technique. Creep tests were performed for the first time on porous α-phase uranium and uranium with 10 wt. % zirconium (U-10Zr) samples. The samples of α-phase uranium and U-10Zr were fabricated from depleted uranium by spark plasma sintering and subjected to uniaxial compressive creep testing. Calculated stress exponents were found to be 2.6±1.6 and 5.7±1.4 for α-U and U-10Zr, respectively, and calculated activation energies were found to be 61.6±1.1kJ/mol for α-U. The creep data were also used to evaluate existing porosity inclusive in creep models.

In this paper, the volume control spark plasma sintering tool has been designed and applied to sinter porous U-10Zr metallic fuels, by which the sintered sample volume can be precisely controlled. Ethanol and NH4HCO3 are used to control the powder compact or as pore formers to control the pore size and pore structure. Without pore formers, the fuel pellet displays an inhomogeneous microstructure consisting of highly porous and highly densified areas. Uneven powder stacking in the green body results in a non-uniform microstructure, and in the closed-packed area, Joule heating accelerates the neck formation and densification. The addition of ethanol reduces the friction between the powders, resulting in isolated pores formed by the stacking of powders during the sintering. By adding NH4HCO3, the pore size, and structure can be well controlled, and an interconnected pore structure can be obtained upon the decomposition of the NH4HCO3. A uniform microstructure and pore distributions can be achieved through the U-10Zr fuel pellets by controlling current flow during the volume control SPS sintering. The microstructure and phase characterization of the sintered porous U-10Zr pellets show major phases of α-U and α-Zr for the sample with short dwelling. For the sample with long dwelling (30 min), the ω UZr2 in the Zr-enriched area has been observed. The strategy of volume control SPS sintering with the assistance of pore formers could be used to fabricate porous U-10Zr metallic fuels to mimic the microstructure evolution of irradiated metallic fuels (including porosity) and could enable a possible solution for the design of new sodium-free metallic fuels for high burnup.

Antimony (Sb) and tellurium (Te) were investigated as potential additives for U-10Zr (wt.%) metallic fuel to limit the fuel-cladding chemical interaction (FCCI) with HT-9 alloy. Neodymium (Nd) was utilized to simulate the formation of lanthanide-based solid fission products which are known to play a detrimental role in FCCI. Fuel alloys of U-Zr-Sb-Nd and U-Zr-Te-Nd were evaluated in their annealed condition and compared against their as-cast conditions. Isothermal diffusion couple experiments were performed between U-Zr-Nd, U-Zr-Sb-Nd, and U-Zr-Te-Nd against the cladding alloy HT9 to evaluate the effectiveness of the additives to stabilize Nd within the fuel alloys, as well as investigate the interaction regions that form between the different fuel alloys and HT9. SbNd and Sb3Nd4, and TeNd are found to be the primary neodymium-based phases formed in the U-Zr-Sb-Nd and U-Zr-Te-Nd alloys, respectively. The zirconium-based phase, Zr2Sb, is also found to form within the former alloy. All phases were found to remain stable through the diffusion experiments and exhibited no interaction with HT9 constituent elements. Preferential interaction between Nd with additivities Te and Sb compared to constituting elements in HT9 was further verified based on density functional theory (DFT) calculated enthalpy of mixing.

Antimony is being investigated as a potential additive to metallic fuel to control fuel-cladding chemical interactions (FCCI). The most detrimental elements involved in FCCI are fission product lanthanides, leading to brittle intermetallics and low melting eutectic phases. Previous investigations of Sb as an additive focused on U-10Zr, in wt. %, as the fuel. The current investigation expands that to include Pu in the fuel. Two alloys, U-20Pu-10Zr-4Sb and U-20Pu-10Zr-4Sb-4Ln (wt. %, Ln=53Nd-25Ce-16Pr-6La) have been investigated using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) to characterize the fuel as-cast microstructure and the microstructure after introduction of lanthanides. Sb reacts with Zr initially, forming Zr2Sb and Zr5Sb3, with as much as 20 at. % interstitial Pu present. In the presence of lanthanides, Sb forms Ln(4)Sb(3) with the lanthanides, containing similar to 14 at. % Pu. The Pu is substitutional for the lanthanides in the crystal lattice.

To accommodate the swelling of metallic fuels induced by the fission gas release for a sodium-free fuel option, advanced U-10Zr fuels with controlled porosities were designed and demonstrated by spark plasma sintering (SPS). U-10Zr fuel pellets with manufactured porosities varying from 35% up to fully dense fuel pellets have been fabricated by controlling mechanical attrition of the starting uranium powders, sintering temperature, pressure, and duration, and the correlation among the microstructure control – porosity – sintering conditions has been established. To further mimick the pore structure in irradiated fuels, different pore formers (NaCl and NH4HCO3) were used to control the pore size and distribution. Microstructure characterization indicates a lamellar reaction zone of enriched Zr and U composed of alpha Zr with 2 at% U and distorted alpha U with 13 at% of Zr, respectively, differing from the arc-melted U-10Zr as a result of rapid consolidation of SPS at lower temperatures and short durations. The thermal conductivity of U10Zr with different porosities were also measured. This work demonstrates the success in manufacturing new metallic fuel forms with controlled porosities and pore distribution, which can be used as model systems to investigate the thermal transfer behavior of metallic fuels in the reactor.

Metallic uranium is the leading fuel form for sodium cooled fast reactors as an enabling technology of future nuclear energy systems. Mechanistic understanding of fuel behaviors and kinetics under thermodynamically-equilibrium and highly non-equilibrium conditions are essential for evaluating fuel performance. It is important to understand and predict the grain and pore evolutions of metallic fuels under thermal and irradiation effects. However, very limited data are available on the grain growth kinetics and mechanisms of pure gamma phase uranium. In this paper, the pure gamma uranium pellets with different length scales were fabricated by combining high-energy ball milling and spark plasma sintering. Isothermal annealing tests with different durations were performed to investigate the grain growth behavior of the gamma phase pure uranium with different initial grain sizes. A parabolic relationship with time was identified for submicron-sized (374 nm) grain growth. In contrast, for the nano-sized (137 nm) sample, the grain size shows a linear relationship with time. The activation energies of grain growth were determined as 199.5 KJ/mol and 80.6 KJ/mol for nano-sized and sub-micron-sized grain structures, respectively. For the nano-sized sample, the rate-control step of grain growth is dominated by the triple-junction migration and thus has higher activation energy than the bulk diffusion, and the dominating mechanism for the sub-micron-sized sample is grain boundary diffusion. The mechanistic understanding and critical data obtained on the kinetics of pure uranium phases are essential to evaluate fuel behavior under thermodynamically-equilibrium conditions and develop a high fidelity model to predict fuel performance.

U-Zr metallic fuel is a promising fuel candidate for Gen Ⅳ fast spectrum reactors. Previous experimental irradiation campaigns showed that the sodium thermal bonded U-10Zr fuel design can achieve a burnup of 10% fissions per initial heavy metal atom (FIMA). Advanced metallic fuel designs are pushing the burnup limit to 20% or even 30% FIMA. To achieve the higher burnup and eliminate the pyrophoric sodium, a prototypical annular fuel has been designed, fabricated, clad with HT-9 in the Materials and Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory (INL) to a peak burnup of 3.3% FIMA. During irradiation, the mechanical contact between fuel and cladding acts as a thermal bond. The irradiation lasted for 132 days in the reactor. Recently, the archived fresh and irradiated fuel samples were characterized using advanced characterization capabilities in the Irradiated Materials Characterization Laboratory (IMCL) of INL. This article summarizes the results of advanced characterization and computer vision-based materials informatics to reveal the irradiation effects on U-Zr metallic fuel. Future work will focus on further implementation of advanced characterization and statistical data mining to improve the fidelity of fuel performance modeling and support U-Zr metallic fuel qualification for fast spectrum reactors.

Phase evolution of as-cast U-35 wt% Zr and U-50 wt% Zr alloys during thermal cycling (303–1073 K) was investigated using in-situ neutron diffraction . Analysis was performed using Rietveld crystal and microstructure refinements from time-of-flight neutron diffraction data, with a focus on evolution of lattice parameter , atom ordering, unit cell volume, and weight fractions during the thermal cycling. Disordered δ-UZr 2 and residual γ-(U,Zr) phase are retained in the field of δ of the U-Zr equilibrium phase diagram for the U-50Zr sample, and in the field between δ and α-U for the U-35Zr sample. The α phase that is present in the reported phase diagrams of U-Zr was not observed in the diffraction patterns. The evolutions of lattice parameter and unit cell volume of δ and γ are affected by both thermal expansion and chemistry.

To support the development of U-10 wt.% Zr (U-10Zr) metallic fuel for Generation IV sodium-cooled fast reactors, we analyzed a Na-bonded solid U-10Zr fuel cross-section that was irradiated to a burnup of approximately 13.2 at.% at the Fast Flux Test Facility (FFTF). Advanced characterization techniques, including site-specific sample preparation by focused ion beam (FIB) and scanning transmission electron microscopy (STEM), were used to reveal the Zr redistribution and characterize the fuel matrix and secondary phases (such as solid fission products) present at the end of life. Results showed that the fuel pin cross-section is divided into three major concentric zones: a Zr-rich central region, a Zr-lean intermediate region, and a Zr intermediate peripherical region. The phase characterization revealed that the irradiation environment enhanced the development and stabilization of phases not predicted by the standard equilibrium U-Zr phase diagrams . Comparing the current results with the ones from previous studies, it is reaffirmed that the radial temperature profile and the time spent in the reactor, rather than the fuel burnup, are the two factors that most influence the formation of redistribution zones and their extension along the fuel cross-section. Various solid fission products , such as lanthanides , ZrRu, BaTe, CsI, and Ba and Sr oxides, precipitated inside the fission gas pores. This study provides unprecedented nanoscale understandings in the irradiated U-10Zr fuel system that may benefit fuel performance modelling and advanced fuel development.

The Zr-rich rind that forms at the fuel-cladding interface in U-Zr nuclear fuels has been known for many years, although it has never been explained. It acts as an effective diffusion barrier preventing the diffusion of fission product lanthanides to the cladding, but the Zr rind breaks down and does not remain intact throughout irradiations. The current study is the investigation of the Zr rind formed in U-10Zr (wt. %). A series of experiments varied the conditions to explore the effects of heat and force on the alloy. The samples were analyzed using scanning electron microscopy and transmission electron microscopy to determine the microstructural changes and phases present. Based on the results presented, a mechanism for the formation of the Zr rind is proposed along with the Zr rind structure. The Zr rind is the metastable face-centered cubic (FCC) Zr phase formed due to δ-UZr2 decomposing and subsequent diffusion of Zr to the interface when the sample is subjected to both a compressive force and heat. If either heat or force are removed from the experiment, the FCC Zr rind does not form.

UZr based metallic nuclear fuel is the leading candidate for next-generation sodium-cooled fast reactors in the United States. US research reactors have been using and testing this fuel type since the 1960s and accumulated considerable experience and knowledge about the fuel performance. However, most of knowledge remains empirical. The lack of mechanistic understanding of fuel performance is preventing the qualification of UZr fuel for commercial use. This paper proposes a data-driven approach, coupled with advanced post irradiation examination, powered by machine learning algorithms, to facilitate the development of such understandings by providing unpreceded quantified new insights into fission gas bubbles. Specifically, based on the advanced postirradiation examination data collected on a neutron-irradiated U-10Zr annular fuel, we developed a method to automatically detect, classify ~19,000 fission gas bubbles into different categories, and quantitatively link the data to lanthanide transpiration along the radial temperature gradient. The approach is versatile and can be modified to study different coupled irradiation effects, such as secondary phase redistribution and degradation of thermal conductivity, in irradiated nuclear fuel.

This work describes the microstructural and elemental characterization of irradiated metallic fuels containing palladium as an additive. The use of additives has been proposed to control Fuel-Cladding Chemical Interaction (FCCI) and thus to promote higher fuel utilization (i.e., higher burnup). In this work, Pd has been investigated as a potential additive to metallic fuel to bind lanthanides, impeding their migration and attack on the cladding. The influence of Pd on the microstructure, chemistry and performance of metallic fuel has been characterized via scanning electron microscopy for two metallic fuel designs—namely, annular and solid fuel. Pd was observed to play an important role in the chemistry of the fuel. Indeed, the addition of Pd leads to the formation of new phases. Pd was detected to combine not only with the lanthanides, as intended, but also with Zr, a main element of the fuel matrix. While Pd proved to be effective in preventing lanthanide migration and their attack on the cladding, the Pd-Zr compound may potentially lead to other unexpected fuel-performance issues, such as the formation of low-melting point phases and increased unalloyed U available for FCCI interaction with Fe in the cladding. Even the increase of Zr to 13wt%. did not completely mitigate this adverse phenomenon generated by the Pd-Zr interaction. Thus, the efficacy of using this additive needs further investigation.

Quaternary fuel alloys containing U, Nb/Mo, Ti, and Zr are proposed as fuel candidates for sodium-cooled fast reactors (SFRs). In this work, two Nb-bearing alloys, i.e., U-NT5Z (U-2.5Nb-2.5Ti-5.0Zr in wt%) and U-NT7Z (U-1.5Nb-1.5Ti-7.0Zr in wt%), and two Mo-bearing alloys, i.e., U-MT5Z (U-2.5Mo-2.5Ti-5.0Zr in wt%) and U-MT7Z (U-1.5Mo-1.5Ti-7.0Zr in wt%) were characterized and compared. The characterization techniques were differential scanning calorimetry (DSC), X-ray powder diffraction (XRD), and scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS). DSC was performed to obtain the transition behaviors, and XRD and SEM/EDS were applied for phase identification. The results were combined to obtain the solid-state phase transitions between 500 degrees C and 850 degrees C for the alloys. It is found that the Nb-bearing alloys comprise similar phase transition behaviors as the Mo-bearing alloys. The phase transitions in U-NT5Z are alpha -> gamma at 608 degrees C and U2Ti -> gamma at 627 degrees C, which are similar to 40 degrees C higher than that of U-MT5Z. The phase transition in U-NT7Z is alpha -> gamma at 645 degrees C, and is 23 degrees C higher than that of U-MT7Z.

Journal Article Understanding Fission Gas Bubble Distribution and Zirconium Redistribution in Neutron-irradiated U-Zr Metallic Fuel Using Machine Learning Get access Fei Xu, Fei Xu Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Lu Cai, Lu Cai Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Daniele Salvato, Daniele Salvato Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Fidelma Dilemma, Fidelma Dilemma Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Jeffrey J Giglio, Jeffrey J Giglio Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Michael Benson, Michael Benson Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Daniel J Murray, Daniel J Murray Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Cynthia A Adkins, Cynthia A Adkins Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Joshua J Kane, Joshua J Kane Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Min Xian, Min Xian University of Idaho, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar ... Show more Luca Capriotti, Luca Capriotti Idaho National Laboratory, Idaho Falls, ID, USA Search for other works by this author on: Oxford Academic Google Scholar Tiankai Yao Tiankai Yao Idaho National Laboratory, Idaho Falls, ID, USA Corresponding author: tiankai.yao@inl.gov Search for other works by this author on: Oxford Academic Google Scholar Microscopy and Microanalysis, Volume 28, Issue S1, 1 August 2022, Pages 82–83, https://doi.org/10.1017/S1431927622001234 Published: 01 August 2022

U-Pu-Zr metallic fuels are important fuel candidates for future advanced and/or test reactors. To better understand the fuel performance and guide future fuel design, a U-20Pu-10Zr (in weight) metallic fuel irradiated in the Experimental Breeder Reactor-Ⅱ was revisited using advanced electron microscopies. This fuel cross section was irradiated to a burnup of 6.15 % at a cladding temperature of ∼500°C. Several transmission electron microscopy samples were extracted from different but representative radial locations of the fuel cross section using focused ion beam technique. Phase identification and chemical analysis with sub-micron spatial resolution were carried out using a scanning transmission electron microscopy. Multiple phenomena that are critical to fuel performance, such as Zr redistribution, were revealed in an unpreceded highly detailed manner. An improved understanding of fuel reconstruction in U-Pu-Zr metallic fuel under reactor irradiation is provided.

Palladium is being investigated as a fuel additive to bind with and potentially immobilize lanthanide fission products. A primary cause of fuel-cladding chemical interaction (FCCI) is the lanthanide fission products migrating to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. Palladium has previously been identified as a promising additive used to prevent or decrease FCCI by reacting with the lanthanide fission products. In the current study, an alloy cast from the four highest abundant lanthanides found in irradiated metallic fuel, Nd, Ce, Pr, and La, with and without Pd, has been characterized using neutron diffraction, scanning electron microscopy, and electron probe microanalysis. In the lanthanide-Pd intermetallic compounds, all of the constituent compounds, i.e. Nd-Pd, Ce-Pd, La-Pd and Pr-Pd are known. There is very good agreement, both structurally and compositionally, between the out-of-pile lanthanide alloy and lanthanide fission products characterized in irradiated fuels. In both cases, the lanthanide elements form a solid solution in a hexagonal crystal structure. The out-of-pile lanthanide alloy follows Vegard's Law, with the measured and calculated (weighted average of constituents) lattice parameters being within 1% for both the a and c parameters. Pd bonds with the lanthanides (Ln) forming the phases LnPd and Ln7Pd3. The results indicate the properties of lanthanide compounds in irradiated metallic fuel can be reliably simulated in out-of-pile experiments.