Hydrides that retain their hydrogen in service are critical for advanced microreactor and space nuclear systems. Hydrogen retention is affected by evolving microstructure under temperature and irradiation extremes. This study investigated microstructural features in neutron-irradiated yttrium hydride and the stability of these features under thermal cycles. Differential scanning calorimetry was used to determine hydrogen desorption and temperatures at which significant phase changes are occurring, and transmission electron microscopy was used for microstructural analysis after each thermal cycle. Cyclic heating led to crystallization and epitaxial growth of surface oxidation and dehydriding of the matrix. Other features such as the bulk matrix crystal structure, irradiation-induced cavities, and matrix precipitation remained constant after thermal cycling. Results allow us to design better hydride alloys through microstructure tailoring and potential irradiation conditioning treatments.

To elucidate radiation defect processes in SiC, Raman spectroscopy was systematically applied to high-purity, polycrystalline beta-SiC that was neutron irradiated at a range of temperature and dose conditions. The analysis specifically focused on formation of carbon homonuclear bonds by irradiation; these bonds were indicated by D and G bands and amorphous carbon peaks. Intensity of the carbon peaks relative to SiC peaks significantly decreased in the case of high temperature and/or high neutron dose of 500 degrees C to 29 displacements per atom (dpa) and about 800 degrees C to 1.38 and 29 dpa. The absence of carbon bond peaks under those conditions was explained by growth of stoichiometric defect clusters, consistent with previous atomistic simulations on SiC defect stability. The lack of Raman bands associated with carbon clusters under high-temperature and high-dose radiation conditions accounts for the resistance of SiC to phase separation under irradiation. The findings further suggest that material compositions and chemical properties that are inherently resistant to chemical disordering under high-dose radiation conditions are indicative of the long-term durability of ceramic compounds in radiation environments.

The neutron irradiation changes the apparent thermal transport of substoichiometric yttrium hydride (YH x ) through coupled effects of defect production and hydrogen redistribution. Below about 400°C, irradiated specimens generally exhibit lower thermal diffusivity than unirradiated material, whereas the higher-temperature response tends to recover toward the unirradiated trend. In the 1/(A + BT) representation, irradiation primarily affects the A term, consistent with enhanced temperature-independent scattering from irradiation-induced defects, while the B term remains only weakly affected. Differential scanning calorimetry (DSC) results further show that the second-order transition temperature evolves systematically with thermal history and hydrogen state, supporting its use as an indirect H retention metric for hydrogen retention and redistribution. Together, these observations indicate that the apparent thermophysical behavior of irradiated YH x is controlled by the interplay of defect microstructure, reversible hydrogen trapping and release, and bulk dehydrogenation.

Functional hydrides are promising candidates for advanced nuclear reactors, particularly in portable or transportable applications, due to their high hydrogen-retention capabilities, enabling efficient neutron moderation, and compact reactor design. However, hydrogen mobility in hydrides at elevated irradiation temperatures poses significant technological challenges, necessitating a comprehensive understanding of their irradiation behavior. This study investigated the microstructural and chemical stability of neutron-irradiated yttrium hydrides to assess their hydrogen-retention capacity. A targeted literature review was also conducted to contextualize neutron-irradiation effects on functional hydrides with regards to structural stability and hydrogen retention. Experimental characterizations revealed that, at high temperatures, irradiated hydrides retained their phase stability, which was likely enhanced by irradiation-induced microstructure evolution. Notably, an amorphous yttrium and oxygen -rich surface layer was present at the free surface of the hydride. Its thickness decreased while a continuous crystalline Y-O-rich layer was formed with increasing neutron damage. Additionally, the number density of dislocation loops and cavities generally increased as a function of neutron dose. First-principles calculations of hydrogen behavior within yttrium vacancy clusters in yttrium hydrides demonstrated vacancy-size-dependent hydrogen stability and configuration, highlighting the role of vacancy geometry in regulating hydrogen retention. Thus, the presence of irradiation-induced dislocation loops and cavities were hypothesized to improve hydrogen retention. Collectively, these findings advance the understanding of hydride behavior under neutron irradiation as well as their technological readiness for portable or transportable nuclear reactors.

For the next-generation high temperature microreactors, yttrium dihydride (YH2) is an attractive solid state neutron moderator. Despite a number of recent investigations, the mechanism of hydrogen transport remains poorly understood. Experimental evaluations of diffusivity are inconclusive with large variations in diffusivities and activation energies. In this work, we perform ab initio molecular dynamics (AIMD) simulations on YH2 for temperatures spanning 300 K to 1200 K. Our main finding is that YH2 shows a superionic-like behavior with hydrogen atoms hopping from one native site to another above a characteristic temperature of 800 K. This correlated motion results in quasi-one-dimensional string-like displacements that enable the hydrogen atoms to diffuse rapidly. We confirm that the octahedral sites are mostly unoccupied, although channeling through them is the most favored pathway between lattice hops above 800 K. At the highest temperature of 1200 K, the string relaxation time is merely of the order of a few picoseconds, which indicates a liquid-like diffusive behavior. Based on the formation of spontaneous thermal vacancies, an order-disorder crossover temperature Tα ~ 800 K is established for YH2 with an activation energy of 0.83 eV for hydrogen diffusion in the superionic-like state.

The impact of the neutron-displacement damage on phase stability and microstructure of substoichiometric yttrium dihydrides (YHx, x <2) were investigated to assess their use as solid moderator in high-temperature nuclear reactors. YHx specimens were, thus, subjected to neutron irradiations in the range of 0.1-2 displacements per yttrium atom (dpa-Y) in the temperature range of 536-878 degrees C at the Oak Ridge National Laboratory's (ORNL's) High Flux Isotope Reactor (HFIR). YHx specimens were initially prepared at stoichiometry (H/Y) ratios of 1.69 and 1.83. HFIR-irradiated specimens were characterized by variety of techniques to investigate H retention characteristics including dimensional analysis, optical microscopy, scanning electron microscopy electron back scatter diffraction (EBSD), transmission electron microscopy, thermal desorption spectroscopy (TDS), and high-energy x-ray diffraction (HE-XRD) characterizations. Overall, YHx exhibited notable structural and phase stability under short-term neutron-irradiation, except for the samples with significant silicon carbide (SiC) interaction at high doses and temperatures. Basic dimensional and mass measurements were misleading for accurate assessment of H retention, as confirmed by EBSD phase maps, XRD line profiles, and TDS signals. Thus, it was discussed that a robust H retention metric is needed to assess irradiated hydrides. Nanoscale cavities were observed as a result of the neutron irradiation in all samples. Although no clear impact of dose and irradiation temperature was determined, the initial H/Y ratio had an impact on the cavity number density where low H/Y specimens had high-resistance to cavity formation. The Y-vacancy cluster formation at the collision stage of the displacement cascade and their stabilization by H were considered to be the likely underlying mechanisms for the observed cavity microstructure.

Irradiation experiment campaigns are critical to advancing nuclear energy technologies by providing data on material performance under relevant radiation conditions. Successful irradiation experiments require integrated design efforts that balance technical goals with facility constraints. This paper presents an expert-informed overview of irradiation experiment design at the High Flux Isotope Reactor. It addresses the nuclear materials research and irradiation experiment communities to guide them toward developing technically sound, facilitycompatible campaigns. The High Flux Isotope Reactor is a multipurpose reactor supporting isotope production, neutron scattering, and materials testing. Its high, steady-state neutron flux is ideal for irradiation experiments, but successful execution demands coordinated thermal, structural, and reactor physics analyses. The paper outlines the complete development workflow from concept definition and design optimization to safety qualification and post-irradiation examination. Standardized capsule platforms are also discussed in terms of flexibility, specimen capacity, and thermal performance. Common failure modes such as unanticipated geometric variations, can impact temperature-dose profiles and compromise data reliability. Therefore, detailed thermal modeling and accurate as-built characterization are essential for meaningful post-irradiation data interpretation. Key recommendations include early engagement all stakeholders, clearly defined design expectations, and alignment of specimen geometries with post-irradiation examination capabilities. This approach reduces design iterations, enhances data quality, and supports more efficient use of irradiation resources. Strategic and well-planned irradiation testing not only improves individual campaign success but also accelerates the deployment of advanced nuclear technologies. By closing critical data gaps and reducing development risks, the nuclear materials community can more effectively contribute to the future of clean, resilient energy systems.

Iron-chromium-aluminum (FeCrAl) class alloys are candidates for use as cladding for accident-tolerant fuels and moderators. In this context, hydrogen isotope permeation in FeCrAl alloys is an important material property. In the present work, the apparent permeability, effective diffusivity, and apparent solubility of hydrogen in the FeCrAl alloys C26M and Kanthal D (KD) were measured with gas-driven hydrogen permeation. Permeation measurements were conducted at temperatures of 400 to 700 degrees C and at gas-driven pressures from 1 to 100 kPa. In particular, the effect of grain size on hydrogen transport was studied with KD samples with three different microstructures: nanocrystalline (NC), ultra-fine grained (UFG), and coarse-grained (CG). The UFG and NC specimens had higher apparent activation energies (73.4 kJ mol(-1) and 65.2 kJ mol(-l), respectively) for hydrogen permeability than the CG sample (46.9 kJ mol(-1)). An aluminum oxide layer formed on the primary- and secondary-side surfaces of all samples subjected to permeation experiments which demonstrated the propensity of FeCrAl alloys to form these innate oxide permeation barriers.

The response of silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composite cladding to mechanical interaction with fissile fuel is a knowledge gap that must be overcome to design and assess SiC-based cladding systems for advanced nuclear applications. This study developed the relevant mechanical testing capability and identified the failure behavior and the critical microstructural features and processing defects. Sections of SiC composite tube were subjected to a modified expansion-due-to-compression (EDC) test in an X-ray computed tomography microscope: a polyurethane plug pressed surrogate Al2O3 into the inner walls of the SiC/SiC composite tubes to achieve hard contact. A pure EDC test with just a polyurethane plug was also performed as a reference. Through the use of displacement fields, digital volume correlation revealed inhomogeneous deformation fields in the tubes, even for pure EDC, which was related to the inherent defects in the structure. Deep learning-aided segmentation and systematic data analysis revealed that the presence of inhomogeneous deformation applied by the hard contact was exaggerated by the presence of inner surface imperfections left behind from the matrix densification process. The findings provide insights into the applications, highlighting the necessity for improvements in inner surface roughness and the incorporation of localized contacts in pellet--cladding mechanical interaction computational models.

The impurity or alloying atoms in YH2 can alter the local electronic structure and so the hydrogen defect stability, as well as the H migration barrier energy. Thus, DFT calculations were employed to determine the effect of foreign elements from alkali and alkaline earth metals to transition metals and one critical impurity element, O, on H vacancy stability and retention characteristics in YH2. Results revealed that alloying elements act as hydrogen vacancy sinks by reducing the vacancy formation energy at neighboring sites. The implantation of nonmagnetic foreign elements (s1, s2, and d10 valence electrons) in hydrogen energy landscape was calculated to be minor; while the hydrogen vacancy formation energy was reduced from 1.37 eV to 1.00 eV, the migration energy barrier of hydrogen was increased from 0.87 eV to 1.15 eV for non-magnetic foreign elements. The migration energy barrier monotonically decreased with increasing d-shell occupancy, reaching as low as 0.4 eV for Cr, Mo (d4), and Fe (d4). Alloying with late transition metals (d8 and d9) moderately impacted the hydrogen vacancy formation. Finally, it was found to be O addition into the YH2- lattice did not alter the energy landscape of hydrogen vacancies. Since alloyed YH2 has not been studied extensively, this study provides an atomistic understanding how alloying elements and impurities trap vacancies and affects hydrogen mobility YH2. Meanwhile, the main findings of this study may serve as guidelines for introducing alloying elements in ZrH2 as well.

The high-temperature motion of hydrogen in near stoichiometric yttrium dihydride (YHx, x=1.62 and 1.87 at.%) was investigated using incoherent quasi-elastic neutron scattering and Density Functional Theory (DFT) calculations as a function of hydrogen stoichiometry. Translational motion (diffusivity) of hydrogen in yttrium dihydride was only observed in a temperature range of 1073-1173K under vacuum environment. The hydrogen motion was found to be limited to the tetrahedral sublattice, and diffusivity of hydrogen was observed to increase with decreasing hydrogen stoichiometry. The same behavior was also supported with DFT calculations. The DFT results also indicated that certain migration paths with smaller energy barriers favored H jump resulting in higher diffusivities.

The development of microreactor technology presents an efficient solution for providing portable electricity, catering to both human space exploration needs within our solar system and supplying power to remote Earth-bound areas. The miniaturization of nuclear reactors poses immediate new challenges for materials science with respect to the capability for controlling nuclear reactions via thermalization of highly-energetic neutrons. In a microreactor, neutron moderation takes place in compact geometries, thus new moderator materials are required to exhibit high moderating power per unit of volume. This challenge is currently being addressed through the development of transition metal hydrides, known for their strong nuclear moderation capability but to date, research on their irradiation response is limited, specifically regarding phase stability, hydrogen in-lattice retention, and their dependence on irradiation temperature and dose. Herein, we present a detailed investigation on the response of yttrium dihydride (YH2) 2 ) to heavy ion irradiation. The experiments indicate that YH2 2 is stable up to an irradiation dose of 2 dpa and below 800 degrees C, degrees C, identified herein as a critical temperature for YH2. 2 . Our study detected the nucleation and growth of voids as a function of the irradiation temperature. They were the predominant type of radiation damage present in the microstructure of YH2 2 that was distinguishable from pre-existing defects in the pristine YH2 2 samples. Below the critical temperature, no phase transformation (degassing/dehydriding) nor amorphization occurred. Experimental results with concomitant density functional theory calculations allowed us to elaborate and propose new strategies to enhance the metal hydride performance in extreme environments.

In situ synchrotron high-energy x-ray diffraction experiments and detailed transmission electron microscopy (TEM) characterization were conducted on as-fabricated and neutron-irradiated yttrium hydrides. The high-resolution synchrotron x-ray diffraction revealed minor α yttrium and major δ yttrium hydride phases in all specimens. Specimens were subject to heat treatments (heating-cooling cycles), and the intensity of α yttrium partially and completely disappeared in as-fabricated and neutron-irradiated specimens, respectively. The disappearance of α yttrium was unforeseen because hydrogen was expected to leave δ phase, causing an increase in α yttrium diffraction peak intensity. This observation indicated a surplus of hydrogen in the specimens where it was odd for hydride-forming early transition metal elements. The subsequent through-focus TEM characterization discovered nanometric cavities in both as-fabricated and neutron-irradiated yttrium hydride specimens for the first time. Two types of cavities were identified as fabrication-caused and irradiation-induced. The fabrication-caused cavities were associated with regions having linear deformation features, interfaces, and inclusions. The irradiation-induced cavities were observed as being formed isolated in the yttrium hydride phase. The presence of such nanometric cavities was considered as potential hydrogen storage pockets where the overall hydrogen storing capacity of yttrium hydride would be enhanced.

Modeling the full-range deformation behaviors of materials under complex loading and materials conditions is a significant challenge for constitutive relations (CRs) modeling. We propose a general encoder-decoder deep learning framework that can model high-dimensional stress-strain data and complex loading histories with robustness and universal capability. The framework employs an encoder to project high-dimensional input information (e.g., loading history, loading conditions, and materials information) to a lower-dimensional hidden space and a decoder to map the hidden representation to the stress of interest. We evaluated various encoder architectures, including gated recurrent unit (GRU), GRU with attention, temporal convolutional network (TCN), and the Transformer encoder, on two complex stress-strain datasets that were designed to include a wide range of complex loading histories and loading conditions. All architectures achieved excellent test results with an rootmean-square error (RMSE) below 1 MPa. Additionally, we analyzed the capability of the different architectures to make predictions on out-of-domain applications, with an uncertainty estimation based on deep ensembles. The proposed approach provides a robust alternative to empirical/semi-empirical models for CRs modeling, offering the potential for more accurate and efficient materials design and optimization.

Yttrium hydrides are considered as candidate materials for neutron moderation applied in microreactors (akin transportable miniaturized nuclear reactors) owing to their superior thermal stability and hydrogen retention. The evolution of elastic properties of these materials at elevated temperatures, needed for predicting the thermomechanical response and performance of the moderator during in-service reactor conditions, however, is lacking. Here, we report the Young’s and shear elastic moduli of three stoichiometries of bulk yttrium hydride (YHx, x = 1.61, 1.82, and 1.84) from room temperature to 1000°C. In situ temperature-dependent measurements of the longitudinal and shear wave velocities were performed using a laser ultrasonic technique while heating the sample in a vacuum-pumped heating stage. The elastic moduli increased linearly with increasing hydrogen content and decreased by ~10% during heating from room temperature to 1000°C in the three YHx compositions. The linear relationship between the elastic moduli and the hydrogen content in yttrium hydride was verified by atomistic calculations based on density functional theory (DFT). The absence of abrupt changes in the temperature-dependent measurements of elastic modulus of the YHx samples suggested negligible loss of hydrogen at elevated temperatures. Excellent agreement was found between the measured and calculated dependence of the elastic moduli on the stoichiometry, thereby providing a new approach for investigating the effects of fabrication-induced parameters (such as porosity) on the elastic moduli. This study demonstrates the utility of the combined approach involving DFT-based atomistic calculations and measurements of the elastic moduli for the informative development of metal hydrides and can be used as a metric for novel moderator materials investigations for emerging microreactors and beyond.

This work investigates the micromechanical deformation and failure mechanisms of advanced FeCrAl alloys developed for use in nuclear reactors as cladding material. Three different FeCrAl alloys (82-X)Fe-13Cr-5Al-X (X=Nb, TiC, 0 for base alloy) were investigated both in their as-received and welded states. In-situ neutron diffraction with simultaneous digital image correlation was used to determine micromechanical deformation mechanisms not only as a function of elemental composition but also as a condition of state (as-received vs. welded). Ex-situ X-ray computed tomography was used on as-deformed samples to help determine the failure mechanisms and void initiation strains.

Solid state hydrides such as early transition metal hydrides are of inestimable importance for the future of hydrogen energy and are actively being investigated for energy conversion and storage applications such as fuel cells, solid-state batteries and neutron moderators. The retention and transport behavior of hydrogen in these hydrides has a huge role on the extended performance of components. While early transition-metal-based compounds exhibit many peculiar properties due to their unique correlated electronic signatures arising from d-orbital electrons, the fundamental chemistry and transport behavior of hydrogen in such hydrides is not well understood. In the present work, using density functional theory, a highly intricate bonding feature is revealed through the theoretical investigation of the electronic structure of early transition metal hydrides YH2 and ZrH2. In particular, a pronounced charge transfer from the transition element to H, results in localized electron densities at deep energy levels. The interplay between intrinsic charge transfer, charge localization, and metallicity in YH2 and ZrH2 leads to strong chemical bonding between metal and hydrogen atoms and large energy barriers for the migration of hydrogen vacancies. Specifically, hydrogen vacancies are found to be stable in the neutral state due to electron screening effects, accompanied by substantially high migration barriers between 0.8-1.2 eV along different crystallographic directions. In contrast, recent literature shows the migration barrier for charged H vacancies in insulating s-block metal hydrides lie between 0.1-0.4 eV, which is suitable for fast conduction applications. This pivotal electron structure difference exploited between early transition metal hydrides and alkali/alkaline earth metal hydrides determines extended hydrogen retention in these early transition metal hydrides. This work explains fundamental differences between the electronic structure of s-block and d-block metal hydrides, and its impact on the mobility of hydrogen vacancies. Published by Elsevier Ltd on behalf of Hydrogen Energy Publications LLC.

Journal Article Advanced Crack Analytics on 3D X-ray Tomography of Irradiated Silicon Carbide Claddings Get access Fei Xu, Fei Xu Idaho National Laboratory, Idaho Falls, ID, United States Corresponding author: fei.xu@inl.gov Search for other works by this author on: Oxford Academic Google Scholar Joshua J Kane, Joshua J Kane Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Peng Xu, Peng Xu Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Nikolaus Cordes, Nikolaus Cordes Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Jason L Schulthess, Jason L Schulthess Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Mahmut Nedim Cinbiz, Mahmut Nedim Cinbiz Idaho National Laboratory, Idaho Falls, ID, United States Search for other works by this author on: Oxford Academic Google Scholar Sean Gonderman, Sean Gonderman General Atomics, San Diego, CA, United States Search for other works by this author on: Oxford Academic Google Scholar Christian Deck, Christian Deck General Atomics, San Diego, CA, United States Search for other works by this author on: Oxford Academic Google Scholar Jack Gazza Jack Gazza General Atomics, San Diego, CA, United States Search for other works by this author on: Oxford Academic Google Scholar Microscopy and Microanalysis, Volume 28, Issue S1, 1 August 2022, Pages 208–210, https://doi.org/10.1017/S1431927622001696 Published: 01 August 2022

Stress-driven microstructure evolution imposes significant challenges on mechanical behavior modeling. Here, a new artificial neural network architecture, consisting of a temporal convolutional neural network and fully connected neural network (TCN-FCN), was applied to learning complex stress-strain data. The causal convolution operation implemented in this TCN-FCN architecture can correlate the most informative loading history information to the current stress state in a high-dimensional material parameter space. The TCN-FCN model was benchmarked against a similarly motivated gated recurrent unit (GRU)-based recurrent neural network model and showed ~50% error reduction in modeling complex loading histories. Such TCN-FCN architecture demonstrates excellent generalization ability and universal capability in modeling high-dimensional complex stress-stress data (e.g., temperature, strain rate, materials conditions, etc.), thus offering a robust alternative to conventional empirical/semi-empirical models for constitutive relationship modeling and materials optimizations.