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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Lattice response to ambient self-irradiation of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures

Self-irradiation in plutonium is a continuous process that introduces radiation damage ranging from point defects to nanoscale radiation damage as well as decay daughter product impurities that require to be accommodated by the host material. These changes are affecting the structure and properties of the material by lattice swelling, mechanical strengthening, and other alterations. X-ray diffraction and dilatometry studies have shown that the room temperature self-irradiation of δ-phase Pu-Ga alloys (2 and 7 at%Ga) results in swelling of the lattice that saturates after 0.1–0.2 displacement per atom (dpa) of accumulated lattice damage. Here, for a better understanding of these self- irradiation processes, defect accumulation and damage evolution, this time-of-flight neutron diffraction study investigates the impact of ambient self-irradiation effects on the structural evolution of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures (<120 K), associated with a varied history of storage time and thermal treatments. Two key behaviors were observed in the lattice evolution after exposure to cryogenic temperatures with respect to room-temperature accumulated self-irradiation damage. Samples with no self- irradiation damage accumulation under ambient conditions showed swelling of the lattice at the end of the measurement, while samples that accumulated significant self-irradiation damage under ambient conditions showed only a partial recovery of the lattice parameter value upon completion of the measurements (cooling followed by heating back to room temperature). Possible causes of the behavior are proposed, and their associated effects are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Thermal conductivity suppression in uranium-doped thorium dioxide due to phonon-spin interactions

In this work, impact of low level of uranium (U) atom substitution on thermal conductivity of thorium dioxide (ThO 2 ) is investigated. ThO 2 is an electronic insulator with a wide optical band-gap and no unpaired electrons whose thermal transport is governed by phonons. U-substitution introduces unpaired f-electrons resulting in paramagnetic behavior of U-ThO 2 at room temperature, which significantly suppresses its thermal conductivity. A single crystal of U-ThO 2 with graded composition of U is grown using a hydrothermal synthesis method, and thermal conductivity measurements are performed in regions with uniform composition of U at levels of 0%, 6%, 9% and 16%. Measured thermal conductivity profiles over 77–300 K temperature range are analyzed using an analytical expression for phonon-mediated thermal transport based on Klemens-Callaway model. Temperature dependent thermal conductivity is found to deviate significantly from the Rayleigh scattering trend expected for a simple substitutional point defect with a small perturbation to mass and interatomic forces. With the resonant scattering term, observed large suppression of thermal conductivity at low temperatures can be closely reproduced. Additionally, the extracted phonon-spin coupling constants imply a nonlinear relation of phonon-spin interaction intensity with respect to U doping percentage. Our study reveals how phonon-spin scattering contributed by unpaired f-electrons in U atoms influences thermal transport in the U-ThO 2 system.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Investigation of the irradiation effects in additively manufactured 316L steel resulting in decreased irradiation assisted stress corrosion cracking susceptibility

While additive manufacturing (AM) is an area of increasing interest to the nuclear industry, there are significant research and development needs prior to successful implementation of additively manufactured components into reactors. AM steel behaves differently than traditional wrought material, as has been observed in numerous mechanical testing studies. The nuclear industry must also understand the material response to irradiation. This work examines the irradiation damage and the irradiation assisted stress corrosion cracking behavior (IASCC) of both proton irradiated AM and wrought 316L steel. For IASCC testing, specimens were strained to 4% plastic strain in a simulated boiling water reactor environment (288°C, 0.2 µS/cm). Transmission and scanning electron microscopy, confocal microscopy, and X-ray computed tomography were used to examine coupons of the material as well as tensile bars used in the IASCC testing. This work found that the wrought 316L was more susceptible to IASCC than the AM steel, however, some IASCC was observed in the AM steel when tested with the tensile axis parallel to the build direction. Electron microscopy has shown significantly fewer radiation induced voids formed in the AM steel. While the pre-existing pore network in the AM steel may act as a sink to point defects, this is not expected to be the major factor in the reduced swelling observed in the AM steel. The lower density of radiation induced voids was expected to contribute to the IASCC resistance by decreasing the localization of deformation in the irradiated specimens, however, severity of localized deformation in AM steel, measured in dislocation channel spacing and height, was similar or more severe to that found in wrought. In conclusion, the presence of severe cracking in the wrought specimens did not allow for a direct comparison of localized deformation severities between the AM and wrought steel.

Additive manufacturing↗
DOE OSTI2021

A density functional theory and neutron diffraction study of the ambient condition properties of sub-stoichiometric yttrium hydride

Several mechanical and thermophysical properties are required as a function of non-stoichiometry for the successful implementation of YH 2-x for nuclear reactor moderator applications. Density functional theory calculations, in combination with neutron diffraction experiments, were used to study the structural and mechanical properties of YH 2-x . Point defect analysis indicated H occupation primarily at the tetrahedral site within an fcc Y sub-lattice, confirming a fluorite YH 2 structure. The small positive formation energy for H vacancies under Y-rich conditions predicted that hypo-stoichiometry is accommodated by Y+YH 2 at ambient conditions and by H vacancies in the YH 2-x single phase that is relevant to high temperatures. Neutron diffraction studies were used to confirm both the occupation of H on tetrahedral sites and the near-stoichiometric composition of the hydride phase in the two-phase Y+YH 2 region of the phase diagram that dominates at room temperature. Energy minimized special-quasirandom-structures of H vacancies were used to calculate lattice parameters, elastic constants, and several other properties as a function of composition for the single phase YH 2-x . The lattice parameter of YH 2-x decreased by only 0.004 Å with increasing H/Y for the range 1.31 ≤ H / Y ≤ 2.0 indicating a negligible effect on lattice parameters due to vacancy formation. In the two-phase region, however, calculations predicted the density of z Y + ( 1 - z ) Y H 2 - x to increase with decreasing H/Y at lower temperatures due to the increased fraction of high-density Y metal. Additionally, for the high temperature single phase, decreasing H/Y reduced the density as a consequence of the lattice expansion associated with vacancy formation. All elastic constants and moduli increased with increasing hydrogen content in single-phase YH 2-x .

36 MATERIALS SCIENCE↗
DOE OSTI2022

Understanding radiation effects in friction stir welded MA956 using ion irradiation and a rate theory model

We report an outstanding challenge in the manufacturing and joining of oxide dispersion strengthened steels is retaining the nanofeatures in the alloy throughout the fabrication and welding process. MA956 was friction stir welded with two different sets of welding parameters, resulting in a medium and high heat input. After welding, 5 MeV Fe ++ ion irradiations were performed at doses ranging from 50 to 200 dpa in the temperature range of 400 to 500°C. Post-irradiation characterization was performed with scanning transmission electron microscopy and energy-dispersive x-ray spectroscopy to investigate the Y-Al-O dispersoids, voids, and dislocations. After welding, the dispersoid microstructure coarsened, resulting in fewer and larger dispersoids regardless of heat input. After irradiation, the dispersoid behavior in the welded material was sensitive to temperature, exhibiting growth behavior attributed to Ostwald coarsening at 500 °C but a mixture of nucleation and more muted growth at 400 and 450 °C, attributed to competing mechanisms of radiation-enhanced diffusion and Ostwald coarsening. Void swelling correlated to heat input; being more prevalent in the welded conditions occurring at lower doses and in higher values relative to the base material. The low values of swelling despite microstructure coarsening caused by welding demonstrate the excellent swelling resistance of MA956, even after welding with the highest swelling values of 0.5% noted in the stir zone high heat input condition at 450 °C, 200 dpa. The dislocation behavior was inconsistent: the strongest trend was that network density was higher for welded versus base material, and an increase in loop diameter with temperature was observed. A rate theory model based on the observed microstructure suggests at high temperature interstitial loss to sinks was more likely to be dominant compared to mutual annihilation via point defect recombination, because of an increase of the radiation diffusion coefficient with temperature regardless of initial welded microstructure.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Atomistic and cluster dynamics modeling of fission gas (Xe) diffusivity in TRISO fuel kernels

TRISO fuel particles are candidates for use in next generation reactors including gas reactors, fluoride salt-cooled high temperature reactors, and micro-reactors. The UCO fuel kernel consists of a uranium dioxide (UO) and uranium carbide mixture. The addition of UC helps suppress the formation of carbon monoxide gas, which led to failures during initial TRISO development. The addition of uranium carbide alters the chemistry of the UO kernel, which is known to influence performance parameters such as fission gas diffusivity, although the impact has not been quantified and no models exist that take the change in chemistry into account. Therefore, better understanding and more accurate models of the impact of chemistry on fuel performance are of high priority. In this paper, a first-principles density functional theory (DFT) and empirical potential based multi-scale study has been carried out to model the diffusivity of fission gas xenon (Xe) in UCO TRISO fuel kernels. The focus is on the UO component in the UCO fuel kernels, as that represents the largest volume fraction of the fuel kernels. The study relies on DFT and empirical potential calculations to determine Xe and point defect properties, which are then used in thermodynamic and kinetic models to predict diffusion for intrinsic conditions. In addition, the information is utilized in cluster dynamics simulations using the Centipede code to estimate the impact of irradiation on defect transport. Additionally, the presence of UC or UC in the UCO fuel kernels is shown to have a substantial impact on the UO non-stoichiometry by inducing oxygen vacancies and driving UO sub-stoichiometric, which causes much slower Xe diffusion in UCO compared to light water reactor UO fuel. The application of this model in fuel performance simulations using the Bison code is also demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Depth-resolved thermal conductivity and damage in swift heavy ion irradiated metal oxides

We investigated thermal transport in swift heavy ion (SHI) irradiated insulating single crystalline oxide materials: yttrium aluminum garnet- Y 3 Al 5 O 12 (YAG), sapphire (Al 2 O 3 ), zinc oxide (ZnO) and magnesium oxide (MgO) irradiated by 167 MeV Xe ions at 10 12 – 10 14 ions/cm 2 fluences. Depth profiling of the thermal transport on nano- and micro- meter scales was assessed by time-domain thermoreflectance (TDTR) and modulated thermoreflectance (MTR) methods, respectively. This combination allowed us to isolate the conductivities of different sub-surface damage-regions characterized by their distinct microstructure evolution regimes. Thermal conductivity degradation in SHI irradiated YAG and Al 2 O 3 is attributed to formation of ion tracks and subsequent amorphization, while in ZnO and MgO it is mostly due to point defects. Additionally, notably lower conductivity when probed by very low penetrating thermal waves is consistent with surface hillock formation. An analytical model based on Klemens-Callaway method for thermal conductivity coupled with a simplified microstructure evolution capturing saturation in defect concentration was used to obtain depth dependent damage across the ion impacted region. The studies showed that YAG has the highest damage profile resulting in the less dependence of thermal conductivity with the depth, while MgO on the contrary has the strongest dependence. The presented work sheds new light on how SHI induced defects affect thermal transport degradation and recovery of oxide ceramics as promising candidates for next generation nuclear reactor applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Solute segregation and precipitation across damage rates in dual-ion–irradiated T91 steel

We report that dual-ion irradiations using 5.0 MeV defocused Fe 2+ ions and co-injected energy degraded 2.00 to 2.85 MeV He 2+ ions were conducted on a Fe9CrMo ferritic-martensitic steel T91 to 17 dpa at a damage rate range of 5 × 10 -5 dpa/s to 3 × 10 -3 dpa/s at 445°C, followed by characterization of the microstructure using conventional and scanning transmission electron microscopy. Radiation induced Ni/Si clusters and radiation induced segregation were quantified using energy dispersive X-ray spectroscopy at each condition and were compared with the same material irradiated in the BOR-60 reactor and in the as-received condition. No significant Cr segregation was found at lath boundaries after dual-ion irradiation, while Ni and Si enrichments both decreased with increasing damage rate leading to a sharp decrease in the density of Ni/Si clusters with damage rate. Increased point defect recombination at higher ion damage rates likely reduced the Ni/Si cluster density compared with BOR-60. Although the overall vacancy concentration and diffusion are enhanced by the irradiation damage rate, the lack of time for thermal diffusion and ballistic displacements of solutes are significant limiting factors for Ni/Si cluster formation. This work demonstrates the effect of irradiation damage rate on elemental segregation and clustering when using ion irradiation to simulate reactor irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Cluster dynamics modeling of Mn-Ni-Si precipitates coupled with radiation-induced segregation in low-Cu reactor pressure vessel steels

Formation of precipitates enhanced or induced by irradiation causes hardening and embrittlement of nuclear structural materials. Post-irradiation microstructure characterization of reactor pressure vessel (RPV) steels has shown that precipitation can be strongly associated with radiation-induced segregation (RIS) of solutes on dislocations. However, RIS and precipitation have not been coupled in previous precipitation modeling of RPV steels. Here in this study, a new hybrid and spatially-dependent precipitation model is developed that couples cluster dynamics with RIS, providing an unique way to account for concurrent evolution of heterogeneous cluster densities as well as solute and point defect concentration profiles. The model is applied to study the segregation of Mn, Ni, and Si on dislocations and heterogeneous nucleation of Mn-Ni-Si rich precipitates (MNSPs) in a low-Cu RPV steel. The result shows that the onset of MNSP nucleation on dislocations occurs at the fluence of 2 x 10 23 n.m -2 . The number density and mean radius can reach ~ 10 24 m -3 and 1-2 nm at high fluence, respectively. These observations suggest that the role of RIS and heterogeneous nucleation can be significant for RPV steels under a high fluence of irradiation. Results of simulations adopting various dose rates and dislocation densities show an increased fraction of MNSPs on dislocations at the condition of low dose rate, high dislocation density, and high fluence.

36 MATERIALS SCIENCE↗
DOE OSTI2023

The effect of secondary phases on microstructure and irradiation damage in an as-built additively manufactured 316 L stainless steel with a hafnium compositional gradient

Additive manufacturing (AM) or rapid prototyping has become a crucial tool for reducing both cost and time while increasing efficiency in qualifying structural materials for reactor use. In this study, directed energy deposition (DED) was used to develop an as-built 316 L stainless steel sample with three regions of increasing Hf-dopant to study the effects of Hf on the irradiation response of the material. Morphological and microstructural changes were analyzed before and after 2 MeV proton irradiation at 360 °C to a damage of 2.5 dpa at ~ 10 µm below the surface. The addition of Hf effectively suppressed radiation-induced damage (dislocation loops, radiation-induced segregation) due to enhanced point defect recombination. The radiation damage seen in the as-built sample was further compared to a thermo-mechanically treated counterpart of the same fabrication and irradiation parameters which was found to behave superiorly. In conclusion, the increased radiation resistance of this material may be attributed to the as-built microstructure, which includes undissolved Hf particles, delta ferrite grains and cellular sub-grain boundaries that can hinder defect motion.

316L Stainless Steel↗
DOE OSTI2023

Improving charge transport in integrated MoO 3 /C electrode materials for water-in-salt energy storage systems by incorporating oxygen vacancies

Improvements in the charge storage properties of α-MoO 3 used as an electrode with a 30m ZnCl 2 water-in-salt electrolyte have been achieved by enhancements in electron and ion transport enabled by an inventive synthesis route. Electron transport was improved through the integration of MoO 3 with dopamine-derived carbon via a chemical preintercalation route, and enhanced ion transport was achieved by incorporating oxygen vacancies in MoO 3 structure through ethanol 2 reduction under hydrothermal conditions. Here, the presence of carbon was confirmed by corresponding D and G bands observed in Raman spectroscopy measurements. The presence of oxygen vacancies was proven through correlated XPS, TGA, Raman spectroscopy and XRD analyses, with the introduction of oxygen vacancies leading to an expanded interlayer region. Four-point probe measurements provided evidence of increased electronic conductivity due to the incorporation of carbon, and cyclic voltammetry-based charge storage mechanism analyses revealed increases in ion transport kinetics due to oxygen vacancy formation. Tuning the oxygen vacancy concentration is critical, as excessive concentrations of these point defects leads to structural instability and poor capacity retention. This work demonstrates the combined potential of carbon and oxygen vacancies in moderate concentrations to enhance the charge storage properties of transition metal oxides. The strategies developed in this study offer a path to the development of promising materials for high-rate, high-capacity, and long-duration electrochemical energy storage technologies.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2022

The role of Cr concentration and temperature on cavity swelling with co-injected helium in dual-ion irradiated Fe and Fe-Cr alloys

The level of chromium plays an essential role in irradiation tolerance of Fe-Cr ferritic alloys. However, conflicting results have been reported regarding the dependence of cavity swelling under irradiation on Cr level and temperature. Here, we have performed a comprehensive set of simultaneous dual-ion (Ni + He) irradiations to high dose (~30 displacements per atom, dpa) at 400–550 °C on a series of ultra-high purity Fe and Fe-Cr binary alloys (3–14 wt.%Cr). Helium co-implantation rates of 0.1 and 10 appm He/dpa were selected to examine He synergistic effects relevant for fission and fusion reactor conditions, respectively. Cavities were observed in all irradiated samples by transmission electron microscopy. The results show that higher He implantation rate causes a shift in the swelling peak to higher temperatures in both Fe and Fe-Cr alloys. When assuming smaller cavities as biased sinks, the non-monotonic nature of the cavity swelling behavior is related to the ratio of biased to unbiased point defect sink strengths. Cr-enriched precipitates were observed in Fe-14Cr irradiated at 400 °C by atom probe tomography. Our analysis suggests the formation of Cr-enriched precipitates could suppress cavity swelling for Fe-Cr alloys with Cr content above 10 wt%.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Thermodynamic origin of nonvolatility in resistive memory

Electronic switches based on the migration of high-density point defects, or memristors, are poised to revolutionize post-digital electronics. Despite significant research, key mechanisms for filament formation and oxygen transport remain unresolved, hindering our ability to predict and design device properties. For example, experiments have achieved 10 orders of magnitude longer retention times than predicted by current models. Here, using electrical measurements, scanning probe microscopy, and first-principles calculations on tantalum oxide memristors, we reveal that the formation and stability of conductive filaments crucially depend on the thermodynamic stability of the amorphous oxygen-rich and oxygen-poor compounds, which undergo composition phase separation. Including the previously neglected effects of this amorphous phase separation reconciles unexplained discrepancies in retention and enables predictive design of key performance indicators such as retention stability. Furthermore, this result emphasizes non-ideal thermodynamic interactions as key design criteria in post-digital devices with defect densities substantially exceeding those of today’s covalent semiconductors.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Analysis of orientation-dependent deformation mechanisms in additively manufactured Zr using in-situ micromechanical testing: Twinning and orientation gradient

Here, orientation-dependent plasticity in zirconium (Zr)-alloy, which is produced through ultrasonic additive manufacturing (UAM) with subsequent hot-isostatic pressing (HIP), was analyzed by recording electron backscattered diffraction (EBSD) data during in-situ micromechanical testing. Three (0001) plane orientations of a hexagonal close-packed (HCP) structure were analyzed parallel to (1) the rolling direction [X||tensile direction (TD)], (2) the build direction [Z], and (3) the transverse direction [Y]. The analysis revealed that the grains with ~<0001>||TD show twin-dominant plasticity with three variants from {10 $\overline{1}$ 2}<$\overline{1}$ 011>; minor slipping with ($10\overline{1}$ 1)[$\overline{2}$ 113], and ($11\overline{2}$ 2)[$\overline{1} \overline{1}$ 23] pyramidal slip has also been observed. However, grains with orientation ~<0001>⊥TD are mainly sensitive to dislocations assisted plasticity-leading to orientation gradients formation. Furthermore, the neck formation was identified as originating from higher populated micro-crack locations and their association with localized plasticity at defect points in the UAM material. These results demonstrate that minimization of impurities to enable grain growth across prior foil interfaces makes HIP an effective methodology for producing Zr plate, with deformation characteristics expected for conventionally manufactured Zr.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Spatial variation of the low-temperature micro-photoluminescence of THM-grown CdZnTeSe

The quaternary compound CdZnTeSe (CZTS) has emerged as a next-generation detector material. Unlike CdZnTe (CZT), CZTS has the distinct advantages of little or no sub-grain boundary networks and a much lower concentration of Te inclusions, plus better compositional homogeneity. Thus, the material is expected to offer better spatial charge-transport homogeneity compared to CZT. However, an inhomogeneous distribution of point defects in both CZT and CZTS, such as Cd vacancies and residual impurities, can impose additional spatial inhomogeneity of the resistivity, carrier mobilities, and carrier lifetimes of the material. To investigate the spatial distribution of such defects in CZTS, low-temperature photoluminescence (PL) spectroscopic studies were performed at different positions along a single crystalline CZTS sample surface, which was grown by the traveling heater method (THM). In conclusion, the intensity variation of the PL emission of excitons bound to a neutral acceptor defect (A 0 , X), which is likely dominated by an acceptor like Cu-related level, with respect to the PL emission line from a neutral donor-bound exciton (D 0 , X) exhibited significant spatial variation, while the peak energy positions were approximately uniform due to the high compositional and bandgap homogeneity of the material.

II-VI semiconductor↗
DOE OSTI2021

Amorphization resistance of nanocrystalline 3C-SiC implanted with H2+ ions

Compared to single-crystal SiC, nanocrystalline SiC with high densities of stacking faults has been reported to be much more resistant to amorphization under self-ion and electron irradiations. This study examines H2+ ion irradiation-induced amorphization in nanocrystalline 3C-SiC with dense stacking faults using transmission electron microscopy. The results show that full amorphization at room temperature occurs at a comparable dose to that for its single-crystal SiC counterpart under the identical irradiation conditions. Both materials are amorphized as a result of local damage accumulation. The formation of the nucleation sites for amorphization is not appreciably affected by the presence of stacking faults and grain boundaries. The behavior may be attributed to the significant chemical effects that may completely immobilize the point defects in SiC at room temperature. The results suggest cautions be excised to use nanocrystalline SiC materials in high H irradiation environment at room temperature. Further studies of the H behavior at elevated temperatures are warranted.

Amorphization, ion irradiation, SiC↗
DOE OSTI2025

Thermal conductivity measurement of U-Mo and U-Mo/Al interaction layers generated from in-pile irradiation using the suspended-bridge method

Here, this study presents the first measurement of the individual thermal conductivity of U-7wt.%Mo fuel particles and U-Mo/Al interaction layers (ILs) from in-pile irradiated dispersion fuel plates, using the suspended-bridge method. Nanorods of U-7wt.%Mo fuel and U-Mo/Al ILs were extracted by focused ion beam (FIB), and their microstructures were characterized with transmission electron microscopy (TEM). TEM revealed finely distributed nanobubbles in the U-7wt.%Mo matrix, along with an amorphous structure in the ILs. The thermal conductivity of in-pile irradiated U-7wt.%Mo was approximately 30% lower than that of the unirradiated material, ranging from 6.7 W/m·K at 300 K to 8.5 W/m·K at 380 K. The ILs exhibited even lower thermal conductivity, from 2.1 W/m·K at 300 K to 2.7 W/m·K at 380 K. These reductions, attributed to nanobubbles, fission products, and irradiation-induced point defects, were analyzed through a combination of microstructural characterization and literature-based transport models, which successfully reproduced the observed degradation trends.

42 - ENGINEERING↗
DOE OSTI2026

Irradiation-induced grain boundary migration in an ion irradiated Fe-Ni-Cr-Mn alloy

Irradiation-induced grain boundary migration (RIGM) has previously been observed in different metallic systems. However, the mechanisms and dependence on irradiation conditions remained unclear. Here, we report on radiation-induced GB migration (RIGM) in a coarse-grain Fe-Ni-Cr-Mn austenitic alloy irradiated with Fe ions at 400 °C to 15, 35, and 70 dpa. Grain boundary migration is only observed in the irradiated regions and exhibits a linear dependence on the locally delivered dose, as estimated from SRIM calculations. Notably, the same linear dependence is noted across the nominal doses. Furthermore, the evidence points to grain boundary migration being mediated by the flux of point defects to grain boundaries and motivates additional systematic studies of RIGM in these systems.

Grain boundary↗