Analysis of fission gas bubbles (FGBs) in light-water reactor (LWR) fuel is needed to improve the understanding and predictive capability of fuel evolution under normal- and transient conditions. One of the most important parameters of a FGB is the pressure of the gas, primarily Xe, inside it. Bubble volume and location (inter- vs. intragranular) are important considerations as well. However, experimental analysis of such bubbles is challenging due to their small size and embedded nature, and usefulness of the data requires large numbers be analyzed. This paper proposes a method to measure the pressure of Xe bubbles using X-ray spectrum imaging (XSI) in scanning transmission electron microscopy (STEM). From X-ray generation and instrumental parameters, the number of Xe atoms yielding a given number of Xe L-series X-ray counts is estimated, and then from the estimated volume of the bubble the Xe density is estimated, which can then be converted to pressure via an equation-of-state. We apply the method to XSI datasets from high burnup (HBu) fuel from North Anna 1 reactor in as-irradiated and post-LOCA-test conditions and find Xe pressures in the FGBs clustered around 1 GPa.

Ultrasonic atomization builds a hierarchical microstructure in densified Yb-filled skutterudites, leading to multi-scale phonon scattering and improved average thermoelectric figure of merit zT .

Focused ion beam (FIB) milling has become the dominant approach for site-specific transmission electron microscopy (TEM) specimen preparation; however, FIB damage remains a critical limitation for reliable microstructural characterisation, particularly in radiation effects studies. Tungsten is especially susceptible to FIB damage due to its high nuclear stopping power, which promotes the formation and strong diffraction contrast of FIB-induced 'black spot' defects that are indistinguishable from very fine irradiation-induced loops/defects resulting from low to intermediate temperature neutron irradiation. In this work, flash electropolishing is systematically evaluated as a post-FIB treatment for minimising preparation-induced artefacts for TEM analysis of tungsten-based alloys. Using a range of non-, ion-, and neutron-irradiated tungsten materials, the effectiveness of flash electropolishing has been assessed through direct comparison with conventional FIB and plasma-FIB preparation including low-energy Ga, Ar, Xe ion cleaning. The results demonstrate that flash electropolishing effectively removes FIB-damaged layers and 'black spot' defects, thereby enabling reliable observation of irradiation-induced dislocation structures. Key processing parameters governing flash electropolishing quality - including lamella thickness, applied voltage, polishing duration, electrolyte chemistry, and cathode geometry - have been systematically evaluated, and clear criteria were established for determining when flash electropolishing is required to ensure reliable microstructural analysis. This work also provides practical guidance for implementing flash electropolishing as an artefact-controlled specimen-preparation approach for TEM characterisation of FIB-produced specimens. The systematic protocol can be extended to other, non-tungsten materials. LAY DESCRIPTION: Flash electropolishing is demonstrated as an effective post-FIB treatment for removing preparation-induced damage in tungsten TEM specimens, including FIB-induced 'black spot' defects that can obscure irradiation microstructures. Through systematic comparison with conventional and plasma-FIB methods across irradiated and non-irradiated tungsten alloys, key processing parameters and reliability criteria are established. The work provides a practical, artefact-controlled preparation protocol for accurate radiation-damage characterisation and offers a broadly adaptable strategy for other material systems.

At high temperature conditions relevant to fossil and nuclear energy plants, Laves phase (Fe2X, X=Mo, W) precipitation is observed in common ferritic martensitic (FM) structural steels, with various reported effects on creep behavior. Despite being valuable metrics to correlate with mechanical properties and other precipitate phases, the volume fraction and number density of Laves phase precipitates has been difficult to quantify accurately using common techniques such as transmission electron microscopy (TEM) due to the relatively large size (similar to 0.25 mu m) and low number density (similar to 10(11) cm(-3)) of Laves precipitates. To address this characterization challenge, we developed and demonstrated a high-throughput and widely accessible method to quantify the volume fraction and number density of the Laves phase based on scanning electron microscope (SEM) images with a backscattered electron signal and the information depth (ID) of backscattered electrons. We applied this new technique in creep ruptured Grade 92 FM steel to study the effect of Laves phase on creep properties and determine the influence of stress on Laves phase precipitation. The quantitative accuracy of the SEM-based volume fraction and number density values was verified using synchrotron high energy X-ray diffraction and serial sectioning tomography. Stress did not significantly affect the Laves phase size or volume fraction during creep testing at 550 - 650 degrees C and stress levels of 90 - 260 MPa (vs. unstressed conditions). Conversely, a moderate but statistically significant stress-enhanced increase in Laves phase number density, corresponding to an increase in nucleation rate, occurred during creep exposure above 110 MPa.

A shear band is a heterogeneous, narrow seam within a solid material whose formation is caused by intense localized shearing when a sufficiently large amount of deformation occurs. If that deformation occurs at a sufficiently rapid rate, with operative friction, then co‐located ephemeral heating will occur in the shear band. In this study, shear bands were produced from dynamic shear‐induced compaction of a granular form of crystalline α‐quartz (SiO 2 ). The produced shear bands were approximately 25‐µm thick and were examined with scanning electron microscopy/electron backscatter diffraction, transmission electron microscopy, Raman spectroscopy, and nanoindentation. They were found to contain a mixture of vitreous silica and small‐sized crystallites. This finding is significant because the presence of the vitreous silica within the shear band is a postmortem indicator that the localized temperature had reached or exceeded the melting temperature (∼1723°C) of crystalline SiO 2 during the rapid shear and compaction and then sufficiently rapid cooling quenched in that vitreous state.

Uranium-based intermetallic Laves phases (UM2) provide valuable insight into the interplay of 5f orbital hybridization and magnetic properties. However, ferromagnetism in these compounds is rare. UFe2 is a cubic C15 Laves phase with greatest Curie temperature (TC) of all such compounds, at 162 K. In this study, we report the synthesis and characterization of a novel U-based Laves phase, UFe2+Ni, effectively forcing excess Ni in to the UFe2 structure. UFe2+Ni crystallizes in the cubic C15 Laves phase, appearing to deviate from the typical UM2 formula and exhibiting a significantly extended homogeneity range. Annealing this material leads to an increased ferromagnetic transition temperature, between 35 and 155 K, and decreased lattice parameter with higher annealing temperatures. This variation is attributed to changes in Fe and Ni solubility with annealing temperature. Our findings highlight the potential for tuning magnetic properties in U-based Laves phases through compositional and thermal modifications, expanding the understanding of magnetically ordered intermetallics.

Four monoblocks coming from one ITER-like plasma-facing unit from the Q3B sector of the lower divertor, named as monoblock (MB)3, MB9, MB20, and MB30, were exposed to the deuterium and helium plasma mixture during the C3 (2018) and C4 (2019) campaigns of the Tungsten Environment in Steady-state Tokamak (WEST), followed by a detailed ex-situ X-ray photoelectron spectroscopy investigation. The surface and in-depth chemistry of the tungsten monoblocks indicated the formation of a re-deposited mixture in the deposition-dominated area of the divertor, thicker than 218 and 172 nm for MB3 and MB9, respectively. The redeposition layer was dominated by a mixture of boron carbides accompanied by tungsten carbides in MB3, while in MB9, the redeposition layer was dominated by tungsten borides. The remaining two monoblocks, MB20 and MB30, were collected from the erosion region and showed similar chemical behavior with a blended mixture of oxidized and metallic tungsten followed by boron carbides within a 50 nm depth range. Boron fixation in the layers is an expected consequence of the boronizations used during the operation, but the chemical status of redeposited elements was characterized for the first time in this work.

Refractory metal composites are desirable for use in extreme environments that require materials with high specific strengths and resilience to external environments such as that found in nuclear and aerospace. However, due to the high melting temperature of refractories, liquid state joining processes such as welding remains difficult. Ultrasonic additive manufacturing (UAM) provides a potential route for processing refractory composites because it is a solid-state (i.e., no-melting) process and allows for intermittent machining operations to be performed between welds. To demonstrate refractory composite fabrication, this study utilized UAM to machine a cavity to locate and sequester a Mo foil in a Zircaloy-4 (Zry-4) baseplate and additively build over the top with Zry-4 foils, thereby embedding the Mo in Zry-4 matrix. Significant deformation of the Zry-4 microstructure was observed: this deformation caused adiabatic heating and subsequent dynamic recrystallization through the transformation from alpha ->beta and then back to alpha as the material cooled. Flexural testing of the Zr-Mo composite revealed no delamination or failure, but the strength was not as expected, falling lower than a cold-worked Zry-4 sample. Finite element analysis supported that some bonding must have existed between the Zry-4 and Mo. There was indeed an interdiffusion zone at the Zry-4 foil-Mo foil interface, observing a metastable body-centered cubic beta-Zr lathe. It was determined that sufficient strain energy was present to encourage the nucleation of the beta-Zr grain along alpha-Zr grains. Future work is warranted to investigate UAM for refractory composite fabrication.

Steady-state deuterium plasma exposures were performed on ultra-high temperature ceramics titanium diboride (TiB2) and zirconium diboride (ZrB2) using the PISCES-RF linear plasma device (LPD) as early screening for first wall, plasma-facing applications. Deuterium plasma exposures were performed using 40 eV ion energies at 240, 525, and 800 degrees C sample temperatures and 90 eV ion energies at 240 degrees C sample temperatures to analyze TiB2 and ZrB2 surface morphology and chemistry evolution behavior. Post-plasma exposure chemistry characterization of the near surface ( <50 nm) region of the samples all show transition metal enrichment, indicating boron preferential erosion. Transition metal to boron fractions vary with plasma exposure temperature under the 40 eV ion energy; metal enrichment is maximized at 800 degrees C and then minimized at 525 degrees C. SEM micrographs of all plasma exposed sample surfaces show no significant or noticeable plasma induced damage from cracking or blistering.

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. To accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300∘C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8×1021n/cm2 fast fluence (En>1.0MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

Deuterium and helium plasma exposures on bulk TiB2 and ZrB2 samples were performed using the PISCES-RF linear plasma device. 40 and 90 eV deuterium ion plasma exposures were performed at 240 and 800 °C sample temperatures, and 80 eV helium ion plasma exposures were performed at 800 °C sample temperatures. Following plasma exposures, it was discovered that two plasma conditions (90 eV deuterium and 80 eV helium at 800 °C) resulted in thick (>200 nm) tantalum-rich (>10 at%) surface features on the targets, presumably from tantalum sourced from a tantalum adapter mask or cap used as part of the target holder. This work aims to characterize these tantalum-rich features and examine the mechanisms of impurity deposition.Plasma-induced surface morphology of the tantalum-rich surface layers depends on plasma properties and target temperature and chemistry. Greater titanium sputtering compared to zirconium resulted in more distinct surface features in the TiB2 samples compared to the ZrB2 samples via increased, prompt deposition onto tantalum surface impurities. There is still uncertainty as to why thick tantalum deposition only occurred under some plasma exposure conditions but not others; it is likely due to tantalum sputtering by a combination of boron molecules from the targets and carbon-impurities in the tantalum mask or targets. Impurity driven surface features are a well-documented phenomena in samples exposed to plasma from linear plasma device facilities—this work confirms the occurrence of this and emphasizes the need for chemistry characterization of isolated post-mortem surface features in plasma-exposed samples.

This long-standing collaboration between the US DOE and the National Institutes for Quantum Science and Technology (QST), Japan, jointly pursues activities using the advanced capabilities for materials irradiation and post-irradiation examination at ORNL, especially using HFIR, hot cells, and LAMDA.The focus is on ferritic steels and other advanced materials, especially the development of the materials database for the design of facilities beyond ITER.The goals include achieving DEMO-relevant performance data on structural materials neutron irradiated in HFIR to high levels of displacement damage.The DOE-QST collaboration continues to steadily build the materials databases for the RAFM alloys that include F82H, ODS, and Eurofer97 steels and other materials such as nuclear grade SiC composites and copper alloys.This database is being used by QST to predict performance of components of a DEMO fusion reactor.

This report describes the ORNL aspects of an NSUF project in collaboration between EPRI, ORNL, and NCSU. Preliminary characterizations at ORNL LAMDA laboratory on piezoelectric sensors and bonds before and after NCSU PULSTAR reactor irradiation. Five pre-irradiation and seven post-irradiation samples have been examined by scanning electron microscopy and related methods so far. Samples consisted of (Pb,Bi)(Sc,Ti)O3 or LiNbO3 ceramics. Bonding layers consisted of epoxy, zirconia, alumina, aluminosilicate, or a complex soda-lime type slurry.

Journal Article STEM Analysis of High Burnup Structure in LWR Fuels Get access Chad M Parish, Chad M Parish Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA Corresponding author: parishcm@ornl.gov Search for other works by this author on: Oxford Academic Google Scholar Jesse Werden, Jesse Werden Nuclear Energy and Fuel Cycle Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA Search for other works by this author on: Oxford Academic Google Scholar Tyler Gerczak, Tyler Gerczak Nuclear Energy and Fuel Cycle Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA Search for other works by this author on: Oxford Academic Google Scholar Jason Harp, Jason Harp Nuclear Energy and Fuel Cycle Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA Search for other works by this author on: Oxford Academic Google Scholar Casey McKinney, Casey McKinney Nuclear Energy and Fuel Cycle Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA Search for other works by this author on: Oxford Academic Google Scholar Nathan Capps Nathan Capps Nuclear Energy and Fuel Cycle Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA Search for other works by this author on: Oxford Academic Google Scholar Microscopy and Microanalysis, Volume 29, Issue Supplement_1, 1 August 2023, Pages 1543–1545, https://doi.org/10.1093/micmic/ozad067.794 Published: 22 July 2023

Attenuation-based neutron computed tomography (CT) has been used to non-destructively characterize the uncoated tristructural-isotropic (TRISO) nuclear fuel kernels in this work. Particularly, the effect of two different types of carbon blacks (Raven 3500 and Mogul L) on the internal gelation process of UO3-C kernels has been investigated. With 3D reconstructed kernel volumes and digital imaging processing techniques, heterogenous density distributions are mapped in both types of kernels. It is found that the kernels produced with Mogul L are - 20 % denser and - 10 % larger (in equivalent diameter) than the Raven 3500 kernels. Furthermore, less neutron attenuating regions, which are most likely to be carbon agglomerates as scanning electron microscopy (SEM) and energy-dispersive X-ray spectroscopy (EDS) results show, are observed in the Mogul L kernels. The size distribution of such carbon agglomerates (ranges from 50 & mu;m to 850 & mu;m with a peak at - 200 & mu;m) has been determined by analyzing the CT data. Furthermore, multiple metrics, including equivalent diameter, surface area, volume, sphericity, have been extracted to evaluate the fuel kernels. This work demonstrates that neutron imaging is an excellent, nondestructive tool to efficiently characterize, understand, and explore fuel materials for nuclear material research and development.

Silicon carbide (SiC) is often used as a passive temperature indicator for uninstrumented in-core nuclear experiments.For the past several decades, Oak Ridge National Laboratory has relied primarily on SiC thermometry manufactured by Dow Chemical Company, formerly Rohm and Haas because of the material's high density and reduced grain boundary elements.Although Dow SiC has performed well in this application, this material is no longer commercially available, and a new supplier is needed for future experiments.To determine a suitable replacement, a study was initiated on five types of SiC from four commercial vendors.Several critical properties, including density, electrical resistivity, chemical purity, grain structure, crystal structure, and strength were measured in samples from each material.Additionally, thermometry specimens were manufactured for irradiation in the High Flux Isotope Reactor at nominal temperatures of 300°C, 600°C, and 900°C, which will be measured using continuous dilatometry at a later date.This report provides an update of commercially available SiC material characterization and the irradiation capsule design.

Studies on high burnup UO2 subjected to loss-of-coolant accident conditions have shown that restructured regions of the fuel are susceptible to pulverization and eventual dispersal. Due to a lack of pre-test characterization, the distinct microstructural features rendering the fuel prone to fragmentation remain ambiguous. Four samples of commercially irradiated light-water reactor UO2 have been characterized utilizing electron backscatter diffraction to assess the susceptible microstructure. The microscopy focused on determining the burnup and temperature conditions responsible for the formation of the different microstructural regions where the regions were denoted as the high-burnup structure (HBS), HBS transition, mid-radial, restructured central, and central region. Previous works have outlined the specific conditions required for the restructuring of the microstructure into the HBS, but the conditions responsible for the restructuring in the central region of the fuel are not well understood. The four analyzed samples confirm a burnup threshold of 61 GWd/tU, and an unknown temperature range is needed to facilitate the formation of the restructured central region. Additional fuel performance evaluations are needed to quantify the temperature range promoting restructuring in the central region.