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DOE OSTI2026

A dual dynamic shutter system for accelerating ion irradiation sample throughput via lateral gas implantation gradients

Ion irradiation for material performance testing is limited due to its serial nature, which allows for only one value of the implantation (appm) versus dose (dpa) parameter space to be explored for each ion and experiment at a time. While ion irradiation can accelerate the process by up to three orders of magnitude compared to neutron irradiation experiments, the sample throughput for ion irradiation remains relatively low. To address these limitations, a novel capability has been developed at the Michigan Ion Beam Laboratory (MIBL), enabling for the creation of single- and two-dimensional lateral ion implantation gradients using recently installed motorized-controlled ion-beam shutters. This advancement can generate a wide scope of the two-dimensional (H+, He2+) implantation parameter space within a single sample. Integration of this new capability now allows for dual- and triple-ion beam experiments to be performed with full user control over not only the ion implantation depth, but also laterally across the sample by imposing ion implantation concentration gradients, thus providing researchers with a high-throughput means for material testing under various irradiation conditions. Furthermore, recent improvements in MIBL's microbeam ion-beam analysis (IBA) target station now allow for probing these concentration gradients in irradiated alloys with exceptional spatial resolution, down to 10 µm. These two approaches promise to significantly improve ion irradiation capabilities and increase the sample throughput by several orders of magnitude. The application of the shutter technique plus the subsequent microbeam characterization of the imposed implantation gradients are showcased by two proof-of-principle ion-irradiated experiments, one performed on single-crystal Si and the other on the fusion-candidate alloy F82H-IEA. These advancements mark a substantial leap in ion-beam technology, offering researchers a robust, high-throughput method to efficiently investigate candidate alloys with high technological readiness for both advanced fission and fusion reactor applications, in a time- and cost-effective manner.

36 - MATERIALS SCIENCE↗

DOE OSTI2021

Organic Hole Transport Material Ionization Potential Dictates Diffusion Kinetics of Iodine Species in Halide Perovskite Devices

Iodine-containing volatiles are major degradation products of halide perovskite materials under irradiation, yet iodine diffusion kinetics into and throughout organic hole transport materials (HTMs) and consequent reactions are largely unexplored. In this work, we modify the Ca:O 2 corrosion test to Ag:I 2 to quantify I 2 transmission rates through common organic HTMs. We observe I 2 permeability to inversely correlate with HTM ionization energy, or the highest occupied molecular orbital (HOMO) energy. Tracking electronic conductance during exposure to I 2 confirms shallow HOMO HTMs are strongly oxidized (i.e., doped), leading to substantial I 2 uptake and increased transmission rates. Finally, relationships between HOMO level, doping, and transmission rate are maintained when methylammonium lead triiodide (MAPbI 3 ) photolysis products are the only source of iodine. While HTM energetics influence the initial performance of halide perovskite devices by selective charge extraction, our results further suggest they will affect device stability; deeper HOMO energy HTMs will suppress iodine migration and associated degradation mechanisms.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Post-Irradiation Evaluation of Eurofer97 Fracture Toughness Using Miniature Multinotch Bend Bar Specimens

In this study, we performed fracture toughness characterization of ten neutron-irradiated Eurofer97 variants using precracked miniature multi-notch bend bar (M4CVN) specimens based on the Master Curve method in the ASTM E1921 standard. The neutron irradiation was performed in the flux trap position of the High Flux Isotope Reactor (HFIR) of the Oak Ridge National Laboratory (ORNL) with the nominal irradiation temperature of 300°C and irradiation dose of 2.5 displacements per atom (dpa). Depending on the irradiation temperature and materials, we observed different degrees of irradiation hardening and embrittlement for ten Eurofer97 variants. The upper shift in the Master Curve reference temperature T0Q vs. the increase in Vickers microhardness values showed a liner relationship for only a few materials indicating different irradiation responses of the Eurofer97 variants.

Chen, Xiang↗

DOE OSTI2024

Thermal Gradient and Neutron Irradiation Experiment Design for Fusion Reactor Materials in the Advanced Test Reactor

This work outlines a hypothetical coupled thermal gradient and neutron irradiation experiment in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. Although the ATR is a thermal spectrum test reactor and doesn’t inherently produce a flux spectrum dominated by the high-energy neutrons typical in a fusion reactor, it’s multitude of experiment positions and dynamic flux environment make it a suitable platform for investigating fusion related issues.

36 MATERIALS SCIENCE↗

DOE OSTI2024

AGC-4 Specimen Post-Irradiation Examination Data Interim Report

This interim report documents preliminary results of the post-irradiation examination material property testing from the fourth advanced graphite creep (AGC), AGC 4, capsule specimens. This is the fourth of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades to moderate dose levels (=7 dpa). The AGC 4 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 800°C and to a peak dose of 8 dpa. Half of the AGC-4 specimens were subjected to compressive stresses to induce irradiation creep. Post-irradiation testing and measurement results are reported with the exception of thermal testing, which is still in progress, and irradiation mechanical strength testing. Additionally, some specimens initially deemed too hot to be examined in the ART Graphite laboratory may still be measured. The data reported includes specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and dimensional data necessary to derive density, elastic constants (Young?s modulus, shear modulus, and Poisson?s ratio) from ultrasonic time of flight velocity measurements, Young?s modulus from the fundamental frequency of vibration, and electrical resistivity. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to the irradiation environment and applied load on the specimens, will be discussed later in AGC 4 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

AGC-4 Specimen Post-Irradiation Examination Data Interim Report

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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 OSTI2022

Radiation effects on the structure and alteration behavior of an SiO 2 –Al 2 O 3 –B 2 O 3 –Na 2 O glass

As borosilicate glasses are used in many countries to immobilize fission products and minor actinides after spent fuel reprocessing before storage in a deep geological repository, assessing that their chemical durability is of paramount importance. Here, pristine and preirradiated (952 MeV, 136Xe) SiO 2 –B2O 3 –Al2O 3 –Na 2 O glasses with the same molar ratios as in the French SON68 and ISG glasses have been subjected to aqueous corrosion in deionized water and in silica-saturated solution to measure the initial and longer term alteration rates. Pristine and preirradiated glasses corrode following the same mechanisms, but the preirradiation has a strong impact on the initial dissolution rate (increase by a factor of 5.6), and on the alteration layer depth in silica-saturated conditions (by two- to threefolds). The later result is related to the formation of a more porous, less passivating gel on the preirradiated glass specimen. Using both experimental spectroscopies (NMR, IR, and SFG) and classical molecular dynamics, the radiation effects on the glass structure and water diffusion have been assessed. After preirradiation, the density and the polymerization degree of the glass decrease, whereas the topological disorder increases. In consequence, water diffusion accelerates. These observations allow to correlate the radiation impact on the alteration behavior to the structural changes.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Modeling and Design of Micro Flux Monitors for Neutron Fluence Measurements

Measuring neutron flux is critical to understanding the characteristics of nuclear reactors and can provide information on the energy spectrum, total neutron fluence, and operational history of a facility. Such measurements are typically performed by irradiating high-purity materials in an operating reactor for a known length of time. By measuring and identifying activation products in the irradiated materials and referencing the energy-dependent cross section of each isotope, the neutron energy spectrum and total fluence can be determined. Historical techniques use a variety of small, individually pressed foils or wires for dosimeter devices, which can result in excessively high levels of radioactivity and complications in handling after irradiation. The present work describes a new neutron dosimetry concept, named the micro flux monitor, that uses semiconductor device fabrication processes to deposit thin films of flux monitoring materials in precise geometries onto a SiC wafer. This approach allows multiple dosimetry materials to be located on a single device, simplifying irradiation and subsequent analysis. The geometry, and therefore mass, of each material can be custom-tailored for a specific application—considering the expected neutron flux, duration of irradiation, and device retrieval time of a facility. Additionally, hundreds of identical dosimetry monitors can be produced simultaneously, reducing fabrication costs. Finally, the extremely thin metal dimensions of these devices could improve the fidelity of neutron fluence measurements by reducing neutron self-shielding and gamma self-attenuation effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

Patch-Based Convolutional Neural Networks for Multiple Microstructural Features Detection in FIB-SEM Micrographs of Irradiated Nuclear Fuel

Focused ion beam scanning electron microscopy (FIB-SEM) tomography has increasingly been utilized for acquiring three-dimensional (3D) microstructure features at the sub-micron scale in irradiated nuclear materials. This technique involves sequential ion beam slicing followed by electron beam imaging and compositional mapping using energy dispersive spectroscopy (EDS). Despite its growing use, several challenges persist. These include the time-intensive nature of data collection of EDS data, difficulties in distinguishing between various microstructures, and issues with image alignment. These challenges currently limit the broader application of FIB-SEM tomography in the field. To overcome these limitations, we propose using convolutional neural networks (CNNs) to automate microstructure identification in SEM images. Our study introduces a new framework for identifying microstructures in irradiated U-10Zr (wt. %) metallic fuel with limited annotated data. The framework includes the creation of a reliable annotated dataset with paired SEM and ground truth data from EDS maps, the applications of CNNs for microstructure identification, and the validation of model performance. Specifically, we employed the Segment Anything Model (SAM) to align SEM images with corresponding EDS maps and focused ion beam (FIB) tomography SEM data. We evaluate several models, including Patch-based U-Net, Attention U-Net, and Residual U-Net, finding that patch-based U-Net exhibits superior segmentation performance and consistency. This approach reduces reliance on EDS detectors and aids in accelerating nuclear material analysis process, highlighting the potential of advanced deep learning techniques to improve microstructural understanding in nuclear material. This is the first framework to integrate SAM and Patch-based CNN models for semantic segmentation of irradiated nuclear materials, with potential applicability to other tomography datasets.

36 - MATERIALS SCIENCE↗

DOE OSTI2023

Irradiation-enhanced Interactions at UO 2 /Al 2 O 3 /Al Interfaces

Combustion synthesis is used to prepare thin UO 2 films on aluminum alloy substrates. This simple preparation method involves electrospraying uranyl nitrate + acetylacetone + 2-methoxyethanol solution on the substrate, followed by a short annealing at 350 or 550 °C. The irradiation of films with a 40 Ar 2+ ion beam (energy of 1.7 MeV and fluences of 7.7 × 10 16 and 1.3 × 10 17 ions/cm 2 ) is conducted to investigate irradiation-induced restructuring processes. High-resolution transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS) investigations show that the annealing temperature significantly influences the crystallinity and stability of materials during irradiation. A small amount of Mg in the alloy substrate diffuses into the amorphous Al 2 O 3 interfacial layer between the film and the substrate. Local thermal spikes from the incoming ions facilitate the irradiation-induced mixing of immiscible Al 2 O 3 and UO 2 for the materials prepared at 350 °C. This mass diffusion produces relatively large cavities at the interface. Selective diffusion of a more significant amount of Mg for the materials prepared at 550 °C suppresses the mixing of the Al 2 O 3 interlayer with the film but forms Mg y U 1–y O 2±x solid solutions during irradiation. Local thermal heating triggers the precipitation of a discontinuous crystalline MgO layer close to the film surface. As a result, the enhanced and selective diffusion of Mg into the film makes the materials prepared at 550 °C more robust and mechanically stable than those prepared at 350 °C.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Temperature threshold for preferential bubble formation on grain boundaries in tungsten under in-situ helium irradiation

Understanding a material's radiation tolerance requires examining its performance under different irradiation conditions. In this work, we investigate the radiation tolerance in terms of helium bubble damage in tungsten irradiated in-situ with 16 keV helium at 1073 K and 1223 K. Damage evolution represented by helium bubble density, size and total change in volume in the grain matrices and the grain boundaries are quantified as a function of fluence. Preferential large bubble formation and a higher change in volume on the grain boundaries occurred at 1223 K, suggesting faster migration of certain helium-vacancy complexes as confirmed by a diffusion-reaction model.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Changes to the morphology and coefficient of thermal expansion in HDPE and UHMWPE following irradiation-based crosslinking

This work describes recent findings of the Los Alamos National Laboratory concerning the alterations to polyethylene’s (PE’s) mechanical and chemical properties after crosslinking via gamma irradiation. Three materials were examined as part of this study, including one HDPE and two UHMWPE specimens. Changes to degree of crystallinity (X C ) were qualitatively determined via measurements of density, and results suggest that X C is not necessarily correlated to crosslink density, depending on polymer manufacturing process and treatment following irradiation. TMA data suggests an anisotropic response in linear thermal expansion, and that the response may be more correlated to X C than crosslink density. Finally, FTIR analysis has led to the conclusion that oxidation occurs as a result of irradiation in open air, the degree of oxidation is proportional to the maximum radiation dose received, and that oxidation is largely a surface effect, penetrating no deeper than 1.5 mm. Due to time constraints, a full dataset could not be generated for changes in creep and thermal properties as a result of irradiation, and TMA analysis is only complete for one material tested in this study. These measurements are ongoing and will provide more context to the hypotheses drawn in this work.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Multi-energy ion irradiation effects and distinctive features of color center generation in yttria-stabilized zirconia driven by structural discrepancy

Regarding a requirement for inert matrix fuel (IMF) to burn minor actinides and plutonium and high-level waste immobilization, Yttria-stabilized zirconia (ZrO 2 -6.5 wt% Y 2 O 3 ) (YSZ) crystals with cubic phase are selected to comprehensively evaluate the irradiation resistance by irradiation-induced structural discrepancies across multiple energy conditions. Electronic excitation and nuclear collisions reveal distinct color centers, influenced by varying component ratios that reflect their formation mechanisms. Key factors driving ultra-fast structural transitions include collision cascades, pressure waves, and energy dissipation from electronic excitation. In this work, we find that in the electronic energy loss (E ele )-dominant region, internal latent tracks with unique surface nanostructures emerge, distinct from the defect clusters seen in the nuclear energy loss (E nuc )-dominant region, while the threshold for latent track formation and melting are identified. In conclusion, structural discrepancies driven by different energy-loss mechanisms promote the generation of singly ionized and nearest-neighbor doubly ionized oxygen vacancies, which respectively dominate the formation of F⁺ and T-centers, resulting in bandgap narrowing (4.80 eV → 4.73 eV), and furthermore triggering visible light emission (2.17 eV → 2.20 eV) in irradiated YSZ crystals, thus, enabling the design of novel irradiation-tailored material functionalities.

36 MATERIALS SCIENCE↗

DOE OSTI2022

A fusion relevant data-driven engineering void swelling model for 9Cr tempered martensitic steels

The UCSB database on cavity evolution in 9-12Cr tempered martensitic steels (TMS), includes the results for both dual heavy and helium ion (DII), and High Flux Isotope Reactor (HFIR) in situ helium injection (ISHI) neutron irradiations at 500°C. These results were combined with literature single ion and fission neutron irradiation data to derive a model for the void volume fraction, f v , as a function of displacements per atom (dpa) and transmutant helium concentrations in atomic parts per million (appm). The scientific foundation for the paper is described in a companion paper entitled “Cavity Evolution and Void Swelling in Dual Ion Irradiated Tempered Martensitic Steels”. Here, in this study, we show that f v (dpa, He/dpa) is described by the incubation dose, dpa i , for the onset of void growth, and the post-incubation growth rate, f v ’(%/dpa). Both dpa i and f v ’ decrease with increasing He/dpa at > ~ 5. The dpa i is also lower for the ISHI neutron irradiations at the same He/dpa. Single heavy ion and fission reactor neutron irradiations, with low He/dpa ratios, have a much larger dpa i . Based on a combined analysis of DII, single ion, ISHI and fission neutron data, we further show that the post-incubation f v data analyzed here have a common empirical curve shape, with f v ’ reaching up to ~ 0.2%/dpa at very high dpa. We also show that f v ’ can be predicted based on a physical model of defect partitioning between evolving sinks. At 500°C and fusion relevant He/dpa ≈ 10, the best-fit model predicts nominal swelling, S = f v /(1-f v ), of ~ 1.1, 4.9 and 16% at 50, 100 and 200 dpa, respectively. The physically motivated, data-driven model includes estimated uncertainties for both dpa i and f v ’.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Thermal diffusivity of irradiated tungsten and tungsten-rhenium alloys

The Japan-US PHENIX project irradiated tungsten materials in the RB-19J capsule experiment in the High Flux Isotope Reactor (HFIR). A gadolinium (Gd) shielding was used to absorb the thermal neutrons and reduce rhenium and osmium generation in tungsten. Pure tungsten and K-doped W-3% Re samples were irradiated at 532 – 662 °C to dose of 0.21-0.46 dpa, with the grain orientation perpendicular or parallel to the disk surface. Thermal diffusivity measurements were performed from 100 °C to 500 °C. Furthermore, additional measurements followed after annealing up to 900 °C. Irradiated pure tungsten specimens showed similar thermal diffusivity results compared with an unirradiated W-1% Re specimen in another study. The transmutation amount of Re was calculated to be about 0.52% for those specimens that showed good agreement with this study. Specimens irradiated in this study to different doses presented almost the same thermal diffusivity. Annealing up to 800 °C resulted in no recovery of thermal diffusivity. These results show that the contribution of crystalline defects to degradation of thermal diffusivity is quite limited. In addition, the thermal diffusivity of the irradiated specimens was getting close to that of the unirradiated specimens at elevated temperature.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Surrogate isotope-containing materials for emergency response training and Methods of formation and dispersal

Surrogate materials are in the form of solid particles that include surrogate isotopes, namely, short-lived isotopes selected and formed to serve as surrogates for the radioactive materials of a nuclear fallout without including isotopes that are, or that decay to, biologically or environmentally deleterious and persistent isotopes. The surrogate material may be formed using high-purity reactant material and irradiation and separation techniques that enable tailoring of the isotopes and ratios thereof included in the surrogate material, and the surrogate material may be dispersed, e.g., in a training environment, in solid form.

Carney, Kevin P.↗

DOE OSTI2021

Impact of neutron irradiation on the thermophysical properties of additively manufactured stainless steel and inconel

The deployment of additive manufacturing (AM) technologies to support the nuclear energy industry is limited by the current lack of data related to the irradiation performance of AM parts. The Colorado School of Mines is conducting a multi-year irradiation and post irradiation examination project to study the impacts of neutron irradiation on stainless steel 316L and Inconel 718 and 625 specimens produced by several commercially available AM techniques and irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). This paper presents initial results on the impact of neutron irradiation on the heat capacity, the thermal diffusivity, and the thermal conductivity of the specimens at temperatures between 100 oC and 400 oC. The unirradiated AM specimens have heat capacity values similar to, or lower than, the reference values, regardless of manufacturing method or material. The irradiated AM specimens have average heat capacities greater than the unirradiated specimens and reference values in every case. Additionally, the thermal diffusivities of the unirradiated AM specimens are either larger or statistically similar to the reference values. The irradiated AM specimens exhibit lower thermal diffusivities compared to the reference thermal diffusivity values. One of the stainless steel 316L laser powder bed fusion samples exhibited thermal diffusivity values with statistically significant anisotropy prior to and after irradiation. The thermal conductivities of the unirradiated AM stainless steel 316L specimens and the unirradiated Inconel 718 specimens are lower or statistically similar to the reference values. The unirradiated Inconel 625 specimens exhibit slightly higher thermal conductivity values than the reference values. The irradiated specimens show a significant increase in thermal conductivity compared to the unirradiated specimens and the reference values for all materials.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗