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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Irradiation Driven Restructuring of Nanocrystalline ThO 2 and Th 1–x U x O 2 Thin Films

Irradiation induced structural changes of actinide oxide materials is a key consideration in their development and use as nuclear fuels. This study reported on the synthesis of ThO 2 and Th 1–x U x O 2 (x = 0.15, 0.50) thin films, fabricated using electrospray-assisted solution combustion synthesis, and their responses to ion irradiation. Krypton ion irradiations, up to a fluence of 1 × 10 16 ions/cm 2 , were carried out to simulate radiation damage induced by fission products in a reactor environment. Structural and chemical changes induced by irradiation were analyzed using high-resolution scanning transmission electron microscopy (STEM), energy-dispersive X-ray spectroscopy (EDS), and electron energy-loss spectroscopy (EELS). It was determined that the extent and nature of irradiation-induced damage are strongly correlated with the uranium content. ThO 2 films were most susceptible to radiation-induced damage, with significant cavity formation and delamination from the substrate at high fluence. Of the compositions studied, Th 0.85 U 0.15 O 2 films showed the highest stability, characterized by moderate grain growth and the absence of voids or severe defect structures. In contrast, Th 0.5 U 0.5 O 2 films accumulated extensive damage, including the formation of a nanocrystalline central region. EELS analysis indicated that oxygen displacement is the primary driver of structural degradation in Th 0.5 U 0.5 O 2 films. α-particle spectroscopy confirmed minimal actinide loss across all compositions, underscoring the mechanical robustness of the films. These findings provide insight into the irradiation-induced damage mechanisms in Th O2 and Th 1–x U x O 2 systems, supporting their development as potential materials for nuclear fuels and irradiation-tolerant thin film targets in nuclear physics measurements.

Th1−xUxO2↗
DOE OSTI2023

Comparison of the Thermal Stability in Equal‐Channel‐Angular‐Pressed and High‐Pressure‐Torsion‐Processed Fe–21Cr–5Al Alloy

Nanostructured steels are expected to have enhanced irradiation tolerance and improved strength. However, they suffer from poor microstructural stability at elevated temperatures. In this study, Fe–21Cr–5Al–0.026C (wt%) Kanthal D (KD) alloy belonging to a class of (FeCrAl) alloys considered for accident‐tolerant fuel cladding in light‐water reactors is nanostructured using two severe plastic deformation techniques of equal‐channel angular pressing (ECAP) and high‐pressure torsion (HPT), and their thermal stability between 500–700 °C is studied and compared. ECAP KD is found to be thermally stable up to 500 °C, whereas HPT KD is unstable at 500 °C. Microstructural characterization reveals that ECAP KD undergoes recovery at 550 °C and recrystallization above 600 °C, while HPT KD shows continuous grain growth after annealing above 500 °C. Enhanced thermal stability of ECAP KD is from significant fraction (>50%) of low‐angle grain boundaries (GBs) (misorientation angle 2–15°) stabilizing the microstructure due to their low mobility. Small grain sizes, a high fraction (>80%) of high‐angle GBs (misorientation angle >15°) and accordingly a large amount of stored GB energy, serve as the driving force for HPT KD to undergo grain growth instead of recrystallization driven by excess stored strain energy.

Arivu, Maalavan↗
DOE OSTI2021

Enhancement of Nanostructured Ferritic Alloy 14YWT Properties via Heat Treatment for Post-Consolidation Processing

The nanostructured ferritic alloy 14YWT is a promising candidate for in-core use in generation IV nuclear reactors, due to a dense dispersion of insoluble, ultrafine-scale Y-Ti-O nano-oxides, which provide a high degree of irradiation tolerance and thermal stability. This work investigates the effects of heat treatment on the workability of 14YWT, along with the effect of processing history on abnormal grain structures and radial microstructural uniformity. In this study, a 14YWT rod consolidated at 850 °C was heat treated in argon at 1100 °C, 1150 °C, 1200 °C, or 1250 °C for 1 or 8 hours and changes in mechanical properties and microstructure were examined using microhardness and electron backscattered diffraction (EBSD). Two distinct types of large abnormal grains were observed, each with unique processing origins, including one with high and the other with low strain energy. Consolidation via direct extrusion resulted in radial microstructural gradients, where the center of the rod was softer with larger grain sizes and lower strain energies. These gradients persisted and intensified throughout heat treatment. Based upon this work, the recommended heat treatment for increased workability with minimal microstructural change is 1150 °C for 1 hours.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The effect of Zr on precipitation in oxide dispersion strengthened FeCrAl alloys

Oxide dispersion strengthened (ODS) FeCrAl alloys are promising candidate materials for advanced nuclear reactor applications requiring high-temperature strength, corrosion resistance, and irradiation tolerance. As these alloys have increased in compositional complexity through attempts to use highly reactive elements such as Zr to refine particle sizes and optimize nanoprecipitate dispersion characteristics, much debate has ensued as to the effects of these alloying element additions on alloy properties. In an attempt to reconcile differences in nanoprecipitate distributions reported in the literature over the past decade, a detailed investigation of a recently developed ODS FeCrAl alloy with nominal composition Fe–10Cr-6.1Al-0.3Zr+0.3Y 2 O 3 is presented here using a combination of atom probe tomography (APT), scanning/transmission electron microscopy (S/TEM), and computational thermodynamics modeling. It is illustrated that based on the amount of Zr available in the lattice, Zr competes with Al and Cr to form carbides and nitrides as opposed to oxygen-rich precipitates. This alloy system has a high number density (>10 23 m -3 ) of ~2–4 nm diameter (Y,Al,O)-rich nanoprecipitates, but it is shown that due to the compositional spread and unknown partitioning of Al between the matrix and precipitates, significant challenges still exist for quantifying the exact compositions of these precipitates using APT. However, the noted compositional spread is supported by identified complex oxides yttrium aluminum monoclinic (YAM) and yttrium aluminum garnet (YAG) using S/TEM. As a result of these findings, researchers developing ODS FeCrAl with reactive element additions must pay careful attention to C and N impurities when optimizing reactive element additions.

electron microscopy↗
DOE OSTI2026

Radiation damage in tungsten under high-energy He-ion irradiation

Tungsten (W) is a primary candidate for plasma-facing components (PFCs) in fusion reactors and hybrid fission-fusion reactors. However, W suffers from severe microstructural damage under Helium (He) ion irradiation in operation conditions. Key irradiation parameters, including incident ion energy, fluence, and irradiation temperature, are known to determine the formation of He bubbles and dislocation loops, which ultimately lead to the degradation of mechanical properties, such as radiation hardening and ductile-to-brittle transition temperature (DBTT) shifts. Various strategies, for instance, interface engineering and alloying, have been developed to enhance He irradiation tolerance in W to address these challenges. Studies in the recent two decades have enhanced the understanding of the underlying mechanisms that drive defect evolution and mechanical performance degradation, thereby offering more insights into the design of more resilient W-based materials for extreme environments.

Liu, Yinghang [Purdue University]↗
DOE OSTI2022

Fabrication, thermal analysis, and heavy ion irradiation resistance of epoxy matrix nanocomposites loaded with silane-functionalized ceria nanoparticles

This paper describes a detailed understanding of how nanofillers function as radiation barriers within the polymer matrix, and how their effectiveness is impacted by factors such as composition, size, loading, surface chemistry, and dispersion. Here, we designed a comprehensive investigation of heavy ion irradiation resistance in epoxy matrix composites loaded with surface-modified ceria nanofillers, utilizing tandem computational and experimental methods to elucidate radiolytic damage processes and relate them to chemical and structural changes observed through thermal analysis, vibrational spectroscopy, and electron microscopy. A detailed mechanistic examination supported by FTIR spectroscopy data identified the bisphenol A moiety as a primary target for degradation reactions. Results of computational modeling by the Stopping Range of Ions in Matter (SRIM) Monte Carlo simulation were in good agreement with damage analysis from surface and cross-sectional SEM imaging. All metrics indicated that ceria nanofillers reduce the damage area in polymer nanocomposites, and that nanofiller loading and homogeneity of dispersion are key to effective damage prevention. The results of this study represent a significant pathway for engineered irradiation tolerance in a diverse array of polymer nanocomposite materials. Numerous areas of materials science can benefit from utilizing this facile and effective method to extend the reliability of polymer materials.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Displacement of carbon atoms in few-layer graphene

Molecular dynamics simulations were performed to study the susceptibility of carbon atom displacement under electron irradiation. The mapping of threshold displacement energies at different recoiling directions showed that the energies are very sensitive to the layer configurations and positions of neighboring atoms. Carbon atoms on the top and the bottom layers of few-layer graphene are most vulnerable to irradiation damage due to lack of constraints from the neighboring graphene layers. As indirect experiment evidence, transmission electron microscopy was performed on the edge of folded few-layer graphene, which made it possible to reveal “the inside” and compare irradiation tolerance of atoms at different layers, by using an electron analysis beam for both displacement creation and in situ characterization.

74 ATOMIC AND MOLECULAR PHYSICS↗
NASA NTRS1980

Characterization of reflected light from the space power system

Sunlight reflected off large space structures is examined. To assure that this illumination does not exceed the irradiance tolerances of the eye, reflections from these satellites must be controlled by vehicle orientation and surface specifications. The components of various space power system vehicles to determine the reflectances which will significantly contribute to the ground illumination are evaluated. Calculations of reflected solar intensity from various satellite system elements requires description of the elements and of the geometry potential reflectance paths. Surface intensity and the conditions under which it will illuminate a portion of the Earth are also determined.

Liemohn, H. B.↗
DOE OSTI2022

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗
NASA NTRS1987

Neutron radiation tolerance of Au-activated silicon

Double injection devices prepared by the introduction of deep traps, using the Au activation method have been found to tolerate gamma irradiation into the Gigarad (Si) region without significant degradation of operating characteristics. Silicon double injection devices, using deep levels creacted by Au diffusion, can tolerate fast neutron irradiation up to 10 to the 15th n/sq cm. Significant parameter degradation occurs at 10 to the 16th n/sq cm. However, since the actual doping of the basic material begins to change as a result of the transmutation of silicon into phosphorus for neutron fluences greater than 10 to the 17th/sq cm, the radiation tolerance of these devices is approaching the limit possible for any device based on initially doped silicon.

Joyner, W. T.↗
DOE OSTI2024

NON-DESTRUCTIVE POST IRRADIATION EXAMINATION OF FIRST CYCLE ACCIDENT TOLERANT AND ADVANCED ZIRCONIUM ALLOY HIGH BURNUP FUEL RODS

Post irradiation examination (PIE) of accident tolerant fuel concepts and high burnup uranium dioxide advanced zirconium alloy clad fuel contribute to near term nuclear industry goals of enhancing light water reactor safety and economics. Westinghouse is developing chromium coated zirconium alloy cladding as an accident tolerant fuel option for current pressurized water reactors. Examination of high burnup fuel further augments the technical foundation for extending the peak rod average burnup limits beyond the current regulatory limit of 62 MWd/kgU. Seven fuel rods were received at the Oak Ridge National Laboratory Irradiated Fuel Examination Facility. This includes three chromium coated zirconium alloy fuel rods and four high burnup advanced zirconium alloy fuel rods. Six of the seven rods were fueled with standard uranium dioxide while one of the chromium coated rods contained Westinghouse’s ADOPT fuel. The overall hot-cell PIE plan and some PIE results have been presented previously. This paper focuses on the additional non-destructive examination and fission gas release. Highlights from visual examination of these rods are presented. The chromium coating showed little to no visual change following irradiation. Observations of the axial burnup trends in the fuel rods are derived from full rod axial gamma-ray scans with additional observations related to local fission product migration. The axial dimensional changes of the seven rods are evaluated. As expected, creep down had not yet occurred in the chromium coated rods, and the maximum diameter strain in the high burnup advanced zirconium alloy cladding was less than 0.5%. Finally, the steady state fission gas release for these rods was measured ranging from 4.1% to 17.6% which matched expectation from literature. These fuel rods now await further destructive examination.

Harp, Jason↗
DOE OSTI2023

Solution Processed Schottky Diodes Enabled by Silicon Carbide Nanowires for Harsh Environment Applications

Silicon carbide nanowires (SiC NWs) exhibit promising features to allow solution-processable electronics to be deployed in harsh environments. By utilizing a nanoscale form of SiC, we were able to disperse the material into liquid solvents, while maintaining the resilience of bulk SiC. Here, this letter reports the fabrication of SiC NW Schottky diodes. Each diode consisted of just one nanowire with an approximate diameter of 160 nm. In addition to analyzing the diode performance, the effects of elevated temperatures and proton irradiation on the current–voltage characteristics of SiC NW Schottky diodes were also examined. The device could maintain similar values for ideality factor, barrier height, and effective Richardson constant upon proton irradiation with a fluence of 10 16 ion/cm 2 at 873 K. As a result, these metrics have clearly demonstrated the high-temperature tolerance and irradiation resistance of SiC NWs, ultimately indicating that they may provide utility in allowing solution-processable electronics in harsh environments.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2021

Irradiation Experiments on High Efficiency Nanowire Solar Cells Including Tilted Incidence Angle

Nanowire based solar cells have promising potential to become the next generation of space solar cells. In this study we corroborate their high radiation tolerance by irradiating high efficiency GaAs (efficiency > 15 %) nanowire solar cells with 1 MeV electrons at a fluence of 10 16 e - /cm 2 . Additionally, we theoretically show that the damage in nanowire solar cells is expected to be slightly dependent on the incidence angle and the geometry of the nanowire array. However, experimentally, two sets of GaAs nanowire solar cells have been irradiated with 1 MeV electrons at normal incidence and with a 25° tilted angle exhibiting a similar degradation.

high specific power↗
DOE OSTI2023

Accident Tolerant Fuel Test (ATF-2c) Irradiation Test Report

A new fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2C has completed its first cycle of prototypic steady state irradiation. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 24 test pins has been calculated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1980

Radiation tolerance of boron doped dendritic web silicon solar cells

The potential of dendritic web silicon for giving radiation hard solar cells is compared with the float zone silicon material. Solar cells with n(+)-p-P(+) structure and approximately 15% (AMl) efficiency were subjected to 1 MeV electron irradiation. Radiation tolerance of web cell efficiency was found to be at least as good as that of the float zone silicon cell. A study of the annealing behavior of radiation-induced defects via deep level transient spectroscopy revealed that E sub v + 0.31 eV defect, attributed to boron-oxygen-vacancy complex, is responsible for the reverse annealing of the irradiated cells in the temperature range of 150 to 350 C.

Rohatgi, A.↗
DOE OSTI2021

Development of a Reconstruction Methodology Based on X-Ray Computed Tomography to Generate Realistic 3D Concrete Microstructures in MOSAIC

Despite being passive components, concrete structures represent a major capital investment in nuclear power plants. However, some concrete components are critical for the safety and long-term operation of the reactors: the concrete containment building protects the reactors from external aggression. The concrete biological shield (CBS) contains the radiation exiting the reactor to protect equipment and personnel. Depending on the design, the CBS also has a function for supporting the reactor’s systems. This function needs to be ensured in-service and during accident scenarios. After several decades of use, the CBS is exposed to high-energy irradiation over an extended operation time, exceeding the fluence level currently considered detrimental for the concrete’s structural properties. Such conditions occur at different times depending on the reactor’s operation and design. Nevertheless, from the perspective of subsequent license application, it is mandatory to assess the potential effects of prolonged irradiation on concrete. Unlike steel components, concrete structures are built with local constituents (cement, sand, and coarse aggregates) to avoid prohibitive transportation costs. Thus, concrete performance is determined by the tolerance or the susceptibility of its constituents against any aging mechanism. The tolerance against irradiation of the CBSs may vary greatly among the 56 nuclear power plants and 93 nuclear power reactors that form the current US light-water reactor fleet. Irradiation-induced degradation is mainly a result of amorphization-induced expansions of the silicated minerals present in the aggregates. The neutron irradiation–induced expansion varies from one rock-forming mineral to another: well framework–structured silicates (e.g., quartz and feldspars) exhibit higher swelling than chained (e.g., pyroxene) or isolated silicates (e.g., garnet). Silicates make up nearly 90% of the Earth’s crust. Other common minerals present in construction aggregates are carbonates found, for example, in limestone. Irradiated carbonates exhibit only very minor expansions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Proton Radiation Tolerance of Wide and Narrow Band Gap Perovskite Solar Cells

Two different single junction mixed organic-inorganic halide perovskite solar cells (PSCs) were assessed for their tolerance to high energy proton radiation. Both the narrow band gap mixed tin-lead iodide and wide band gap lead iodide-bromide show remarkable tolerance for fluences up to 1x10 11 p/cm 2 . Both materials show a retention of open circuit voltage and decrease in dark current densities indicating minimal effect on SRH recombination that typically plagues traditional III-V based solar cells. Additionally, a control device that was processed without the perovskite absorber layer showed better diode characteristics and lower dark current densities after proton irradiation. The tolerance of these perovskite as a class of materials is promising for multijunction tandem flexible photovoltaic devices.

perovskite↗
DOE OSTI2026

Synthesis, Characterization, and Testing of High-Lithium-Density Composite Breeders

Solid tritium breeder materials must first and foremost have sufficiently high concentrations of lithium to enable a plant-scale tritium breeding ratio greater than 1:1. However, in addition to lithium content, such breeder materials must also meet other performance metrics including high tritium release rates, thermal conductivities, and irradiation damage tolerance. Perhaps most importantly, tritium breeders must maintain their mechanical integrity during reactor operation so as to avoid degradation which can jeopardize the functionality of the tritium breeder blanket module, which in most designs takes the form of a pebble bed geometry. Unfortunately, the mechanical robustness of most lithium-bearing ceramics under investigation for fusion applications is often inversely related to the lithium atom density. For example, a material such as lithium oxide (Li2O), which has one of the highest lithium atom densities, has a much lower mechanical splitting strength than lithium metatitanate (Li2TiO3), though Li2TiO3 has less than half the lithium atom density of Li2O. This work seeks to provide an alternative to monolithic ceramic tritium breeders, in the form of metal-reinforced composite tritium breeders. Specifically, composite tritium breeders have been synthesized combining Li2O with various ferrous metal reinforcements via electric field assisted sintering (EFAS), also known as spark plasma sintering (SPS). As the metal reinforcement content is increased, metallic networks are observed, via electron microscopy and X-ray computed tomography, to form throughout the composite material. Through destructive mechanical testing, even dilute metal reinforcement loading enables drastic mechanical strength improvements over pure Li2O while higher loadings give rise to quasi-ductile behavior and higher ultimate strengths than Li2TiO3 – while still maintaining a higher density of lithium atoms than Li2TiO3 and many other breeder candidates. In addition to microstructural characterization and mechanical testing, thermal property measurements and hydrogen permeability testing are underway to further assess the suitability of such composites for fusion reactor applications.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗