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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Swift heavy ion irradiation effects in zirconium and hafnium carbides

The behavior of microcrystalline zirconium carbide (ZrC) and hafnium carbide (HfC) was studied under highly ionizing irradiation conditions at room temperature. The induced structural modifications were characterized via synchrotron-based X-ray diffraction experiments. Unit-cell expansion and buildup of microstrain were determined across a wide fluence range and linked to chemical compositions of the target material. The observed swelling resulting from irradiation with 198 MeV Xe ions in both carbide materials is characterized by two distinct mechanisms that operate within different fluence regimes. Unit-cell expansion initially proceeds by a direct-impact behavior that reaches saturation, followed at higher fluences by a second, linear swelling regime. The overall behavior, particularly the direct-impact regime, is similar for ZrC and HfC, with a more pronounced second defect accumulation process in HfC. Swelling in ZrC shows the same two distinct mechanisms upon irradiation with 198 MeV Xe ions and 946 MeV Au ions, but swelling induced by the lower-energy ions is greater across the entire fluence series. Accounting for the difference in energy deposition density between the two irradiation conditions reveals that the first swelling mechanism (direct-impact behavior) is likely related to the formation of more simple defects. In contrast, the second damaging mechanism at higher fluences (linear increase) cannot be fully explained by the induced energy density, and swelling remains somewhat higher for the low-velocity Xe irradiation. This may suggest that more complex defects and defect clusters are responsible for this swelling regime, with either their size and/or morphology modified at different energy densities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

In-situ ion irradiation induced nanograin growth in a spent UO 2 fuel

This study investigates the irradiation-driven evolution of nanograins in the early-stage restructured rim region of medium burnup spent uranium dioxide (UO 2 ) fuel. Transmission electron microscopy (TEM) lamellas prepared from Belgium Reactor 3 (BR-3) fuel were subjected to in-situ 300 keV Xe ion irradiations under varying doses, fluxes, and temperatures to evaluate their effect on the evolution of the nanograins. Our results reveal that the nanograins grow during ion irradiation. Additionally, the growth is most pronounced at elevated temperatures (300 °C), moderate at room temperature, and negligible at cryogenic temperature (−223 °C). This behavior indicates that thermal activation, alongside irradiation effects, is essential to overcome grain boundary pinning by fission gas bubbles, metallic precipitates, and porosity. Furthermore, while nanograins (<200 nm) consistently coarsened under irradiation, larger grains did not undergo further restructuring, which can be attributed to the strong defect annihilation at TEM lamella surfaces combined with the limited electronic stopping power of low energy Xe ions used in this work. These findings highlight the roles of thermal spike effects, defect mobility, and impurity pinning in governing grain evolution in the rim region of spent UO 2 fuel during ion irradiation, providing key insights for predictive models of restructuring and performance of the high burnup nuclear fuel.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

The role of excess vacancies in stabilizing solute clusters in low-alloy steels

The formation of solute clusters in irradiated low-alloy steels, such as reactor pressure vessel steels, is a critical cause of radiation hardening and embrittlement. However, the chemical interactions among solute elements and excess vacancies remain to be fully understood. This study employs density functional theory, cluster expansion, and lattice-based Monte Carlo simulations to examine the stability and morphology of nano-size coherent solute clusters with excess vacancies. The findings reveal that excess vacancies are crucial in stabilizing and promoting the growth of Mn-Ni-Si-vacancy clusters. A minimum of seven vacancies is required for stable nucleation and growth of these clusters. These Mn-Ni-Si clusters act as defect sinks, effectively trapping and absorbing mobile vacancies generated under irradiation. Furthermore, phosphorus (P) preferentially dissolves in Mn-Ni-Si clusters due to chemical coupling with vacancies. In conclusion, this study offers new insights into solute-vacancy interactions, enhancing our understanding of solute-defect cluster formation and embrittlement in low-alloy steels.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Revealing a Pathway for Low-Temperature Recrystallization in Germanium

Thermally activated annealing in semiconductors faces inherent limitations, such as dopant diffusion. Here, a nonthermal pathway is demonstrated for a complete structural restoration in predamaged germanium via ionization-induced recovery. By combining experiments and modeling, this study reveals that the energy transfer of only 2.4 keV nm −1 from incident ions to target electrons can effectively annihilate pre-existing defects and restore the original crystalline structure at room temperature. Moreover, it is revealed that the irradiation-induced crystalline-to-amorphous (c/a) transformation in Ge is reversible, a phenomenon previously considered unattainable without additional thermal energy imposed during irradiation. For partially damaged Ge, the overall damage fraction decreases exponentially with increasing fluence. Surprisingly, the recovery process in preamorphized Ge starts with defect recovery outside the amorphous layer and a shrinkage of the amorphous thickness. After this initial stage, the remaining damage decreases slowly with increasing fluence, but full restoration of the pristine state is not achieved. These differences in recovery are interpreted in the framework of structural differences in the initial defective layers that affect recovery kinetics. This study provides new insights on reversing the c/a transformation in Ge using highly-ionizing irradiation and has broad implications across materials science, radiation damage mitigation, and fabrication of Ge-based devices.

athermal recovery↗
DOE OSTI2021

Size-dependent radiation damage mechanisms in nanowires and nanoporous structures

Nanostructures with a high density of interfaces, such as in nanoporous materials and nanowires, resist radiation damage by promoting the annihilation and migration of defects. This study details the size effect and origins of the radiation damage mechanisms in nanowires and nanoporous structures in model face-centered (gold) and body-centered (niobium) cubic nanostructures using accelerated multi-cascade atomistic simulations and in-situ ion irradiation experiments. Our results reveal three different size-dependent mechanisms of damage accumulation in irradiated nanowires and nanoporous structures: sputtering for very small nanowires and ligaments, the formation and accumulation of point defects and dislocation loops in larger nanowires, and a face-centered-cubic to hexagonal-close-packed phase transformation for a narrow range of wire diameters in the case of gold nanowires. Smaller nanowires and ligaments have a net effect of lowering the radiation damage as compared to larger wires that can be traced back to the fact that smaller nanowires transition from a rapid accumulation of defects to a saturation and annihilation mechanism at a lower dose than larger nanowires. These irradiation damage mechanisms are accompanied with radiation-induced surface roughening resulting from defect-surface interactions. Comparisons between nanowires and nanoporous structures show that the various mechanisms seen in nanowires provide adequate bounds for the defect accumulation mechanisms in nanoporous structures with the difference attributed to the role of nodes connecting ligaments in nanoporous structures. Altogether, our results shed light on the compounded, size-dependent mechanisms leading to the radiation resistance of nanowires and nanoporous structures.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Radiation-Induced Long-Lived Transients and Metal Particle Formation in Solid KCl–MgCl 2 Mixtures

Here, the electron paramagnetic resonance and diffuse reflectance-optical absorption spectra of room-temperature γ-irradiated KCl–MgCl 2 binary solid salt mixtures (98:2 mol % and 2:98 mol %) and the eutectic (68:32 mol %) are reported. Additionally, powder X-ray diffraction of the pristine salts and thermal annealing studies of the irradiated salts were performed to evaluate the radiolysis product stability, annihilation, and association in metallic particles. The main long-lived transient species detected in 98:2 mol % KCl–MgCl 2 salts were perturbed F-centers, that is, trapped electrons (e t – ) in the vicinity of Mg ions (λ max at 561 nm), and the radiolytic reduction of Mg 2+ to Mg + and Mg 0 . Thermal annealing promoted the diffusion of defects to yield polycations (Mg n + ). On the other hand, irradiation of 2:98 mol % KCl/MgCl 2 salts showed the formation of cationic and neutral Mg dimers (Mg 2 + and Mg 2 ) and trimers as well as centers with a rhombic powder pattern apparently consisting of an electron shared between three Mg 2+ nuclei associated with an anion vacancy (ν a + –Mg 3 5+ ). Trapped electrons (e t – ) (F-centers) were not observed in the irradiated eutectic mixture; instead, Mg 2 , Mg 0 , and Cl 3 – were observed. It was observed that the higher temperature for thermal ionization of radiation-reduced Mg species decreased the extent of electron recombination reactions and the disproportionation of Cl 3 – compared to the pure KCl but enhanced the aggregation of Mg into larger metallic microstructures (metallic particles).

36 MATERIALS SCIENCE↗
DOE OSTI2024

Radiation-induced alteration of apatite on the surface of Mars: first in situ observations with SuperCam Raman onboard Perseverance

Abstract Planetary exploration relies considerably on mineral characterization to advance our understanding of the solar system, the planets and their evolution. Thus, we must understand past and present processes that can alter materials exposed on the surface, affecting space mission data. Here, we analyze the first dataset monitoring the evolution of a known mineral target in situ on the Martian surface, brought there as a SuperCam calibration target onboard the Perseverance rover. We used Raman spectroscopy to monitor the crystalline state of a synthetic apatite sample over the first 950 Martian days (sols) of the Mars2020 mission. We note significant variations in the Raman spectra acquired on this target, specifically a decrease in the relative contribution of the Raman signal to the total signal. These observations are consistent with the results of a UV-irradiation test performed in the laboratory under conditions mimicking ambient Martian conditions. We conclude that the observed evolution reflects an alteration of the material, specifically the creation of electronic defects, due to its exposure to the Martian environment and, in particular, UV irradiation. This ongoing process of alteration of the Martian surface needs to be taken into account for mineralogical space mission data analysis.

Science & Technology - Other Topics↗
DOE OSTI2023

A quinary WTaCrVHf nanocrystalline refractory high-entropy alloy withholding extreme irradiation environments

In the quest of new materials that can withstand severe irradiation and mechanical extremes for advanced applications (e.g. fission & fusion reactors, space applications, etc.), design, prediction and control of advanced materials beyond current material designs become paramount. Here, through a combined experimental and simulation methodology, we design a nanocrystalline refractory high entropy alloy (RHEA) system. Compositions assessed under extreme environments and in situ electron-microscopy reveal both high thermal stability and radiation resistance. We observe grain refinement under heavy ion irradiation and resistance to dual-beam irradiation and helium implantation in the form of low defect generation and evolution, as well as no detectable grain growth. The experimental and modeling results—showing a good agreement—can be applied to design and rapidly assess other alloys subjected to extreme environmental conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mg and native defects in cubic silicon carbide from first principles

The diffusion of Mg defects in 3C-SiC is studied using the density functional theory. Mg has the highest burn-in rate as a transmutant in 3C-SiC when it is placed in high-energy neutron irradiation environment of a fusion reactor. The presence and evolution of transmutant defects impact thermal and mechanical properties of this important structural material. This study is focused on understanding the structure, stability, and evolution of Mg defects and the interaction of Mg with native defects in 3C-SiC. Our calculations of diffusion coefficients for different Mg defects suggest that Mg is likely to diffuse faster in pristine 3C-SiC than in the damaged one, in agreement with earlier experimental observations.

36 MATERIALS SCIENCE↗
DOE OSTI2023

An Introduction to Radiation Damage in Minerals and Ceramics [Slides]

Radiation effects in minerals and ceramics are quite distinct from effects observed in metals. This is largely due to interactions between radiation and the electronic structure of insulating materials that are unique compared to metallic conductors. Most notably, insulators exhibit property changes when energetic particles lose energy to electrons in the solid, whereas metals are unaffected by electronic energy losses. In addition, displacive radiation damage effects (atoms being knocked off their lattice sites by energetic particles) are much more complex in insulators, due to the complexities of their crystal structures (multiple cation and anion sublattices). In this presentation, we will focus on the atomistic mechanisms responsible for radiation damage in minerals and ceramics, when energetic particles lose energy while penetrating a solid. We will consider several examples, including (i) irradiation induced color center formation in gemstones; (ii) latent defects in alkali halides (relevance to thermoluminescent detectors); (iii) radioactive decay-induced amorphization of natural minerals; and (iv) extended defect formation and phase stability in complex oxides.

36 MATERIALS SCIENCE↗
NASA NTRS1971

Recombination luminescence in irradiated silicon - Effects of thermal annealing and lithium impurity.

Use of luminescence in irradiated silicon to determine the thermal stability of the defects responsible for the recombination. It is found that the defect responsible for the zero-phonon line at 0.97 eV has an annealing behavior similar to that of the divacancy and that the zero-phonon line at 0.79 eV anneals in a manner similar to the G-15 or K-center. Annealing at temperatures up to 500 C generates other defects whose luminescence is distinct from that seen previously. Addition of lithium to the material produces defects with new characteristic luminescence. Of particular importance is a defect with a level at E sub g -1.045 eV.

Johnson, E. S.↗
DOE OSTI2024

Complex dislocation loop networks as natural extensions of the sink efficiency of saturated grain boundaries in irradiated metals

The development of radiation-tolerant structural materials is an essential element for the success of advanced nuclear energy concepts. A proven strategy to increase radiation resistance is to create microstructures with a high density of internal defect sinks, such as grain boundaries (GBs). However, as GBs absorb defects, they undergo internal transformations that limit their ability to capture defects indefinitely. Here, we show that, as the sink efficiency of GBs becomes exhausted with increasing irradiation dose, networks of irradiation loops form in the vicinity of saturated or near-saturated GB, maintaining and even increasing their capacity to continue absorbing defects. The formation of these networks fundamentally changes the driving force for defect absorption at GB, from “chemical” to “elastic.” Using thermally-activated dislocation dynamics simulations, we show that these networks are consistent with experimental measurements of defect densities near GB. Our results point to these networks as a natural continuation of the GB once they exhaust their internal defect absorption capacity.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Back to functional hydrides: Effects of neutron-irradiated microstructure on hydrogen retention in yttrium hydride

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

Defect clusters↗
DOE OSTI2021

Helium-bubble-enhanced disordering of intermetallic phase under irradiation

Understanding and controlling phase transformations in metals and alloys under irradiation is vital to the design of irradiation-tolerant materials. The L1 2 -ordered intermetallic phase in nickel-based superalloys provides superior high-temperature mechanical properties and corrosion resistance. Here, we studied the order-disorder transformation kinetics of an L1 2 -structured gamma-prime phase in nickel-based superalloy Rene N4 under krypton ion irradiation and the effect of pre-implanted helium gas bubbles on the kinetics of phase transformation. Furthermore, the presence of pre-implanted helium gas bubbles significantly enhances the order-disorder transformation of L1 2 -structured gamma-prime phase under irradiation at 300 °C. The evolution of pre-implanted helium gas bubbles under irradiation implies that helium gas bubbles could effectively trap irradiation-induced vacancies and impede the reordering process of anti-site defects in gamma-prime phase.

36 MATERIALS SCIENCE↗
DOE OSTI2022

AGR-5/6/7 Final Release-to-Birth Ratio Data Analysis

AGR-5/6/7 is the last of a series of Advanced Gas Reactor (AGR) experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low-enriched fuel for use in the high-temperature gas cooled reactor (HTGR). AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through the use of helium coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO-coated particle fuel. The AGR tests provide valuable irradiation-performance data to support fuel process development, qualify fuel for normal operation and accident conditions, and support development and validation of fuel performance and fission-product (FP) transport models and codes. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. Release-to-birth ratios (R/B) for fission-gas isotopes released from each capsule are calculated from release rates, measured by germanium detectors in the Fission Product Monitoring System (FPMS) installed downstream from each capsule, and birth rates calculated using numerical models of FG generation. The R/Bs are a critical measure of the ability of the fuel kernel, the particle coating layers, and the compact matrix to retain fission-gas atoms, preventing their release into the sweep-gas flow, and the impact of initially defective particles and/or particle-coating failures that occur during irradiation. For fission-gas isotopes, particle failure is defined as failure of all coating layers, allowing gaseous fission atoms to escape from a particle. During the first five cycles (162B ? 165A), R/Bs were stable in the 10-8?10-6 range, and no in-pile particle failures were observed, based on the gross gamma counts. The maximum R/B value of around 2 ? 10-6 for Kr-85m resulted from the presence of as-fabricated exposed kernels (based on the high exposed kernel fraction), the dispersed uranium, and high fuel particle temperatures in Capsule 1. Comparison of capsule-measured R/Bs from these early cycles to predictions using the previously developed AGR R/B model demonstrated FG release from the AGR-5/6/7 TRISO fuel was comparable to that of previous experiments. In addition, the Kr-85m R/B per-exposed-kernel values are comparable to R/B values obtained in AGR-3/4 irradiation experiment and four irradiation experiments performed during 1980s: (1) HRB-17/18, (2) COMEDIE-BD1, (3) HFR B1, and (4) HRB-21. In contrast, all measured R/B values are lower than predictions by the commonly used Richards and German models, which are intentionally conservative. A large number of in-pile particle failures occurred in Capsule 1 by the end of Cycle 166A. During the final four cycles (166A ? 168A), apparent damage to the Capsule 1 gas line appeared to cause FG leakage from that capsule into the other four capsules, resulting in an increase in fission gas (FG) detected in the effluent for all capsules. Isolation of the Capsule 1 gas line during the last three cycles also prevented measurement of its FG release. Thus, R/Bs in all capsules after Cycle 166A are highly uncertain because of undefined amount of leakage from Capsule 1, especially for long-lived isotopes. A few hundred in-pile particle failures were estimated for Capsule 1 before the end of Cycle 166A, but the total number of failures is unknown due to the lack of FG release data in the later cycles. Based primarily on evidence from the gross gamma counts during Cycle 168A, approximately 15 particles failed in Capsule 3 and four particles failed in Capsule 2. In-pile failures in Capsule 3 were anticipated because this capsule was designed to operate beyond the HTGR normal operating temperature range. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR model, developed in INL/EXT-14-32970, for R/B of the short-lived isotopes (Kr-89 and Xe-137) with minimal leakage from Capsule 1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Microstructural evolution of a silicon carbide-carbon coated nanostructured ferritic alloy composite during in-situ Kr ion irradiation at 300°C 450°C

This work focuses on irradiation behaviors of a novel silicon carbide and carbon coated nanostructured ferritic alloy (SiC-C@NFA) composite for potential applications as a cladding and structural material for next generation nuclear reactors. The SiC-C@NFA samples were irradiated with 1 MeV Kr ions up to 10 dpa at 300 and 450 °C. Microstructures and defect evolution were studied in-situ at the IVEM-Tandem facility at Argonne National Laboratory. The effects of ion irradiation on various phases such as α-ferrite matrix, (Fe,Cr) 7 C 3 , and (Ti,W)C precipitates were evaluated. The α-ferrite matrix showed a continuous increase in dislocation density along with spatial ordering of dislocation loops (or loop strings) at >5 dpa. The size of the dislocation loops at 450 °C was larger than that at 300 °C. The nucleation and growth of new (Ti,W)C precipitates in α-ferrite grains were enhanced with the ion dose at 450 °C. This study provides new insight into the irradiation resistance of the SiC-C@NFA system.

(FeCr)(7)C-3↗
DOE OSTI2022

Grain boundary strain as a determinant of localized sink efficiency

The opportunity to achieve radiation tolerance in crystalline materials hinges on understanding the structure and response of grain boundary sinks to irradiation. A common descriptor of grain boundary efficiency as a defect sink is the denuded zone, which is a defect free zone adjacent to the grain boundary dictated by its ability to absorb radiation induced defects. This descriptor is often used at the mesoscale, which requires an averaging of absorption events. In this paper, we resolve the defect sink efficiency as a function of interfacial strain with respect to grain boundary character, and correlate high levels of grain boundary strain to an enhanced absorption efficiency. Here, we also introduce a key relationship between localized strain in proximity with the grain boundary sink and the variation absorption efficiency associated with these regions, revealing the pitfalls of averaging absorption events along a grain boundary, and presenting a path forward toward improved models for denuded zones and localized grain boundary absorption phenomena.

36 MATERIALS SCIENCE↗
NASA NTRS1977

EPR and transient capacitance studies on electron-irradiated silicon solar cells

One and two ohm-cm solar cells irradiated with 1 MeV electrons at 30 C were studied using both EPR and transient capacitance techniques. In 2 ohm-cm cells, Si-G6 and Si-G15 EPR spectra and majority carrier trapping levels at (E sub V + 0.23) eV and (E sub V + 0.38) eV were observed, each of which corresponded to the divacancy and the carbon-oxygen-vacancy complex, respectively. In addition, a boron-associated defect with a minority carrier trapping level at (E sub C -0.27) eV was observed. In 1 ohm-cm cells, the G15 spectrum and majority carrier trap at (E sub V + 0.38) eV were absent and an isotropic EPR line appeared at g = 1.9988 (+ or - 0.0003); additionally, a majority carrier trapping center at (E sub V + 0.32) eV, was found which could be associated with impurity lithium. The formation mechanisms of these defects are discussed according to isochronal annealing data in electron-irradiated p-type silicon.

Lee, Y. H.↗