X-ray characterization of anisotropic defect formation in SiC under irradiation with applied stress
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99 GENERAL AND MISCELLANEOUS↗
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Results for “Irradiation induced defects”
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Study of lithium mobility in irradiated silicon
Effects of fast neutron and electron induced defects on diffusion properties of lithium in silicon
FAST NEUTRON↗
Alpha-decay induced shortening of fission tracks simulated by in situ ion irradiation
The diffusion of defects (e.g., vacancies and interstitials) and elements used for dating (e.g., He and Pb) in a mineral structure is a thermal process: This is the primary assumption used to determine the age and thermal history of minerals. For instance, thermal history reconstruction, through the number and length distribution of tracks produced by spontaneous fission of 238 U, is obtained by assuming a thermal event to be the only energy source for shortening of fission tracks. Here, we report a new, non-thermal energy source that induces additional shortening of fission tracks by the irradiation of alpha-recoils from the alpha-decay of 238 U and 232 Th. We simulate alpha-decay induced track-shortening by combining ion accelerator irradiations with transmission electron microscopy. This allows for the first observation of track-shrinkage during in situ ion irradiation. We show that rather than alpha-particles, alpha-recoils induce a significant shortening of fission tracks by nuclear-collisions. The shortening of track-length can be quantified as a function of alpha-decay event dose. However, apatite is less sensitive than zircon to this non-thermal process. Furthermore, the findings exemplify the interactions among different types of self-irradiation from alpha-particles, alpha-recoils and fission-fragment nuclei in single mineral grains and have important implications for the use of zircon and apatite for radiometric dating and thermochronology.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Phase-field modeling of radiation-induced composition redistribution: An application to additively manufactured austenitic Fe–Cr–Ni
Multicomponent alloys undergoing irradiation damage develop radiation-induced composition redistribution at point defect sinks such as grain boundaries (GBs) and dislocations. Such redistribution results in undesired changes to their mechanical behavior and corrosion resistance. Additively manufactured alloys proposed for future nuclear applications are expected to demonstrate a distinct response to irradiation owing to their unique microstructure with as-solidified dislocation density and chemical microsegregation. To capture the composition redistribution in such systems, we develop a mesoscale model with coupled evolution of atomic and point defect components in the presence of dislocation density, dislocation heterogeneity, and thermodynamic interactions at the GB. The model is parameterized for an FCC Fe–Cr–Ni alloy as a representative system for austenitic stainless steels, and simulations are performed in 1D and 2D as a function of irradiation temperature, dose, dislocation density, and grain size. Radiation-induced segregation (RIS) characterized by Cr depletion and Ni enrichment is predicted at both the GB and the dislocation cell wall, with RIS being lower in magnitude but wider at the cell wall. Strongly biased absorption of self-interstitials by dislocations is found to suppress Ni enrichment but slightly enhance Cr depletion under certain conditions. Thermodynamic segregation at the GB is predicted to be narrower and opposite in sign to RIS for both Cr and Ni. Importantly, non-monotonic segregation is found to occur when both thermodynamic and RIS mechanisms are considered, providing a novel physical interpretation of experimental observations. The model is expected to serve as a key tool in accelerated qualification of irradiated materials.
additively manufactured microstructure↗
Dislocation loop evolution in F/M steel T91 under in-situ ion irradiation: Influence of the presence of initial dislocations
Ferritic/Martensitic (F/M) steel T91 was irradiated in-situ to 4 and 10 dpa at 470 °C using 1 MeV Kr 2+ . Additionally, the microstructure evolution under irradiation was followed and characterized by in-situ Transmission Electron Microscopy (TEM), in terms of dislocation loop Burgers vector, average size and density as a function of dose. Furthermore, the influence of the presence of dislocations in the pre-irradiated matrix was investigated by comparing the radiation induced defect evolution in grains with and without initial dislocations.
36 MATERIALS SCIENCE↗
Effects of Galactic Irradiation on Thermal and Electronic Transport in Tungsten
The impact of irradiation on the thermal and electronic properties of materials is a persistent puzzle, particularly defect formation at the atomic and nanoscales. This work examines the nanoscale effects of low-energy irradiation on tungsten (W), focusing on defect-induced modifications to thermal and electronic transport. Using the site-projected thermal conductivity (SPTC) method [A. Gautam et al. PSS-RRL, 2400306, 2024], bulk and twin-grain boundary W with vacancy defects based on the Norgett–Robinson–Torrens displacements per atom model is analyzed. SPTC provides a detailed prediction of postcascade spatial thermal conductivity distribution. Electronic conductivity activity is estimated using the “N 2 method” [K. Nepal et al. Carbon, 119711, 2025] to explore the consequences of vacancies and grain boundaries, highlighting the defect-dependent nature of charge transport behavior. These findings offer high-resolution insights into irradiation-driven transport phenomena, with implications for space-exposed materials and nanoscale thermal/electronic management.
75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
The mechanism behind the high radiation tolerance of Fe-Cr alloys
With the great demand for high radiation tolerant materials for advanced nuclear energy technologies, Fe-Cr alloys are at the forefront with long standing validated performance. Yet, the real mechanism behind their high radiation resistance is in question and understanding the effect of varying Cr percentage is a grand challenge limiting further improvements. Here we applied depth resolved atomic scale probe of defects to uncover the real mechanism on how Cr improves radiation resistance and explain the controversial impact of increasing Cr percentage. By combining depth-resolved positron annihilation lifetime spectroscopy and Doppler broadening spectroscopy we investigated the effect of Cr alloying on the formation and evolution of atomic size clusters induced by ion irradiation in Fe. We also used atom probe tomography to investigate the possible presence of Cr clusters or a’ phase with high Cr composition. The study reveals that the well-known resistance to radiation in Fe-Cr alloys arises from the stabilization of vacancy clusters around Cr atoms which act as sinks for radiation-induced defects. Thus, Cr atoms do not provide a direct sink for interstitials; rather defect complexes for that consist of Cr atoms and vacancies in turn act as sinks for irradiation-induced vacancies and interstitials. Most importantly, we find that lower amounts of Cr create smaller, uniformly distributed defect clusters that act as efficient sinks for radiation damage, but larger quantities of Cr form a defect structure that is less homogenous in size and spatial distribution, resulting in less efficient damage recombination. No evidence of phase a’ was found before or after irradiation, which indicates that it does not contribute to the observed radiation tolerance.
36 MATERIALS SCIENCE↗
Effect of microstructure and neutron irradiation defects on deuterium retention in SiC
Retention of hydrogen isotopes is a critical concern for operating fusion reactors as retained tritium both activates components and removes scarce fuel from the fuel cycle. Radiation-induced displacement damage in SiC influences the retention of hydrogen isotopes compared to pristine SiC. Deuterium retention in neutron irradiated high purity SiC has been compared to different microstructures of non-irradiated high purity SiC using thermal desorption spectroscopy after gas charging and low energy ion implantation. Experimental results show lower deuterium retention in single crystal SiC than in polycrystal SiC indicating that grain boundaries are key trapping features in unirradiated SiC. Deuterium is released at lower temperatures in neutron irradiated polycrystal SiC compared to pristine polycrystal SiC, suggesting weaker trapping by radiation-induced defects compared to grain boundary trapping sites in the pristine materials. Low energy ion implantation caused a high deuterium release temperature, highlighting the sensitivity of deuterium release behaviour to radiation defect characteristics. First principles calculations have been conducted to identify energetically favourable trapping sites in SiC at the H ABc V Si and H TSi V C complexes, and migration barriers between interstitial sites. This helps interpret experimental results and derive effective diffusivity of hydrogen isotopes in SiC in the presence of vacancies.
36 MATERIALS SCIENCE↗
Migration energy barriers and diffusion anisotropy of point defects on tungsten surfaces
We report tungsten is one of the most promising candidates for plasma-facing materials in future fusion devices, owing to its high performance under extreme irradiation conditions. However, irradiation-induced surface morphology varies significantly, depending on the irradiation type, fluence, and flux. Therefore, it is critical to examine the dynamics of point defects on tungsten (W) surfaces to understand how irradiation affects surface morphology. In this study, we employ the Self-Evolving Atomistic Kinetic Monte Carlo (SEAKMC) method to search for potential migration paths of point defects on W (1 0 0), (1 1 0) and (1 1 1) surfaces. The first-principles calculations are then used to accurately determine the migration energy barriers. The obtained paths and barriers are incorporated into a kinetic Monte Carlo (KMC) model to determine trajectories and diffusivities, which are described by diffusion tensors to demonstrate their anisotropic features. Multiple diffusion mechanisms have been identified on different surfaces, with various anisotropy factors. Particularly, point defects on the W (1 1 0) surface have the highest diffusivities, with anisotropy factors independent of temperatures. In comparison, the anisotropy of the W (1 0 0) surface decreases as temperature increases, while the W (1 1 1) surface is isotropic for point defects. This study provides insights into defect transport properties on different surfaces, which are essential for understanding the early stages of irradiation-induced microstructural evolutions and surface morphology of tungsten.
36 MATERIALS SCIENCE↗
Accelerated radiation tolerance testing of Ti-based MAX phases
MAX phases have recently attracted significant attention for potential nuclear applications due to their novel properties such as unique hexagonal-compact nanolayered crystal structure, high-machinability due to lower hardness levels compared to conventional ceramics, and high-chemical inertness. In order for MAX phases to be used in nuclear reactors, two aspects deserve detailed investigations: (i) their phase stability at high-temperatures and (ii) microstructural defect formation and recovery induced by energetic particle irradiation. To date, degradation mechanisms of MAX phases at high-temperatures and following irradiation are largely unexplored fields of research. This work focuses on the evaluation of two Ti-based MAX phases – Ti 2 AlC and Ti 3 SiC 2 – within the context of extreme environments. To accomplish this, a one-of-a-kind comparison between neutron irradiations, performed over a decade of research at the High Flux Isotope Reactor (HFIR), and heavy-ion irradiations, carried out in situ in a Transmission Electron Microscope (TEM) at the MIAMI-2 facility, has been conducted. The results show Ti-based MAX phases are prone to accelerated decomposition under the conditions investigated. This questions the hypothesis that MAX phases exhibit high phase stability, especially when used in future nuclear energy systems where energetic particle irradiation is a dominating degradation mechanism.
36 MATERIALS SCIENCE↗
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↗
Screening and qualification methodology for SiC end plug processing methods
Deployment of SiC-ceramic-based fuel cladding for light water reactors requires a hermetic end plug–to–cladding joint that can withstand neutron irradiation during normal operation and maintain integrity during design-basis accidents. Reactor experiments have shown that some SiC composite tubes with SiC end plugs can retain hermeticity after irradiation. However, achieving consistent joint performance under irradiation remains a key challenge. Resolving this issue is essential to enable integral irradiation testing and to demonstrate fuel integrity under commercial-reactor irradiation conditions. This report aims to: (1) provide guidance for designing radiation-tolerant end plug joints for SiC cladding; (2) demonstrate experimental methods to detect processing defects that are unstable under neutron irradiation at light-water-reactor-relevant temperatures and doses; and (3) outline a step-by-step approach for designing and conducting reactor experiments to screen joining methods. The resulting data will be used to improve joint processing and to define critical defect types and sizes that must be detected and eliminated through non-destructive evaluation for quality assurance. Based on prior irradiation experiments at the High Flux Isotope Reactor, differential swelling among the cladding, bonding layer, and end plug was identified as an underlying mechanism for irradiation-induced joint degradation. Accordingly, this effect must be considered in the design of radiation-tolerant joining techniques. In this work, miniature SiC end plug joint specimens irradiated during the previous project were analyzed using X-ray computed tomography to characterize the joint microstructure. Digital volume correlation of the tomography data quantified radiation-induced microstructural changes and enabled evaluation of defect-related risks. Finally, ongoing neutron irradiation efforts using larger specimen volumes are presented. These efforts aim to statistically assess joint performance and to build a microstructure–performance (e.g., leak-tightness) dataset to inform processing improvements and quality control.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Electron beam irradiation effects on bulk metals: a comparative study of polycrystalline versus single-crystalline structures
This study investigates the effects of electron beam (e-beam) irradiation on the mechanical and structural properties of eight bulk metallic samples, comprising both polycrystalline (PC) and single-crystalline (SC) forms of Ni, Cr, V, and Ti. These metals were evaluated as potential candidates for beam exit windows in high-power (MW-class) particle accelerators. The primary objective is to identify metals capable of withstanding the conditions of high-power/MW-class e-beam accelerators and serve effectively as exit windows. Selection criteria were based on each metal’s intrinsic properties, power dissipation capability, and irradiation-induced changes in mechanical behavior, including hardness, elastic modulus, and defect density. Comprehensive characterization was conducted using field-emission scanning electron microscopy (FESEM), atomic force microscopy (AFM), scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM–EDS), X-ray diffraction (XRD), high-resolution transmission electron microscopy (HRTEM), selected area electron diffraction (SAED), and nanoindentation, performed both before and after exposure to a ~¿66 kGy dose from a 10 MeV e-beam accelerator. Results revealed that e-beam irradiation induced hardening in PC Ni, whereas PC Ti, commonly used in beam exit windows, exhibited softening. The observed softening in PC Ti is attributed to grain coarsening, elongation, and the formation of twins and twin boundaries, in contrast to the smaller, compressed grains in the pristine (Pr) PC Ti samples, consistent with the Hall–Petch relationship. The stresses due to twinning are small and insignificant in influencing the overall hardening of the PC Ti irradiated sample when compared to the stresses due to the dislocation density. Conversely, SC Ti samples exhibited irradiation-induced hardening. The SC Ti irradiated samples developed additional irradiation-induced modifications in crystallographic texture of (100), (101), (110), (200), (112), (004), and (211) as evidenced from the XRD results, which could probably explain the hardening effect that is caused by irradiation.
36 MATERIALS SCIENCE↗
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↗
Dynamic Deformation in Nuclear Graphite and Underlying Mechanisms
Time-dependent deformation in nuclear graphite is influenced by the creation and migration of radiation-induced defects in the reactor environment. This study investigates the role of pre-existing defects such as point defect clusters and Mrozowski cracks in nuclear graphite IG-110. Separate specimens were irradiated with a 2.8 MeV Au 2+ beam with a fluence of 4.38 × 10 14 cm −2 and an 8 MeV C 2+ beam with a fluence of 1.24 × 10 16 cm −2 . Microscopic specimens were either mechanically loaded inside a transmission electron microscope (TEM) or subjected to ex situ indentation-based creep loading. In situ TEM tests showed significant plasticity in regions highly localized around the Mrozowski cracks, resembling slip or ripplocation bands. Slip bands were also seen near regions without pre-existing defects but at very high stresses. Ex situ self-ion irradiation embrittled the specimens and decreased the creep displacement and rate, while heavy ion irradiation resulted in the opposite behavior. We hypothesize that the large-sized gold ions (compared to the carbon atoms) induced interplanar swelling as well as cross-plane channels for increased defect mobility. These findings illustrate the role of pre-existing defects in the dynamic relaxation of stresses during irradiation and the need for more studies into the radiation environment’s impact on the mechanical response of nuclear graphite.
creep deformation↗
Plasmonic Ag nanocomposite phosphate glasses produced via γ-ray irradiation as reduction route
This paper reports on the impact of γ-ray irradiation (10, 100 kGy) on melt-quenched Ag + -doped phosphate glass and the effects of subsequent thermal processing leading to the production of plasmonic Ag nanocomposites. The γ-irradiated glasses were characterized alongside the pristine by differential scanning calorimetry (DSC), Raman spectroscopy, electron paramagnetic resonance (EPR) spectroscopy, optical absorption, and photoluminescence (PL) spectroscopy. DSC characterization showed consistent glass transition temperatures (T g ) before and after γ-irradiation whereas the crystallization temperatures tended to decrease with increasing γ-ray dose. However, a lack of alteration of the glass network structure was supported by Raman spectroscopy. Room temperature EPR spectra clearly showed the formation of phosphorus oxygen hole center (POHC) defects in the undoped host, in addition to another doublet likely associated with a P 3 defect. The presence of paramagnetic silver species encompassing Ag 2+ and 107/109 Ag 0 atoms was also indicated in the silver-activated glass together with POHC defects. Optical absorption spectra were also consistent with the presence of various radiation-induced centers. Further analyzing the glass absorption edge via Tauc plots suggested the formation of electron center (EC) defects in γ-irradiated samples wherein the silver-doped glass exhibited decreasing band gap energies with increasing γ-ray dose. The PL characterization showed the silver-related radio-PL effect was induced exhibiting broad band emission with two maxima around 500 and 625 nm stemming from various molecular Ag$^{x+}_{n}$ clusters. Emission decay analyses revealed that the longer wavelength emission exhibited slower decay. The highest radiation dose of 100 kGy however resulted in weaker emission and faster decay kinetics attributed to energy transfer between the luminescent silver species and POHC defects. Finally subjecting the γ-irradiated Ag-doped glasses to heat treatment near the T g at 490 °C led to the development of the surface plasmon resonance of Ag nanoparticles (NPs) and the vanishing of the Ag$^{x+}_{n}$ clusters luminescence. In conclusion, the presence of the matrix-related EC defects was deemed accountable for the thermally induced reduction and consequent precipitation of Ag NPs making the plasmonic glasses attractive for photonic applications such as nonlinear optics.
36 MATERIALS SCIENCE↗
Tailoring the Electronic Structures and Spectral Properties of ZnO with Irradiation Defects Generated Under Intense Electronic Excitation: A Combined Experimental and DFT Approach
Here, the relationship between the composition of the internal defect states, spectral properties, and correlated electronic structures of wurtzite zinc oxide (ZnO) crystals under 645 MeV Xe 35+ irradiation is systematically investigated, employing experimental characterizations combined with first-principle calculations. Based on the ion irradiation-induced thermal expansion and relaxation processes, the high concentration of vacancy/interstitial defects produced from the transient disordered phase in molten track states trigger photoelectric changes, as follows: i) the generation of internal defect states effectively reduces the intrinsic bandgap (3.25 eV → 2.66 eV); ii) a large number of defective active sites inhibits the recombination between electron–hole pairs, causing dark conductance and photoconductance to increase with increasing damage levels until optimal fluence is achieved. Based on the density functional theory (DFT) with the GGA + U (GGA = generalized gradient approximation) method, the defective models associated with the different electronic structures, density of states, formation energy, and the nature of the chemical bonding are established. The narrowing of the bandgap observed experimentally and the enhancement of carrier concentration originating from the internal electron defect states are qualitatively verified, therefore laying the foundation for designing future nanoscale photoelectronic devices and microelectronics applications.
36 MATERIALS SCIENCE↗
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↗