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Results for “irradiation damage”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Demonstration of the surveillance test article for an advanced reactor surveillance test program

Operational environments in generation IV reactors involve corrosive and irradiative conditions at elevated temperatures. Typical reactor operations consist of transients which impose cyclic loads on reactor components. These cyclic loads, combined with corrosive and irradiative environments, result in synergistic degradation of component materials. However, limited data exists on the coupled damage effects on materials for reactor environments. While the surrogate material surveillance concept has been used in light water reactors to assess irradiation damage, existing material surveillance technologies are not suitable for in-situ monitoring of coupled material degradation. The materials surveillance program focuses on material degradation management and the estimation of remaining life of reactor components through surveillance test articles. This paper presents the design and analysis methodology of a bi-metal surveillance test article, which uses difference in thermal expansion coefficient between two metals to induce in-situ cyclic loads. This report presents the work conducted in FY 25, to test the surveillance test article in air and salt environments. These test specimens were evaluated after thermal cycle exposure and remaining life is measured through creep test.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Irradiation effects in tungsten—From surface effects to bulk mechanical properties

Advanced materials such as tungsten fibre-reinforced composites allow to overcome severe weaknesses of the baseline materials for plasma-facing components — copper and tungsten. The effect of the fusion environment on the mechanical properties of these materials, e.g. the embrittlement by neutron irradiation, plays a key role for the development of future fusion reactors. To simulate this effect, high-energy ions are used as a substitute for the displacement damage by neutrons. We propose the use of very fine tungsten wire as a possibility of studying the influence of irradiation damage on the mechanical properties. This is possible as they allow full-depth irradiation of almost the entire volume despite the limited penetration depth of ions. Geometrical size effects are mitigated due to the nanoscale microstructure of the wire. In addition, similar wire is used in tungsten fibre-reinforced composites. Thus, the investigation of irradiated wire can directly be used for the prediction of the bulk composite properties. For the proof of this concept tungsten wire with a diameter of 16 μm was electrochemically thinned to 5 μm and irradiated with 20.5 MeV W 6+ ions. The mechanical properties were subsequently determined by macroscopic tensile testing. Irradiation to 0.3, 1 and 9 dpa did not lead to a change of the mechanical behaviour. Both strength and ductility, the latter indicated by the reduction of area, were similar to the as-fabricated state.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Revealing the synergistic effects of sequential and simultaneous dual beam irradiations in tungsten via in-situ TEM

Elucidating the synergistic effects of different energetic beams on the radiation response of nuclear materials is critical for developing an improved methodology for their evaluation when exposed to extreme environments. This article describes in-situ sequential (He implantation followed by Kr irradiation and vice-versa) and simultaneous (heavy ion Kr irradiation and He implantation) dual beam irradiations performed on tungsten at 1223 K. Dislocation loop density, average area, and total loop damage as a function of irradiation history and dose/fluence are quantified. The loop Burgers vectors and cavity damage (cavity density, size and total change in volume) are also determined at the final maximum dose for each condition. The loop damage evolution was different in all cases, with the smallest loop damage observed in the simultaneous experiment. Annihilation of Kr generated dislocation loops during He implantation in the Kr + He experiment was an unanticipated observation that may be explained by the dynamic evolution of dislocation loop sink strengths and time-dependent defect fluxes. Dislocation loop raft formation, denuded zones near extended defects, and cavity damage are compared across the different conditions. The phenomena observed and discussed in this work will stimulate further experimental and computational modeling activities leading to improved fundamental understanding of the irradiation response of nuclear materials under reactor-similar environments.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Role of structural defects in mediating disordering processes at irradiated epitaxial Fe 3 O 4 / Cr 2 O 3 interfaces

Disordering processes in oxide materials are complicated by the presence of interfaces, which can serve as either point defect sinks or accumulation sites; the response depends on factors such as interfacial structure, chemistry, and termination. Here, we have characterized the disordering of epitaxial Fe 3 O 4 (111) / Cr 2 O 3 (0001) thin film heterostructures after 400 keV Ar 2+ radiation at room temperature. The density of misfit dislocations in both the Fe 3 O 4 overlayer and Cr 2 O 3 buffer layer is varied by changing the thickness of Cr 2 O 3 to be pseudomorphically strained to the Al 2 O 3 (0001) substrate (5 nm thick) or partially relaxed (20 nm thick), as confirmed by Bragg filtering analysis of scanning transmission electron microscopy images. In both cases, irradiation leads to damage accumulation on the Fe 3 O 4 side of the heterointerface, as shown by Rutherford backscattering spectrometry measurements in the channeling geometry. However, the interface with more misfit dislocations exhibits disordering at a faster rate than the less-defective interface. Likewise, layer-resolved electron energy loss spectroscopy reveals interfacial reduction of Fe after irradiation at the more defective interface. Intermixing of Cr across the interface is observed by atom probe tomography, which is likely facilitated by the generation of Cr interstitials in Cr 2 O 3 under irradiation.

36 MATERIALS SCIENCE↗
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]↗
NASA NTRS1990

Effects Of Dose Rates On Radiation Damage In CMOS Parts

Report describes measurements of effects of ionizing-radiation dose rate on consequent damage to complementary metal oxide/semiconductor (CMOS) electronic devices. Depending on irradiation time and degree of annealing, survivability of devices in outer space, or after explosion of nuclear weapons, enhanced. Annealing involving recovery beyond pre-irradiation conditions (rebound) detrimental. Damage more severe at lower dose rates.

Goben, Charles A.↗
DOE OSTI2021

Effect of sink strength on coherency loss of precipitates in dilute Cu-base alloys during in situ ion irradiation

In situ irradiations with 1 MeV Kr ions at 50~613 K up to a fluence of 6.25 × 10 14 ions/cm 2 (~1.25 displacements per atom, dpa) have been performed on pre-aged dilute Cu-0.9%Co, Cu-0.9%Fe and Cu-0.8%Cr alloys containing uniform matrix dispersions of coherent precipitates in order to study the effects of initial precipitate sink strength, damage dose and irradiation temperature on radiation-induced coherency loss of precipitates. Coherent precipitates with different point defect sink strengths (2πNd, where N and d are the precipitate density and diameter) were used in this work to examine potential differences in atomic relaxation during absorption of point defects. In all cases, irradiation to low doses (<~1 dpa) was very effective at inducing loss of precipitate coherency. At low sink strengths (~10 13 m -2 ), loss of precipitate coherency could be induced for doses ~0.01 dpa. This suggests there might be an efficient preferential medium-range strain-induced bias for absorption of interstitial defects due to tensile strains emanating from the undersized precipitates, which induces relatively rapid loss of coherency at low precipitate sink strengths. High precipitate sink strength (~10 14 m -2 ) conditions were relatively resistant to the radiation-induced loss of coherency (requiring higher doses approaching ~1 dpa) and this might be due to nearly equal numbers of interstitial and vacancy defects arriving at the precipitate interface for such high sink strength conditions. The precipitate coherency loss was observed to have a weak dependence on irradiation temperature. Molecular dynamics simulations confirm a strong effect of precipitate sink strength on the probability of interstitial absorption at precipitates.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Amorphization and siliconization of silicon carbide as a first wall material

The understanding and prediction of silicon carbide (SiC) material evolution exposed to SOL plasma conditions is of prime interest because SiC represents a promising main chamber wall plasma-facing material for next-step fusion devices (low hydrogenic diffusion, good mechanical and thermal properties under neutron irradiation). Gross and net Si erosion rates from SiC surfaces in contact with a well-diagnosed L-mode plasma in the DIII-D tokamak have been simulated and the surface concentrations of impurities have been tracked as a function of time. Coupled simulation of surface model and impurity transport demonstrates amorphization of crystalline SiC exposed to L-mode plasma due to the accumulation of displacement damages under ion irradiation. This affects the lifetime of SiC plasma facing components. Surface evolution is tightly coupled to impurity transport in the plasma and therefore needs to be integrated with impurity transport simulations to effectively predict Si erosion rates and sub-surface concentrations as a function of time. The simulation workflow couples a semi-analytical surface model to the impurity transport code GITR. The surface model is a homogeneous mixed-material model that tracks physical & chemical sputtering and reflection of impurities. Gross erosion is primarily influenced by the background plasma parameters and redeposition patterns are mainly influenced by the prompt redeposition due to the gyro-orbits of impurity ions. Although crystalline form of SiC is preferable for fusion wall applications because of resistance to neutron irradiation, this work indicates that crystalline SiC will undergo amorphization under D plasma contact with implications of higher sputtering and fuel retention. These results direct us to explore the effects of amorphization on crystalline SiC and further the physics basis of SiC usage as first wall material for fusion environments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

An examination of the precipitation behavior of proton irradiated dual phase 308L weldment filler materials

Voids, G phase particles, and Ni-Si rich clusters in proton irradiated dual phase 308L groove filler of a SA508–304 L dissimilar metal weldment are analyzed using advanced characterization techniques. These weldments are often used in light water nuclear reactors and are subject to enhanced corrosion and associated stress corrosion cracking (SCC). Radiation damage is known to accelerate SCC. Ni-Si enriched clusters were observed in proton irradiated γ austenite, while G phase M 6 Ni 16 Si 7 (where M transition metal element) precipitates were observed in proton irradiated δ ferrite. Compositional analysis of the G phase precipitates and Ni-Si clusters from STEM-EDS and APT are compared. Unlike G phase particles in proton irradiated δ ferrite, Ni-Si clusters in proton irradiated γ austenite are not rich in Mn. Both STEM-EDS and APT line-scan profiles of the Ni-Si clusters show that the Fe and Cr concentration gradient between matrix γ austenite and the Ni-Si clusters is not as sharp as those between matrix δ ferrite and G phase. Further, HR-STEM imaging indicates that the lattice parameter of the Ni-Si clusters is commensurate with γ austenite and the clusters do not represent the precipitation of a second phase. Finally, our analysis demonstrates the density and volume fraction of G phase particles and the density of voids scales with proton irradiation damage and the energy to recoils.

308L groove filler↗
NASA NTRS1991

Proton and electron damage coefficients for GaAs/Ge solar cells

A series of electron and proton irradiations of GaAs/Ge solar cells has been carried out for the purpose of deriving radiation damage coefficients for GaAs/Ge solar cells. The electron irradiations were performed at energies of 0.6, 1.0, 2.4, and 12 MeV. The proton irradiation energies used were 0.05, 0.2, 0.3, 0.5, 1.0, 3.0, and 9.5 MeV. I-V characteristics were measured before and after each irradiation. Omnidirectional damage coefficients were calculated for both electrons and protons incident on GaAs/Ge solar cells. The calculations were carried out for infinite backshielding on the cells and for eight different thicknesses of coverglass.

Anspaugh, Bruce E.↗
DOE OSTI2024

Sulfur speciation in dacitic melts using X-ray absorption near-edge structure spectroscopy of the S K -edge (S-XANES): Consideration of radiation-induced changes and the implications for sulfur in natural arc systems

Abstract The synchrotron technique of micro X-ray absorption near-edge structure spectroscopy at the sulfur K-edge (S-XANES) provides a unique opportunity to measure the proportion of different oxidation states of sulfur (S) in silicate glasses. Although applied extensively in the analysis of basaltic silicate glasses, few S-XANES studies have investigated variations in S oxidation states with fO2 in felsic silicate glasses. In addition, no study has systematically compared the S-XANES results obtained from the same samples at different photon flux densities to quantify the relationship between exposure time and changes in S speciation in silicate glass, as has been done for Fe and V. This study evaluates observed differences in S speciation measured in experimentally produced H2O-saturated dacitic glasses over a range of reducing to oxidizing conditions (from log fO2 = ΔFMQ-0.7 to ΔFMQ+3.3; FMQ is the fayalite-magnetite-quartz mineral redox buffer) and equilibrated at 1000 °C and 300 MPa. S-XANES spectra were collected at three different photon flux densities using three microspectroscopy beamlines. As is observed in S-XANES analyses of basaltic silicate glasses, beam-induced changes to the S6+/ΣS are observed as a function of photon flux density and beam exposure time. Our results demonstrate that silicate glasses of dacitic composition undergo beam-induced photo-reduction in samples equilibrated at ΔFMQ > +1.75 and photo-oxidation if equilibrated at ΔFMQ < +1. The time required to observe beam-induced changes in the spectra varies as a function of flux density, and our study establishes an upper photon density limit at ~1.0 × 1012 photons/µm2. The S6+/ΣS calculated from spectra collected below this absorbed photon limit at intermediate flux densities (~1–4 × 109 photons/s per µm2) are affected by beam damage, as no conditions were found to be completely free of beam-induced changes. However, the S6+/ΣS ratios calculated below the limit at intermediate flux densities are consistent with thermodynamic constraints, demonstrating that S6+/ΣS ratios calculated from S-XANES spectra can be considered reliable for estimating the oxygen fugacity. Our results carry important implications for the S budget of felsic magmas and dissolution mechanisms in evolved melts. While our results from all three flux densities show the presence of S4+ dissolved in relatively oxidized (ΔFMQ > +1.75) dacitic glass, even in the spectra exposed to the lowest photon densities, we are unable to rule out the possibility that the S4+ signal is the result of instantaneous X-ray irradiation induced beam damage using S-XANES alone. When our spectra are compared to S-XANES spectra from basaltic silicate glasses, important differences exist in the solubility of S2– and S6+ between dacitic silicate melts, pointing to differences in solubility mechanisms as melt composition changes. This study highlights the need for further investigation into beam damage systematics, presence of S4+, and the solubility mechanisms of different oxidation states of S as silicate melt composition changes.

Geochemistry & Geophysics↗
DOE OSTI2024

Unraveling radiation damage and healing mechanisms in halide perovskites using energy-tuned dual irradiation dosing

Perovskite photovoltaics have been shown to recover, or heal, after radiation damage. Here, we deconvolve the effects of radiation based on different energy loss mechanisms from incident protons which induce defects or can promote efficiency recovery. We design a dual dose experiment first exposing devices to low-energy protons efficient in creating atomic displacements. Devices are then irradiated with high-energy protons that interact differently. Correlated with modeling, high-energy protons (with increased ionizing energy loss component) effectively anneal the initial radiation damage, and recover the device efficiency, thus directly detailing the different interactions of irradiation. We relate these differences to the energy loss (ionization or non-ionization) using simulation. Dual dose experiments provide insight into understanding the radiation response of perovskite solar cells and highlight that radiation-matter interactions in soft lattice materials are distinct from conventional semiconductors. These results present electronic ionization as a unique handle to remedying defects and trap states in perovskites.

14 SOLAR ENERGY↗
DOE OSTI2020

Fluence Capsules for the SAM-2 Experiment MPO# 00239221/SOW-17494, Rev. 0

The statement of work (SOW) applies to the fabrication and analysis of flux wire capsules that will be used in the SAM-2 experiment. This experiment is a drop-in design slated for irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The 8 neutron fluence capsules will be used to determine the neutron exposure that the specimens have received. For Task 1 Pacific Northwest National Laboratory (PNNL) will prepare eight (8) neutron fluence monitors for irradiation in the ATR. Task 2 will involve analysis port-irradiation; the monitors will be returned to PNNL for analysis in FY21 after irradiation. The analysis report will include the neutron fluences and radiation damage calculations for all irradiated positions in the experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Irradiation-induced chemical disordering in ceramics: The case of SiC

Chemical disordering is one feature of damage that occurs in irradiation environments and is common in ceramic compounds. This paper reviews irradiation-induced chemical disordering in SiC-based materials at elevated temperatures. Here the results were obtained using advanced analytical tools to characterize atomistic defects and to evaluate the chemistry of defect clusters and homonuclear bond structures. Finally, this paper demonstrates that chemical disordering is crucial to understanding microstructural stability and establishing a mechanistic modeling of microstructural evolution of irradiated SiC.

36 MATERIALS SCIENCE↗