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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 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Nanoconfinement-induced shift in photooxidative degradation pathway of polystyrene

Polymer nanocomposites with high concentrations of nanoparticles (NPs) possess exceptional mechanical, transport, and thermal properties. To enable their widespread use in structural applications and functional coatings, it is crucial to understand how nanoconfinement and the polymer-NP interface influence polymer degradation under various environmental conditions, including prolonged UV exposure. In this study, we investigate the photooxidative degradation of polystyrene (PS)-confined in the interstices of SiO 2 NP films. These nanocomposite films are prepared by the capillary rise infiltration (CaRI) of PS into interstices of SiO 2 NP packings, and subsequently subjected to UV irradiation. Our investigation reveals that PS degradation progresses uniformly across the thickness, with degradation initiating from the center of the NP interstitial pores and extending towards the NP surface. We rationalize this degradation mechanism based on the disparity in the surface energies of PS and the NP surface, as well as slow oxygen diffusion through confined PS. Here, we also demonstrate that packing smaller NPs at a given thickness or thicker packing of a specific NP size facilitates photooxidative degradation, highlighting the critical role of the number of interstitial pores in influencing degradation processes.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Effect of defect imbalance on void swelling distributions produced in pure iron irradiated with 3.5 MeV self-ions

Ion irradiation has been widely used to simulate radiation damage induced by neutrons. However, there are a number of features of ion-induced damage that differ from neutron-induced damage, and these differences require investigation before behavior arising from neutron bombardment can be confidently predicted from ion data. In this study 3.5 MeV self-ion irradiation of pure iron was used to study the influence on void swelling of the depth-dependent defect imbalance between vacancies and interstitials that arises from various surface effects, forward scattering of displaced atoms, and especially the injected interstitial effect. The depth dependence of void swelling was observed not to follow the behavior anticipated from the depth dependence of the damage rate. Void nucleation and growth develop first in the lower-dose, near-surface region, and then, during continued irradiation, move to progressively deeper and higher-damage depths. This indicates a strong initial suppression of void nucleation in the peak damage region that continued irradiation eventually overcomes. This phenomenon is shown by the Boltzmann transport equation method to be due to depth-dependent defect imbalances created under ion irradiation. These findings thus demonstrate that void swelling does not depend solely on the local dose level and that this sensitivity of swelling to depth must be considered in extracting and interpreting ion-induced swelling data.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Molecular dynamics modeling of atomic displacement cascades in 3C-SiC: Comparison of interatomic potentials

We used molecular dynamics modeling of atomic displacement cascades to characterize the nature of primary radiation damage in 3C-SiC. We demonstrated that the most commonly used interatomic potentials are inconsistent with ab initio calculations of defect energetics. Both the Tersoff potential used in this work and a modified embedded-atom method potential reveal a barrier to recombination of the carbon interstitial and carbon vacancy which is much higher than the density functional theory (DFT) results. The barrier obtained with a newer potential by Gao and Weber is closer to the DFT result. This difference results in significant differences in the cascade production of point defects. We have completed both 10 keV and 50 keV cascade simulations in 3C-SiC at a range of temperatures. In contrast to the Tersoff potential, the Gao-Weber potential produces almost twice as many C vacancies and interstitials at the time of maximum disorder (~0.2 ps) but only about 25% more stable defects at the end of the simulation. Only about 20% of the carbon defects produced with the Tersoff potential recombine during the in-cascade annealing phase, while about 60% recombine with the Gao-Weber potential.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2015

Characterization and comparative analysis of the tensile properties of five tempered martensitic steels and an oxide dispersion strengthened ferritic alloy irradiated at ≈295 °C to ≈6.5 dpa

Tensile test results at 25 and 300 °C on five 9-12Cr tempered martensitic steels and one 14Cr oxide dispersion strengthened alloy, that were side-by side irradiated to 6.5 dpa at 295 °C in the Advanced Test Reactor (ATR), are reported. The engineering stress–strain curves are analyzed to provide true stress–strain constitutive σ(ε) laws for all of these alloys. In the irradiated condition, the σ(ε) fall into categories of: strain softening, nearly perfectly plastic and strain hardening. Increases in yield stress (Δσ y ) and reductions in uniform strain ductility (e u ) are observed, where as the latter can be understood in terms of the alloy's σ(ε) behavior. Increases in the average σ(ε) in the range of 0–10% strain are smaller than the corresponding Δσ y , and vary more from alloy to alloy. The data are analyzed to establish relations between Δσ y and coupled changes in the ultimate stresses as well as the effects of both test temperature and the unirradiated yield stress (σ yu ). The latter shows that higher σ yu correlates with lower Δσ y . In five out of six cases the effects of irradiation are generally consistent with previous observations on these alloys. However, the particular heat of the 12Cr HT-9 tempered martensitic steel in this study has a much higher e u than observed for earlier heats. The reasons for this improved behavior are not understood and may be microstructural in origin. However, it is noted that the new heat of HT-9, which was procured under modern quality assurance standards, has lower interstitial nitrogen than previous heats. As a result, notably lower interstitial solute contents correlate with improved ductility and homogenous deformation in broadly similar steels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Investigation of defect clusters in ion-irradiated Ni and NiCo using diffuse X-ray scattering and electron microscopy

The nature of defect clusters in Ni and Ni$_{50}$Co$_{50}$ (NiCo) irradiated at room temperature with 2–16 MeV Ni ions is studied using asymptotic diffuse X-ray scattering and transmission electron microscopy (TEM). Analysis of the scattering data provides separate size distributions for vacancy and interstitial type defect clusters, showing that both types of defect clusters have a smaller size and higher density in NiCo than in Ni. Diffuse scattering results show good quantitative agreement with TEM results for cluster sizes greater than 4 nm diameter, but find that the majority of vacancy clusters are under 2 nm in NiCo, which, if not detected, would lead to the conclusion that defect density was actually lower in the alloy. Interstitial dislocation loops and stacking fault tetrahedra are identified by TEM. Lastly comparison of diffuse scattering lineshapes to those calculated for dislocation loops and SFTs indicates that most of the vacancy clusters are SFTs.

36 MATERIALS SCIENCE↗
DOE OSTI2018

The defect chemistry of U O 2 ± x from atomistic simulations

Control of the defect chemistry in UO 2±x is important for manipulating nuclear fuel properties and fuel performance. For example, the uranium vacancy concentration is critical for fission gas release and sintering, while all oxygen and uranium defects are known to strongly influence thermal conductivity. Here the point defect concentrations in thermal equilibrium are predicted using defect energies from density functional theory (DFT) and vibrational entropies calculated using empirical potentials. Electrons and holes have been treated in a similar fashion to other charged defects allowing for structural relaxation around the localized electronic defects. Predictions are made for the defect concentrations and non-stoichiometry of UO 2±x as a function of oxygen partial pressure and temperature. If vibrational entropy is omitted, oxygen interstitials are predicted to be the dominant mechanism of excess oxygen accommodation over only a small temperature range (1265 K to 1350 K), in contrast to experimental observation. Conversely, if vibrational entropy is included oxygen interstitials dominate from 1165 K to 1680 K (Busker potential) or from 1275 K to 1630 K (CRG potential). Below these temperature ranges excess oxygen is predicted to be accommodated by uranium vacancies, while above them the system is hypo-stoichiometric with oxygen deficiency accommodated by oxygen vacancies. Our results are discussed in the context of oxygen clustering, formation of U 4 O 9 , and issues for fuel behavior. In particular, the variation of the uranium vacancy concentrations as a function of temperature and oxygen partial pressure will underpin future studies into fission gas diffusivity and broaden the understanding of UO 2±x sintering.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Distribution of defect clusters in the primary damage of ion irradiated 3C-SiC

We report a statistical analysis of sizes and compositions of clusters produced in cascades during irradiation of SiC. The results are obtained using molecular dynamics (MD) simulations of cascades caused by primary knock-on atoms (PKAs) with energies between 10 eV and 50 keV. The results are averaged over six crystallographic directions of the PKA and integrated over PKA energy spectrum derived from the Stopping and Range of Ions in Matter (SRIM) code. Specific results are presented for 1 MeV Kr ion as an example of an impinging particle. We find that distributions of cluster size n for both vacancies and interstitials obey a power law f = A/n S and these clusters are dominated by carbons defects. The fitted values of A and S are different than values reported for metals, which can be explained through different defect energetics and cascade morphology between the two classes of materials. In SiC, the average carbon ratio for interstitial clusters is 91.5%, which is higher than the ratio of C in vacancy clusters, which is 85.3%. Size and composition distribution of in-cascade clusters provide a critical input for long-term defect evolution models.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Characterization of defect clusters in ion-irradiated tungsten by X-Ray diffuse scattering

Defect clusters in ion-irradiated tungsten are characterized by X-ray diffuse scattering. Single crystal tungsten samples are irradiated here at 300 K by 0.5–5 MeV copper ions with total doses of 0.2, 0.6, 2, and 5 displacements per atom (DPA). Analysis of the diffuse scattering intensities allows the determination of the concentration of vacancy- and interstitial-type dislocation loops as a function of loop radius for each sample. More vacancies than interstitials are observed in each sample, and the radius of the vacancy loops increases with irradiation dose. For the 0.2DPA sample the average vacancy loop radius is 10.1 ± 3.0 Å with a defect density (counting all atoms in the loops) of ( 10.2 ± 1.5 ) × 10 - 4 per tungsten atom, increasing to an average loop radius of 16.1 ± 3.1 Å at a density of ( 8.5 ± 1.0 ) × 10 - 4 per tungsten atom for the 5DPA sample.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Study on impact of Cr and Mo on diffusion of H in 2.25Cr1Mo steel using first-principle calculations

Chrome-molybdenum steel (2.25Cr1Mo steel) is one of the main materials of a steam generator (SG) in high-temperature reactor-pebblebed modules (HTR-PM). It is essential to analyze the source term of tritium in this material, because the behavior of tritium in a SG is important for performing source term analysis in normal and accident conditions. In this article, the diffusion behavior of H atom in 2.25Cr1Mo steel was calculated to estimate the diffusivity of tritium using first-principles density functional theory. To develop and simplify the model of hydrogen diffusion in 2.25Cr1Mo steel, the impact of Cr and Mo on the diffusion of hydrogen in bcc-Fe were first calculated, all the possible diffusion paths were considered, and the minimum energy path was obtained. The diffusion activation energy and pre-exponential factor of the diffusion coefficient were obtained from Vienna Ab initio Simulation Package combined with the climbing image-nudged elastic band method. Here the results indicate that the minimum energy path for the impurity H atom is from one tetrahedral interstitial site to an adjacent tetrahedral interstitial site. The function of the diffusion coefficient of H in 2.25Cr1Mo steel with temperature T can be expressed as D = 1.486 X 10 -7 x ($\frac{-16.350kJ/mol}{RT}$) (M2/s). The diffusion coefficient of our calculation and some of the previous experiments have an excellent quantitative agreement, which indicates the reliability of our crystalline model and the practicability of the present theoretical approach. More importantly, the computational results in this work can be treated as a good screening method to collect reasonable experimental data, which will provide a good reference for tritium source term evaluation in the SG of the HTR-PM.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Effects of temperature and pH on the electrochemical behaviour of alloy 600 in simulated pressurized water reactor primary water

The effects of temperature and pH on the electrochemical properties of Alloy 600 were studied in oxygenated borate buffer solutions by the measurements of open circuit potential (OCP), anodic potentiostatic polarization, Mott-Schottky analysis (MSA), and electrochemical impedance spectroscopy (EIS). At each temperature and pH, the stabilized anodic passive current density was independent of the formation potential, and the slope of the M - S plot for the passive film formed at +0.3 V OCP (OCP + 0.3 V) is positive, indicating that the passive films formed on Alloy 600 have n-type semiconductor character. With increasing temperature (in each solution), the OCP and the impedance modulus decreased, while the passive current density and the defect density calculated from MSA increased, indicating a significant reduction in corrosion resistance. With the increase in pH value, the passive current density decreased, the absolute value of impedance increased slightly, revealing a slight increase in corrosion resistance. According to the optimization of Mixed Potential Model (MPM) on the experimentally measured EIS data for Alloy 600 at OCP and +0.3 V OCP , a series of fundamental parameters related to the micro-characteristic of passive film were extracted. In the barrier layer of the passive film, the concentration of cation interstitials is much higher than that of anion vacancies, indicating that the former was a dominant defect. With increasing temperature, the linear increase in the defect density and film thickness resulted in a linear increase of the anodic current density. However, with increasing pH, the decrease of defect density and the slight increase of film thickness caused a linear reduction in anodic current density. Elevated temperature has a more significant deterioration effect on the corrosion resistance of Alloy 600 than the increased pH value. Additionally, all the kinetics parameters describing the generation of cation interstitials and oxygen vacancies at the substrate/film were obtained.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Effects of point defects on oxidation of 3C–SiC

The influence of implantation-induced point defects (PDs) on SiC oxidation is investigated via molecular dynamics simulations. PDs generally increase the oxidation rate of crystalline grains. Particularly, accelerations caused by Si antisites and vacancies are comparable, and followed by Si interstitials, which are higher than those by C antisites and C interstitials. However, in the grain boundary (GB) region, defect contribution to oxidation is more complex, with C antisites decelerating oxidation. Here, the underlying reason is the formation of a C-rich region along the oxygen diffusion pathway that blocks the access of O to Si and thus reduces the oxidation rate, as compared to the oxidation along a GB without defects.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Tensile behavior of dual-phase titanium alloys under high-intensity proton beam exposure: radiation-induced omega phase transformation in Ti-6Al-4V

A high-intensity proton beam exposure with 181 MeV energy has been conducted at Brookhaven Linac Isotope Producer facility on various material specimens for accelerator targetry applications, including titanium alloys as a beam window material. The radiation damage level of the analyzed capsule was 0.25 dpa at the beam center region with an irradiation temperature around 120 oC. Tensile tests showed increased hardness and a large decrease in ductility for the dual a+ß-phase Ti-6Al-4V Grade-5 and Grade-23 extra low interstitial alloys, with the near a-phase Ti-3Al-2.5V Grade-9 alloy still exhibiting uniform elongation of a few % after irradiation. Transmission Electron Microscope analyses on Ti-6Al-4V indicated clear evidence of a high-density of defect clusters with size less than 2 nm in each a-phase grain. The ß-phase grains did not contain any visible defects such as loops or black dots, while the diffraction patterns clearly indicated ?-phase precipitation in an advanced formation stage. The radiation-induced ?-phase transformation in the ß-phase could lead to greater loss of ductility in Ti-6Al-4V alloys in comparison with Ti-3Al-2.5V alloy with less ß-phase.google scho

Tensile Test, titanium, acclerator, proton beam↗
DOE OSTI2022

The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Accurate Fe–He machine learning potential for studying He effects in BCC-Fe

Nanostructured Ferritic Alloys (NFAs) containing oxide nanoparticles are prospective advanced structural materials in future fusion reactors. Lacking fusion neutron sources with an adequate flux, modeling is critical to understand radiation damage in the NFAs, including helium effects. Machine learning interatomic potentials (MLPs) have been demonstrated to be capable of describing atomic interactions in chemically complex systems with comparable accuracy to density functional theory (DFT) but with much less computational costs. Hence, MLPs are chosen to study He bubble accumulation in chemically complex NFAs. As a preliminary step, we develop a Fe–He potential, based on ~10,000 atomic configurations. The developed MLP accurately predicts the bulk properties of BCC-Fe compared to DFT, including the lattice constant (2.832 Å), elastic constants (c 11 = 271 GPa, c 12 = 141, c 44 = 93), and phonon frequencies (maximum error < 3.5%). The MLP predicts He binding energies in He n V bubbles and He n clusters with a mean absolute error (MAE) of only ~70 meV, which is ~3 to ~5 times smaller than the MAE of binding energies estimated using empirical potentials. The MLP also accurately predicts the migration barrier of interstitial He (64 meV). Furthermore, the MLP correctly predicts the maximum number of He atoms (critical size) at which a He n V bubble (He n cluster) can kick-out a self-interstitial atom (SIA), forming a bigger He n V 2 bubble (He n V bubble). Subsequently, we use the MLP to perform molecular dynamics simulations to investigate the effect of temperature on the critical size of He n V bubbles and He n clusters. We find that the critical size decreases with increasing temperature — indicating that smaller bubbles and clusters can kick-out an SIA at higher temperatures, thereby escalating the He bubble growth. We report the current Fe–He MLP can be further developed to include all chemical interactions in NFAs.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Microstructural evolution of compositionally complex solid-solution alloys under in-situ dual-beam irradiation

Here, this work attempts to link the microstructural evolution of single-phase compositionally complex alloy (CCA) compositions under dual-beam irradiation to their Mn-content via the stacking-fault energy (SFE) and vacancy migration energies. Two alloys, Cr 18 Fe 27 Mn 27 Ni 28 and Cr 15 Fe 35 Mn 15 Ni 35 , along with less compositionally complex pure Ni and Fe 56 Ni 44 binary were irradiated at 500 and 600 °C under dual-beam 1 MeV Kr 2+ and 16 keV He + heavy-ions up to 7 displacements per atom (dpa) with a He/dpa ratio of 0.75 %/dpa using in situ transmission electron microscopy (TEM). Due to the bubble-stabilizing effect of implanted He, bubbles were observed in all irradiations, and populations of faulted interstitial loops were characterized in Cr 18 Fe 27 Mn 27 Ni 28 and Cr 15 Fe 35 Mn 15 Ni 35 . A reduction in swelling was observed in the two CCAs compared to pure Ni and Fe 56 Ni 44 . Although swelling increased from 500 to 600 °C in Fe 56 Ni 44 , Cr 15 Fe 35 Mn 15 Ni 35 and Cr 18 Fe 27 Mn 27 Ni 28 both swelled slightly more at 500 °C. This was attributed to the difference in vacancy mobility, stronger pinning effect of vacancies on He, and the sink strength of faulted dislocation loops. Faulted interstitial loops nucleated with a higher number density and dislocation line density in Cr 15 Fe 35 Mn 15 Ni 35 at both temperatures, and at 600 °C in both materials. The differences in faulted loop population and temperature effect on swelling are correlated to the Mn-content and the measured SFE (20.2 ± 6.7 mJ/m 2 for Cr 18 Fe 27 Mn 27 Ni 28 and 9.2 ± 3.4 mJ/m 2 for Cr 15 Fe 35 Mn 15 Ni 35 ).

36 MATERIALS SCIENCE↗
DOE OSTI2024

Cluster dynamics simulations of tritium and helium diffusion in lithium ceramics

Tritium (T) and He diffusion in LiAlO 2 and LiAl 5 O 8 phases influences the performance of tritium producing burnable absorber rods (TPBARs) by affecting the gas release, swelling and thermal conductivity of Li-bearing ceramic pellets. Frenkel pair defects and clusters created by irradiation can attract T and He interstitials and form clusters of the type He i x Li, He i x Al, He i x O, T i x Li, T i x Al, and T i x O, 1 ≤ x ≤ 4 in a Li, Al or O vacancy site (notation denotes x He or T atoms in a 1 Li, 1 Al or 1 O vacant site). The concentration and mobility of each of these clusters collectively contribute to the diffusion of the He and T gases in LiAlO 2 and LiAl 5 O 8 . Here, in this work, free energy cluster dynamics simulations implemented in the Centipede code, are used to obtain the concentration and diffusivities of these clusters which are then used to calculate the total diffusivity of T and He gases in LiAlO 2 and LiAl 5 O 8 . The results show that diffusivity of T is at least one order of magnitude higher in LiAlO 2 as compared to that in LiAl 5 O 8 whereas He diffusion is 2–13 orders of magnitude higher in LiAlO 2 as compared to that in LiAl 5 O 8 . There is a higher concentration of highly diffusive species (T interstitials and T i 03 Li for the case of tritium and He i 01 Li, He i 02 Li, and He i 03 Li for the case of He) in LiAlO 2 than in LiAl 5 O 8 which increase the total diffusion of T and He in LiAlO 2 .

36 MATERIALS SCIENCE↗
DOE OSTI2025

Effect of Mg and Ni impurities on tritium diffusion in lithium ceramics through cluster dynamics simulations

This study investigates the impact of Mg and Ni doping on tritium diffusion in LiAlO 2 and LiAl 5 O 8 ceramics, that are used in tritium-producing burnable absorber rods (TPBARs). Utilizing Centipede simulations across a broad temperature range (500 K to 1250 K), we explore the interplay between defect dynamics, cluster formation, and tritium mobility. In LiAlO 2 , Mg doping significantly enhances tritium diffusivity by increasing tritium interstitial concentrations and diffusion coefficients of key species, thereby doubling the overall tritium diffusivity. Ni doping, while shifting the dominant defect to Li vacancies, maintains high tritium mobility due to the low binding energy of Li vacancy-tritium complexes, which ensures effective tritium migration. In LiAl 5 O 8 , Mg and Ni doping results in a slight reduction in the diffusion coefficients of key species, yet the dramatic increase in tritium interstitial concentrations compensates, leading to a net small increase in tritium diffusivity. In conclusion, the findings highlight the critical role of defects in tritium transport and the effect of Mg and Ni defects on the performance of these ceramics in demanding nuclear environments.

Cluster dynamics↗