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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Phase transformations, microstructural refinement and defect evolution mechanisms in Al-Si alloys under non-hydrostatic diamond anvil cell compression

Non-hydrostatic compression of materials using a diamond anvil cell (DAC) can transform equilibrium microstructure of an alloy to novel, potentially metastable states. In this study, in situ synchrotron X-ray diffraction (XRD) during compression up to 24 GPa and subsequent ex situ, high resolution analytical electron microscopy (AEM) after decompression of an Al-Si alloy provided insight into the crystallographic changes during the compression as well as microstructural refinement, and defect structures caused by such a high pressure compression-decompression process. Pressure resolved in-situ synchrotron XRD was used to detail the phase transformation pathway of the eutectic Si phase in Al-Si alloy, from Si-I → Si-XI → Si-V during compression, and a final transformation predominantly to Si-III after decompression. Using scanning and transmission electron microscopy (S/TEM), site specific analysis of the alloy immediately underneath the anvil contact surface demonstrated a highly complex microstructure. A narrow region of thick amorphous Al oxide interspersed with nanocrystalline grains was found at the top surface. Underneath this Al oxide, while the majority the eutectic Si was transformed into highly-deformed, polycrystalline (PC) Si-III, a complex intermediate layer was discovered at the interface between Al and Si, comprised of a small fraction of Al nanocrystals and a majority of nanocrystalline Si-I. This combination of pressure resolved in-situ synchrotron XRD coupled with subsequent high resolution, electron microscopy resolved the phase transformation as well as non-equilibrium microstructures in a metallic alloy induced by a non-hydrostatic high pressure compression followed by decompression.

36 MATERIALS SCIENCE↗
DOE OSTI2020

In-situ observation of nano-oxide and defect evolution in 14YWT alloys

Nanostructured ferritic alloys (NFAs) are considered as candidates for structural components in advanced nuclear reactors due to their excellent radiation resistance as a result of a high density of nano-oxides (NOs) in the microstructure. Therefore, gaining an understanding on the stability of NOs under irradiation is crucial. In this study, we have investigated the evolution of defects and NOs in 14YWT NFAs under in-situ Kr ion irradiation at room temperature (RT) and 450°C up to 10 dpa. It has been found that irradiations at 450°C do not create any changes in the NOs, similar to the bulk irradiations. On the other hand, elemental mapping indicates that NOs dissolve mostly after 10 dpa irradiations at RT. Thus, while defects are both annihilated and pinned by NOs at low doses (before the dissolution of NOs), glissile loops start to escape to the foil surface at high doses (after the dissolution of NOs), justifying the significantly low fraction of (111) loops compared to the literature values. Furthermore, high resolution transmission electron microscopy analysis has shown that the NOs are mostly coherent Y 2 Ti 2 O 7 particles with pyrochlore crystal structure after both RT and 450°C irradiations, similar to those observed before irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Point defect evolution under irradiation: Finite size effects and spatio-temporal correlations

The evolution of point defect concentrations under irradiation is controlled by their diffusion properties, and by their formation and elimination mechanisms. The latter includes the mutual recombination of vacancies and interstitials, and the elimination of point defects at sinks. Two models are traditionally used to predict the defect concentration evolution, the standard rate theory (SRT) and the atomistic kinetic Monte Carlo (AKMC). We show in this work a large discrepancy in the defect concentrations between both methods when the average number of defects in the AKMC simulation box is close or lower than one. The reason is that AKMC naturally captures strong space and time correlations between vacancies and interstitials generated by the finite size of the periodic simulation box. These correlations strongly affect the recombination rate and the point defect concentrations. SRT fails to predict such correlations, and the corresponding solution deviates from the more accurate solution given by the AKMC simulation under similar conditions. These finite size correlation effects are strong when the elimination of point defects occurs by recombination only, but can still be significant in the presence of sinks. In order to account for the spacio-temporal correlation in a continuum framework, we introduce a Correlated Pair Theory (CPT). This theory fully takes into account the correlations between vacancy and interstitial pairs and predicts point defect concentrations in good agreement with AKMC simulations. Inversely, if the goal is for the AKMC to reproduce defect concentrations in bulk using small simulation boxes, the naturally occurring correlations need to be corrected. We show here that the CPT can be used to modify the elimination rates in the AKMC simulations, so as to yield point defect concentrations in agreement with SRT.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Irradiation-Induced Defect Evolution in Nuclear Graphite

Graphite has historically been used as a moderator material in nuclear reactor designs dating back to the first man-made nuclear reactor to achieve criticality (Chicago Pile 1) in 1942. Additionally, graphite is a candidate material for use in the future envisioned next-generation nuclear reactors (Gen IV); specifically, the molten-salt-cooled (MSR) and very-high-temperature reactor (VHTR) concepts. Gen IV reactor concepts will introduce material challenges as temperature regimes and reactor lifetimes are anticipated to far exceed those of earlier reactors. Irradiation-induced defect evolution is a fundamental response in nuclear graphite subjected to irradiation. These defects directly influence the many property changes of nuclear graphite subjected to displacing radiation; however, a comprehensive explanation for irradiation-induced dimensional change remains elusive. The objectives of this project were focused on the characterization of irradiation-induced defect evolution in nuclear graphite via transmission electron microscopy (TEM). With the use of novel TEM specimen preparation techniques, high-temperature electron-irradiation and characterization of high-temperature neutron-irradiated nuclear graphite, novel fullerene-like defects are shown to be a dominant defect type, especially at higher temperatures. These results contradict the historical models of defect evolution and provide valuable insight into the macroscopically observed property changes in irradiated nuclear graphite.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Development of a MOOSE-based crystal plasticity model with irradiation defect evolution for irradiation creep in 316

Irradiation creep and irradiation swelling are two of the lifetime limiting factors for structural materials in nuclear reactors. These mechanical effects are driven by irradiation defect evolution and the interaction of those defects with dislocations in the microstructure. We present here a coupled cluster dynamics – crystal plasticity approach to model irradiation swelling and creep behavior in 316 SS. The time-dependent evolution of irradiation defects is calculated with a cluster dynamics approach and passed to the crystal plasticity model to compute the dislocation evolution. We show the impact of the irradiation defect evolution on the stress state in the material, which drives inelastic deformation through dislocation motion. The inelastic deformation in the 316 SS is dependent on the dose rate, where the inelastic deformation driven by the early-stage irradiation defect evolution determines the mechanical behavior of the 316 SS.

316 Stainless Steels↗
DOE OSTI2026

Microstructural evolution, defect mitigation, and precipitation behavior in AA6061 via laser powder bed fusion with high-temperature substrate heating

Defects, particularly solidification cracking, remain persistent challenges in the laser powder bed fusion (LPBF) processing of AA6061 aluminum alloy. This study systematically investigates defect mitigation, microstructural evolution, and mechanical properties associated with high-temperature substrate preheating at 500 °C. Comprehensive microstructural analyses, including characterization of defects, grain structures, and precipitation behavior, were performed on samples in both as-built and T6 heat-treated states. Elevated preheating substantially reduced solidification cracking across a wide processing window, while demonstrating decreased crack sensitivity to laser parameters. Columnar cracks along the build direction were observed despite substrate preheating. Lack-of-fusion and keyhole porosity were effectively eliminated, though gas-induced microporosity persisted at higher powers. In-depth characterization of two distinct laser power and speed conditions confirmed the formation of micron-sized, non-coherent Mg 2 Si precipitates under heated substrate conditions, alongside α-AlFeCrMnSi intermetallic phases indicative of in-situ thermal effects during fabrication. Subsequent T6 heat treatment revealed the formation of fine, coherent needle-shaped β″ strengthening precipitates. Despite substantial differences in processing parameters, comparable mechanical properties were measured in the as-built samples (∼52 MPa yield strength, ∼130 MPa tensile strength), primarily due to reduced strain hardening effects and consistent precipitation characteristics. Meanwhile, the T6 heat treatment led to significant improvement in properties, enhancing yield strength by over 400%, aligning closely with the performance of conventional wrought AA6061-T6. These findings underscore that high-temperature substrate preheating offers an effective means to suppress cracks, control precipitation, and enhance mechanical performance in LPBF-processed AA6061.

36 MATERIALS SCIENCE↗
DOE OSTI2025

The effect of heat treatment on the defect evolution in LPBF 316H stainless steel

This work investigates the effect of processing and heat treatment on defect evolution and irradiation response of laser powder bed fusion (LPBF) 316H stainless steel, with comparisons to LPBF 316L and wrought 316L/316H. Using in-situ and ex-situ ion irradiations across a wide parameter space—temperature (300–675 °C), dose (0.2–25 dpa), dose rate (10 -3 –10 -5 dpa/s), and helium co-implantation (20–2500 appm)—we correlated void swelling, dislocation loop evolution, and segregation behavior with pre-irradiation microstructures (as-built, stress-relieved, solution-annealed, and cold-worked). Results show that swelling is strongly controlled by dislocation density: intermediate densities maximize swelling, while solution annealing or cold working reduce susceptibility to levels comparable to wrought alloys. Low dose rates and helium both promote cavity nucleation and lower the incubation barrier, with helium suppressing the role of dislocation density and driving swelling behavior toward wrought-like response. Loop evolution in SA LPBF 316H resembles wrought 316L but shows localized denuded zones near low-angle grain boundaries. STEM-EDS mapping further revealed Ni segregation at voids and sparse Al-rich oxides, without evidence of Ni–Si precipitates. Collectively, these findings identify dislocation density, helium content, and dose rate as the factors governing swelling in LPBF 316H and provide mechanistic datasets for model validation, supporting LAIN and the qualification of AM austenitic steels for nuclear service.

316 stainless steel↗
DOE OSTI2023

Effect of local chemical order on the irradiation-induced defect evolution in CrCoNi medium-entropy alloy

High- (and medium-) entropy alloys have emerged as potentially suitable structural materials for nuclear applications, particularly as they appear to show promising irradiation resistance. Recent studies have provided evidence of the presence of local chemical order (LCO) as a salient feature of these complex concentrated solid-solution alloys. However, the influence of such LCO on their irradiation response has remained uncertain thus far. Here, in this work, we combine ion irradiation experiments with large-scale atomistic simulations to reveal that the presence of chemical short-range order, developed as an early stage of LCO, slows down the formation and evolution of point defects in the equiatomic medium-entropy alloy CrCoNi during irradiation. In particular, the irradiation-induced vacancies and interstitials exhibit a smaller difference in their mobility, arising from a stronger effect of LCO in localizing interstitial diffusion. This effect promotes their recombination as the LCO serves to tune the migration energy barriers of these point defects, thereby delaying the initiation of damage. These findings imply that local chemical ordering may provide a variable in the design space to enhance the resistance of multi-principal element alloys to irradiation damage.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Defect evolution in Ni and solid-solution alloys of NiFe and NiFeCoCr under ion irradiation at 16 and 300 K

Single-phase concentrated solid-solution alloys (SP-CSAs) have shown unique chemical complexity at the levels of electrons and atoms, and their defect evolution is expected to be different from conventional dilute alloys. Single crystals of Ni, NiFe and NiFeCoCr are chosen as model systems to understand the chemical complexity on defect formation and damage accumulation in SP-CSAs under ion irradiation. The high-quality crystals were irradiated at 16 and 300 K to different ion fluences, to form irradiated region with little to heavy damages. Additionally, the ion-induced damage was determined using Rutherford backscattering spectrometry technique along a channeling direction (RBS/C) and the level of lattice damage in irradiated Ni and SP-CSAs was quantified from Monte Carlo (MC) simulations. The results are interpreted using the Multi Step Damage Accumulation model to reveal material damage accumulation kinetics. Key findings of the study are that in case of room temperature irradiations the damage level measured for complex alloys at the highest irradiation fluence of 2 × 10 15 cm -2 (~3 dpa) is significantly higher than that obtained for pure nickel samples and suggest two-step damage accumulation process with a defect transformation taking place at a fluence of about 1.5 × 10 15 cm -2 . Furthermore, structural and damage kinetic differences clearly imply that, with increasing degree of chemical complexity and high solid-solution strengthening effects from Ni to NiFe and to NiFeCoCr, the enhanced lattice stiffness resists to randomization of atomic configurations and inhibits the growth of extended defects.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Formation of I 1 stacking fault by deformation defect evolution from grain boundaries in Mg

I 1 stacking faults (SFs) in Mg alloys are regarded as the nucleation sites of $\langle \text{c+a} \rangle$ dislocations that are critical for these alloys to achieve high ductility. Previously it was proposed that the formation of I 1 SFs requires the accumulations of a large number of vacancies, which are difficult to achieve at low temperatures. In this study, molecular dynamics (MD) and molecular statics (MS) simulations based on empirical interatomic potentials were applied to investigate the deformation defect evolutions from the symmetric tilt grain boundaries (GBs) in Mg and Mg-Y alloys under external loading along $\langle c \rangle$-axis. The results show the planar faults (PFs) on Pyramidal I planes first appear due to the nucleation and glide of $\langle {\frac {1}{2}}c + p\rangle$ partial dislocations from GBs, where $\langle p\rangle$ = ${\frac {1}{3}} \langle10\bar{1}0\rangle$. These partial dislocations with pyramidal PFs interact with other defects, including pyramidal PFs themselves, GBs, and $\langle p\rangle$ partial dislocations, generating a large amount of I 1 SFs. Detailed analyses show the nucleation and growth of I 1 SFs are achieved by atomic shuffle events and deformation defect reactions without the requirements of vacancy diffusion. Our simulations also suggest the Y clusters at GBs can reduce the critical stress for the formation of pyramidal PFs and I 1 SFs, which provide a possible reason for the experimental observations that Y promotes the $\langle{\text{c+a}}\rangle$ dislocation activities.

36 MATERIALS SCIENCE↗
DOE OSTI2024

A Finite Difference informed Random Walk solver for simulating radiation defect evolution in polycrystalline structures with strongly inhomogeneous diffusivity

Diffusivity of species and defects on grain boundaries is usually several orders of magnitude larger than that inside grains. Such strongly inhomogeneous diffusivity requires prohibitively high computational demands for modeling microstructural evolution. Here, this paper presents a highly-efficient numerical solver, combining the Finite Difference method and Random Walk model, designed for accurately modeling strongly inhomogeneous diffusion within polycrystalline structures. The proposed solver, termed Finite Difference informed Random Walk (FDiRW), integrates a customized Finite Difference (cFD) scheme tailored for fast diffusion along thin grain boundaries represented by a single-layer of nodes. Numerical experiments demonstrate that the FDiRW solver achieves an impressive efficiency gain of 1560x compared to traditional Finite Difference methods while maintaining accuracy, making it feasible for personal computer machines to handle diffusional systems with strongly inhomogeneous diffusivity across static polycrystalline microstructures. The model has been successfully applied to simulate radiation defect evolution, showcasing its scalability to engineering scales in both length and time dimensions.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Generalized kinetic model for defect evolution in irradiated materials

A fundamental and open problem in materials science is to determine how the structural properties of irradiated and self-irradiated materials evolve in time. Structural defects generated by irradiation lead to changes in a material’s macroscopic properties, and these macroscopic changes, such as volumetric swelling, can in turn give rise significant alterations of the functionality of a material. Here, we develop a unified analytical model to understand the complex interplay between aggregation, fragmentation, and recombination processes involving mobile clusters of defects such as bubbles, voids, and interstitial atoms in materials subject to radiative sources. Specifically, we employ a mean field approach to derive a system of coupled kinetic equations that describes both the time evolution of the density of each type of defect and how clusters of each defect form and grow as the material ages.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Ta–Zr carbides: Synthesis advances via carbothermal reduction and defect evolution observed through transmission electron microscopy ion irradiation

The thermodynamic stability of six distinct compositions within the Zr—Ta—C ternary system is investigated in this study, marking the first report of their synthesis through carbothermic reduction in vacuum. A prolonged annealing process at 2200°C enabled high densification and phase equilibrium. Detailed phase identification and microstructural characterization through microscopy and X-ray diffraction techniques revealed clear compositional trends and stable phase formations. Two compositions ((Ta 0.2 Zr 0.8 )C 0.6 and (Ta 0.5 Zr 0.5 )C 1 ) were selected for ion irradiation experiments using 200 keV Kr + at 600°C—representing the first-ever irradiation study on the Zr—Ta—C system. The findings indicated defect accumulation and nanoscale cavity formation without any evidence of amorphization, highlighting the system's structural stability under irradiation. Together, the synthesis and irradiation results provide a basis for further investigation of the system and suggest its relevance for applications under extreme environments.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Fullerene-like defects in high-temperature neutron-irradiated nuclear graphite

Irradiation-induced defect evolution in graphite is particularly important for its application in graphite-moderated nuclear reactors. The evolution of defects directly influences macroscopically observed property changes in irradiated nuclear graphite which, in turn, can govern the lifetime of graphite components. This article reports novel defect structures and the irradiation response of microstructural features occurring in high-temperature irradiated nuclear graphite IG-110. High resolution transmission electron microscopy (HRTEM) was used to characterize specimens neutron-irradiated at a high temperature (≥ 800°C) at doses of 1.73 and 3.56 atomic displacements per atom (dpa). Concentric shelled and fullerene-like defects were found to result in swelling along the c-axis and contraction along the a/b-axis of crystallites. Furthermore, such defects are shown to occur within, and partially fill, Mrozowski cracks prior to turnaround dose. In addition, in situ TEM under similar irradiation conditions was used to capture the real-time dynamic evolution of defects, providing unambiguous analysis of the evolution of the graphite structures during irradiation. Here, the results suggest the mainstream theory for radiation damage in nuclear graphite (which assumes additional basal plane formation as the sole reason) to be an incorrect interpretation of defect evolution contributing to irradiation-induced property changes at higher temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Role of α → ε → α phase transformation on the spall behavior of iron at atomic scales

Shock compression of iron microstructures above a threshold stress results in a α (BCC) → ε (HCP) transformation and the propagation of the release wave brings the metal back to the α phase following the ε→α transformation. Predicting failure behavior under shock loading conditions (spallation) relies on understanding the evolution of defects in the microstructure as it undergoes the α→ε→α phase transformation. This study uses molecular dynamics (MD) simulations to investigate the role of defect evolution during α→ε→α phase transformation on the spall strength values of single-crystal (sc) Fe microstructures. The MD simulations aim to characterize the ε phase fraction formed during shock compression and the defects during shock release for variations in loading orientations and shock stresses. The simulations are carried out for loading along the [100], [110], [111], and [112] orientations and for impact velocities ranging from 600 m/s to 1 km/s. The ε phase fractions during compression and defects (dislocations, twins) characterized during spall failure show an orientation dependence that affects the spall strength values. Here, the lowest value for spall strength is observed for the <110> loading orientation that shows a high density of twinning at the spall plane, whereas the highest value is observed for the <100> orientation and is associated with a α (BCC) → γ (FCC) transformation at the spall plane. The correlations of the spall strength values with the strain rates and with the ε phase fractions are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Evolution of extended defects in UO 2 during high temperature annealing

The evolution of thermophysical characteristics of uranium oxide (UO 2 ) under irradiation is a critical aspect of nuclear fuel performance of light water reactors. This study examines the effect of high-temperature annealing at 1000 °C - 1600 °C on defect evolution in krypton irradiated UO 2 at room temperature. Transmission electron microscopy analysis reveals that the size of irradiation-induced bubbles and dislocation loops increases but their density decreases with annealing temperatures. Perfect dislocation loops with = <110> dominate, regardless of the annealing temperature. A rate theory model for extended defect evolution is applied to conclude that current understanding of dynamical loop and bubble evolution at high-temperature annealing is governed by defect coalescence.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗