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At least 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Comparison of Ion and Neutron Irradiations to 3 dpa at 500C in Ferritic-Martensitic Alloys

The growing global demand for energy will increasingly call upon advanced nuclear fission reactors to supply safe and reliable electricity. The structural and fuel cladding components of these reactors will be subject to extreme conditions of irradiation damage up to several hundred displacements per atom (dpa) at temperatures as high as 700°C. Ferritic-martensitic (F-M) steels are leading candidates for these challenging conditions due to their strength and dimensional stability under irradiation. In order to accelerate the process for evaluating F-M alloys, charged particles are increasingly being used to emulate neutron irradiations. Charged particle irradiations allow the possibility of conducting irradiation experiments within a shorter time period (i.e. at a rate up to 4 orders of magnitude faster) and with minimal radioactivation of the material, enabling lower cost and faster turnaround of post irradiation examination and analysis. However, the irradiation dose rate, damage cascade morphologies, and irradiation damage depth profiles all differ widely between protons, heavier ions, and neutrons. Currently, there is limited understanding of the significance of these physical differences and how they manifest in the irradiated microstructure and mechanical properties of F-M steels. The objective of this study is to evaluate charged particles as a surrogate for neutron irradiations in F-M alloys by assessing common irradiation conditions using Fe++ ions, protons, and neutrons. Keeping the temperature and dose consistent enables isolation of the effects of each irradiating particle and their respective dose rates and cascade morphologies.

Swenson, M.J.↗
DOE OSTI2025

Time at Temperature Abstract and Intern Poster

Current NRC regulations for BWR operations dictate that any occurrence reaching the point of Departure from Nucleate Boiling (DNB) disqualifies the use of the fuel rod for further reactor operation. That criteria does not account for duration of rate of the power increase, the corresponding effects on material properties or rewetting that may occur. Previous Anticipated Operational Occurrences (AOOs) show short durations power increase that may require limited amounts of heat removal. Industrial experience has shown evidence that fuel can reach dryout yet continue to safely operate in regular reactor conditions. The gap in research into such occurrences is the motivation for a series of experiments, including this current work. Time-at-Temperature experiments aim to identify and characterize the microstructural changes in Zircaloy-2 under oxygen-free high temperatures. This work uses the FlashDSC instrument to rapidly ramp up and down the temperature of a focus ion beam (FIB) prepared large area lift-out (LALO) of Zircaloy-2 at a rate of 10,000 K/s to desired values. The focus is to characterize the microstructure evolution, if any, of Zircaloy-2 that may impact its performance under typical BWR conditions. This characterization includes analyzing grain structure and size, secondary phase particle (SPPs) size, shape, composition, and location using Transmission Electron Microscopy (TEM). Further data collection and analysis is in progress including diffraction pattern indexing and 4D STEM processing. While current results focus on the testing and characterizing unirradiated material, future plans include expansion to irradiated material to explore the effects of rapid transition rates seen in DNB and dryout conditions on irradiationg damage and defect annealing.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Shining light on nanoscale ‘vine-on-stick’ eutectic structures in the Al-Ce-Ni system

Multiphase eutectics often comprise entangled solid phases with nanoscale periodicity, making it difficult to unravel their 3D connectivity using conventional 2D techniques. Among such systems, the three-phase eutectic Al-Al 11 Ce 3 -Al 3 Ni stands out for its ultrafine (∼100 nm) interphase spacing and promising creep resistance, yet its microstructure remains relatively unexplored despite its potential for high-temperature applications. Here, we use scanning hard X-ray microscopy with an unprecedented ∼10 nm pixel size to resolve its 3D morphology. Reconstructions reveal a novel “vine-on-stick” motif, wherein Al 11 Ce 3 wraps around Al 3 Ni pillars. Analysis of phase tortuosities confirms that Al 11 Ce 3 exhibits more convoluted morphologies than Al 3 Ni. No orientation relationship was observed between the two intermetallics, suggesting that growth is controlled by local solute gradients rather than epitaxy. Fragmentation of intermetallics along their longitudinal axis suggests a Rayleigh-type breakup mechanism in solid state. The “vine-on-stick” pattern may generalize to other alloy systems with variable interfacial anisotropies and low mutual solubilities.

36 MATERIALS SCIENCE↗
DOE OSTI2021

A predictive discrete-continuum multiscale model of plasticity with quantified uncertainty

Multiscale models of materials, consisting of upscaling discrete simulations to continuum models, are unique in their capability to simulate complex materials behavior. The fundamental limitation in multiscale models is the presence of uncertainty in the computational predictions delivered by them. In this work, a sequential multiscale model has been developed, incorporating discrete dislocation dynamics (DDD) simulations and a strain gradient plasticity (SGP) model to predict the size effect in plastic deformations of metallic micro-pillars. The DDD simulations include uniaxial compression of micro-pillars with different sizes and over a wide range of initial dislocation densities and spatial distributions of dislocations. An SGP model is employed at the continuum level that accounts for the size-dependency of flow stress and hardening rate. Sequences of uncertainty analyses have been performed to assess the predictive capability of the multiscale model. The variance-based global sensitivity analysis determines the effect of parameter uncertainty on the SGP model prediction. The multiscale model is then constructed by calibrating the continuum model using the data furnished by the DDD simulations. A Bayesian calibration method is implemented to quantify the uncertainty due to microstructural randomness in discrete dislocation simulations (density and spatial distribution of dislocations) on the macroscopic continuum model prediction (size effect in plastic deformation). Here, the outcomes of this study indicate that the discrete-continuum multiscale model can accurately simulate the plastic deformation of micro-pillars, despite the significant uncertainty in the DDD results. Additionally, depending on the macroscopic features represented by the DDD simulations, the SGP model can reliably predict the size effect in plasticity responses of the micropillars with below 10% of error.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Development of hydrothermal corrosion model and BWR metal coating for CVD SiC in light water reactors

SiC/SiC fiber composites with CVD SiC overcoats are potential candidates for light water reactor advanced accident tolerant cladding materials. Understanding its corrosion kinetics in Light Water Reactor (LWR) conditions is essential to evaluate the concept's viability. Existing models only account for the temperature and oxygen concentration effect on the hydrothermal corrosion behavior applicable to LWR operating conditions. However, the development of a general corrosion rate for CVD SiC that accounts for the impact of irradiated microstructure, flow rate, electrical resistivity, pH, and surface roughness is critical for the practical realization of SiC/SiC-based cladding concepts. After a rigorous experimental campaign, this work updates the existing hydrothermal corrosion model to predict hydrothermal corrosion in LWRs. Numerical radiation and coolant chemistry analysis for LWRs conducted based on the updated corrosion kinetic models suggests that CVD SiC is likely a viable environmental barrier coating for Pressurized Water Reactors while questionable for Boiling Water Reactors (BWR) when the effect of irradiation damage on SiC corrosion is considered. An effective mitigation strategy for the double-layer metal coating is proposed for BWR applications. The double-layer metal coating comprising a FeCrAl overcoat with an intermediate Cr bond coating was observed to provide a stable protective barrier against SiC dissolution in BWR conditions. Finally, the proposed metal coating was also fully adherent following quench and burst tests.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Crack mitigation in additively manufactured AlCrFe 2 Ni 2 high-entropy alloys through engineering phase transformation pathway

The far-from-equilibrium solidification during additive manufacturing often creates large residual stresses that induce solid-state cracking. Here we present a strategy to suppress solid-state cracking in an additively manufactured AlCrFe 2 Ni 2 high-entropy alloy via engineering phase transformation pathway. We investigate the solidification microstructures formed during laser powder-bed fusion and directed energy deposition, encompassing a broad range of cooling rates. At high cooling rates (10 4 –10 6 K/s), we observe a single-phase BCC/B2 microstructure that is susceptible to solid-state cracking. At low cooling rates (10 2 –10 4 K/s), FCC phase precipitates out from the BCC/B2 matrix, resulting in enhanced ductility (~10 %) and resistance to solid-state cracking. Site-specific residual stress/strain analysis reveals that the ductile FCC phase can largely accommodate residual stresses, a feature which helps relieve residual strains within the BCC/B2 phase to prevent cracking. Our work underscores the value of exploiting the toolbox of phase transformation pathway engineering for material design during additive manufacturing.

36 MATERIALS SCIENCE↗
DOE OSTI2020

First Examination of Irradiated Fuel with Pulsed Neutrons at LANSCE (Preliminary Results)

We present preliminary results on the characterization of an irradiated U-lOZr-lPd fuel sample that was prepared from the irradiated AFC-3A-R5A sample. U-lOZr metallic fuels are researched as host materials for potential transmutation fuels and the addition of palladium strives to bind lanthanides, thus preventing fuel-cladding chemical interactions (FCCI). These interactions limit the lifetime of metallic fuels and are caused by migration of lanthanide fission products to the periphery of the fuel slug, where they start to interact with the D9 or HT9 steel cladding, ultimately leading to failure of the mechanical integrity of the cladding. Neutrons offer bulk characterization of irradiated materials for which X-ray tomography methods are not suitable due to the immense gamma background emitted from the samples. In particular pulsed neutrons provide information from the ability to resolve the neutron energy using their time-of-flight and thus the potential to utilize neutron absorption resonance to characterize the spatial distribution of isotopes. This, in turn, may allow to characterize the distribution of fission and neutron capture products non-destructively and may ultimately be applied to the bulk of an irradiation capsule prior to destructive post-irradiation examination to identify regions of interest. To allow the characterization of entire irradiation capsules, a cask is under development in the advanced post-irradiation work package at LANL and progress on this development was reported elsewhere. In parallel, an irradiated U-lOZr-lPd sample cut from the AFC-3AR5A irradiation was shipped to LANL and will be fully characterized with an NSUF funded rapid turnaround experiment (RTE) in the 2020 LANSCE run cycle. The sample emits at a dose rate of ~3R/hr on contact and is therefore manageable with remote handling, without requiring a cask. The disk-shaped material is larger than samples prepared for analysis using electron or X-ray methods and is therefore an intermediate step towards characterization of bulk samples at LANSCE. However, since it covers the full diameter of the irradiated fuel slug, some insight on redistribution of elements, spatially resolved information on microstructure, e.g. phase composition and texture, will be possible using the pulsed neutron-based methods developed for fuel characterization at LANSCE. This report describes the development of procedures to handle the sample at LANSCE as well as preliminary data and results from tests conducted in December 2019 on the energy-resolved neutron imaging (ERNI) beam line at flight path 5 and the high pressure-preferred orientation diffractometer (HIPPO) at LANSCE. This effort is a collaboration between LANL, INL, and ORNL. To compare our capabilities with prior work, we present an overview of previously reported bulk characterization of irradiated or spent fuels. The overview addresses neutron diffraction and neutron absorption resonance spectroscopy, both of which have only few reported applications on irradiated or spent nuclear fuel, as well as neutron radiography. This literature review was already described in a previous report but is repeated here to put our current efforts in context of previous work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

PNNL's Characterization Summary for MP-2 Experiment

Characterization of as-fabricated fuel was performed at Pacific Northwest National Laboratory (PNNL) in accordance with the characterization plan for the fabrication of U 10Mo plate fuel for the U.S. High Performance Research Reactor conversion program’s Fuel Fabrication Pillar (INL 2021). Similar characterization work is also being performed at Idaho National Laboratory to provide a detailed understanding of the as-fabricated foils that would be irradiated in the Mini-Plate 2 (MP 2) experiment. Under the MP 2 characterization plan, foils are studied that have different fabrication parameters (such as rolling condition, rolling thickness reduction, co-rolling with Zr layers). Similar samples from master foils were sent to both the organizations, so that the testing and analysis can be done independently using similar equipment and standardized measurement and analysis procedures. A final, consolidated report will be prepared based on this work and will summarize all the information obtained from the two laboratories. The MP 2 experiment will provide an opportunity to understand the effects of processing conditions on the final fuel microstructure, to compare results obtained independently, and achieve a two-way validation. In Fiscal Year 2022, PNNL received five MP 2 cast (PD STD2) samples to examine the foils’ chemistry and microstructure. For each cast sample, PNNL received samples from three different locations. PNNL also received and characterized 24 U 10Mo foil samples, by sectioning four pieces/specimens from each foil, in accordance with the MP 2 Characterization Plan (INL 2021). These 24 samples consist of four types of foils from BWX Technologies: 0.047 in. thick hot-rolled and annealed samples with Zr layers; 0.025 in. thick cold-rolled and annealed samples with Zr layers; 0.0105 in. thick cold-rolled and annealed samples with Zr layers. Along with these, PNNL also received four plates with Zr layers that were 0.025 in. and 0.0105 in. thick. This report describes the results of PNNL’s MP 2 foil characterization. Microstructure, Mo homogeneity, carbide fraction and morphology, U 10Mo foil thickness, and Zr thickness were evaluated in both the longitudinal and transverse directions for all the foils of the three different thicknesses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Residual Stress Measurements of Cr-coated Cladding

Accident-tolerant fuel concepts for light-water reactor applications have been developed and tested in diverse research programs around the world. Industry teams have developed coated cladding concepts using different coating application methods, such as physical vapor deposition and coldspraying. Recently, variations in thickness and application method were investigated, and their effects on the microstructure and mechanical properties were reported. The results suggest an increase in performance in an unirradiated state, however, it remains unclear what phenomenon drove this performance. One explanation to this performance is associated with the coating process imparting a residual stress. Residual stress analysis was performed on coatings produced with various process parameters to understand how a thin layer of Cr impacts the mechanical properties after coating application and after simulated operating conditions. The gathered information is intended to address safety concerns or support margin identification related to accident-tolerant fuels and to support subsequent modeling and simulation efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Visualizing time-dependent microstructural and chemical evolution during molten salt corrosion of Ni-20Cr model alloy using correlative quasi in situ TEM and in situ synchrotron X-ray nano-tomography

In situ monitoring of corrosion processes is important to fundamentally understand the kinetics and evolution of materials in harsh environments. A quasi in situ transmission electron microscopy technique was utilized to study microstructural and chemical evolution of a Ni-20Cr disc sample 6 exposed to molten KCl-MgCl 2 salt for 60 s in consecutive 20 s iterations. In situ synchrotron X-ray nano-tomography was performed to characterize the morphological evolution of a Ni-20Cr microwire exposed to molten KCl-MgCl 2 . Here, both techniques captured key corrosion events and revealed mechanisms at different time and length scales, potentially bringing greater insights and 10 deeper understanding beyond conventional analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Measuring thermomechanical response of large-format printed polymer composite structures via digital image correlation

Large-format additive manufacturing (LFAM) is a branch of additive manufacturing (AM) research with the ability to create large structures typically measuring several meters in scale. LFAM is advantageous for tooling applications, not only because it offers the ability to create complex geometries not easily made using subtractive manufacturing processes, but the cost savings of pelletized feedstock used by these systems result in larger parts printed at faster speeds than traditional AM systems. Fiber reinforced polymer (FRP) is a commonly used feedstock material in LFAM structures because it reduces the distortion experienced during printing. However, FRP introduces highly anisotropic thermomechanical properties and contributes to a nonhomogeneous microstructure that can result in critical distortion of dimensions during tooling. Measuring the global thermomechanical response of LFAM structures requires a more representative method that accounts for not only anisotropic properties but also the nonhomogeneous nature of the final part. This is where traditional techniques to measure thermomechanical response, such as thermomechanical analysis (TMA), fall short as they assume homogeneity. This study evaluated the coefficient of thermal expansion (CTE) of LFAM structures as measured by TMA as compared to a novel digital image correlation oven (DIC Oven) system. The LFAM structures were made from 20 % by weight carbon fiber reinforced acrylonitrile butadiene styrene (CF-ABS). TMA measurements showed significant variations in CTE across a single LFAM bead, confirming the need for a global technique that captures overall thermomechanical response. The CTE values measured using the DIC Oven compared well to average TMA values obtained from localized measurements across the sample. The DIC Oven was also used to quantify the effects of different layer orientations on thermomechanical properties, which cannot be easily captured using TMA. A predictive model was also developed by using localized TMA values across an LFAM bead to predict the overall thermomechanical response of an LFAM structure.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Variation of effective filament diameter, irreversibility field, anisotropy, and pinning efficiency in Bi-2212 round wires

Abstract In recent years Bi 2 Sr 2 CaCu 2 O x (Bi-2212) received increasing attention due to its round wire multifilamentary architecture, a unique feature in high- T c superconductor. In fact, round wires are preferable for magnet designs, including solenoids for nuclear magnetic resonance (NMR) or research purpose and accelerator magnets. However, due to the narrow over-pressure heat treatment conditions necessary to obtain high J c and to the peculiar microstructure of Bi-2212 wires, a full understanding of the correlations between the different properties has not yet been developed. In this paper we investigate the effect of a vital part of Bi-2212 optimization, the maximum heat-treatment temperature T max in the range of 885 °C–896 °C, on the variations of J c , effective filament diameter d eff , anisotropy γ , INTER- and intra-grain irreversibility fields and pinning energies U 0 , all critical parameters in unravelling the complex mix of vortex pinning and connectivity that ultimately determines the critical current density. We found that d eff of the higher J c wires heat-treated at lower temperature is much smaller than for the lower J c wires. Moreover, a systematic increase of the irreversibility field and a decrease of the intrinsic Bi-2212 anisotropy underpins the higher J c . The analysis of the pinning energies reveals that there is little sample-to-sample variation in the INTER-grain pinning, whereas in all samples the intra-grain pinning has an enhancement below ∼40–45 K becoming more and more evident with increasing J c . These results suggest that the overall J c performance are not only related to the wire microstructure and connectivity, which obviously affect the INTER-grain properties, but they are also intimately related to the intrinsic and intra-grain properties such as γ and U 0 .

Physics↗
DOE OSTI2022

Microstructure and creep strength of simulated intercritical heat-affected zone of grade 91 steel

The intercritical heat affected zone (ICHAZ) has been reported as one of the most Type IV cracking susceptible regions in 9Cr creep-resistant steel weldments. However, creep degradation mechanisms within the ICHAZ itself need further clarifications. In this work, two ICHAZ specimens of Grade 91 steel, low-temperature ICHAZ (LT-ICHAZ) and high-temperature ICHAZ (HT-ICHAZ), were simulated using the Gleeble thermomechanical system by exposing to two peak temperatures (860 °C and 900 °C) between $A_{C1}$ and $A_{C3}$. Dramatically different creep strengths of two simulated ICHAZs were observed and studied. We report dilation curve analysis indicates a high fraction of newly transformed martensite formed in the HT-ICHAZ, which results in a much higher hardness (360 HV0.5) of the HT-ICHAZ than 266 HV0.5 of LT-ICHAZ. Precipitates, especially $M_{23}C_6$ carbides, were not fully dissolved in both ICHAZ specimens. After a typical postweld heat treatment (760 °C-2 hours), the faster recovery of low-carbon martensite and reduced precipitation strengthening due to $M_{23}C_6$ carbides coarsening in the HT-ICHAZ led to a significant hardness reduction. These microstructural degradations in the HT-ICHAZ made its creep lifetime about 36 times shorter than that of the LT-ICHAZ tested at 650 °C. The remaining tempered martensite from base metal in the LT-ICHAZ was the primary contributor to maintain its high creep resistance.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Effect of overcharge on Li(Ni 0.5 Mn 0.3 Co 0.2 )O 2 /Graphite cells–effect of binder

Cells based on NMC/graphite, containing styrene-butadiene rubber/carboxymethylcellulose binder in the anodes and pVdF latex/carboxymethykellulose in the cathodes, were systematically overcharged to 100, 120, 140, 160, 180, 250% and 270% state-of-charge. The impact of the binder was characterized by elemental analysis, SEM, and HPLC. These results were compared to similar cells just using the poly (vinylidene difluoride) binder. Not only did the binder impact the rate of transition metal transport from the cathode to the anode, it also had a marked effect on the microstructure and composition of the materials on the anode surface.

25 ENERGY STORAGE↗
DOE OSTI2021

A study of real-world micrograph data quality and machine learning model robustness

Abstract Machine-learning (ML) techniques hold the potential of enabling efficient quantitative micrograph analysis, but the robustness of ML models with respect to real-world micrograph quality variations has not been carefully evaluated. We collected thousands of scanning electron microscopy (SEM) micrographs for molecular solid materials, in which image pixel intensities vary due to both the microstructure content and microscope instrument conditions. We then built ML models to predict the ultimate compressive strength (UCS) of consolidated molecular solids, by encoding micrographs with different image feature descriptors and training a random forest regressor, and by training an end-to-end deep-learning (DL) model. Results show that instrument-induced pixel intensity signals can affect ML model predictions in a consistently negative way. As a remedy, we explored intensity normalization techniques. It is seen that intensity normalization helps to improve micrograph data quality and ML model robustness, but microscope-induced intensity variations can be difficult to eliminate.

36 MATERIALS SCIENCE↗
DOE OSTI2020

ORNL Post-irradiation Examination (LAMDA) Final Report on Characterization of Irradiated Grade 92 Ferritic-Martensitic Steels

Irradiation effects on ferritic-martensitic steel Grade 92 were investigated in this project. Samples from two heats of Grade 92, i.e., G92-2b and NF616, were selected, which were irradiated in the High Flux Isotope Reactor (HFIR) and the Advanced Test Reactor (ATR), respectively. Samples from two heats of ferriticmartensitic steel T91, irradiated in the ATR, were also selected as references for the Grade 92 samples. The selected G92-2b samples were irradiated to 0.5–14.7 displacements per atom (dpa) at two temperature regimes: 400–497 °C and 683–720°C. The selection of the high irradiation temperatures up to ~720°C is to test the performance at such upper bound temperatures, which would provide insights for the upper bound applicable temperature for Grade 92. The selected NF616 and T91 samples were irradiated to 3–8.2 dpa at 292–448°C. Except for T91 heat2 samples in 3-mm diameter discs, all the other selected samples are of type SS-J2 miniature tensile specimens. Two Nuclear Science User Facilities (NSUF), i.e., the Irradiated Materials Examination and Testing (IMET) hot cell facility and the Low Activation Materials Design and Analysis (LAMDA) facility, were used in the Post-Irradiation Examination (PIE) activities for the selected samples, which include: Tensile tests at room temperature (IMET); Fractography of the tensile fracture surfaces (LAMDA: scanning electron microscopy and energy; dispersive spectroscopy); Vickers hardness measurements (LAMDA); and Microstructural characterization (LAMDA: electron backscatter diffraction, focused-ion beam, and transmission electron microscopy).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Through-thickness fracture behavior of neutron-irradiated nuclear graphite NBG-17 Using X-ray micro-CT

Achieving precise control over crack propagation in nuclear graphite and conducting quantitative analysis remain challenging. In this study, the through-thickness fracture behavior of pristine and neutron-irradiated (700 °C, ∼7 dpa) NBG-17 nuclear graphite was investigated using split-disc testing coupled with micro-computed tomography (micro-CT). A notably lower number of micropores was observed in the neutron-irradiated specimen. The fracture toughness of neutron-irradiated NBG-17 was measured to be 1.45 MPa√m, compared to 1.17 ± 0.05 MPa√m for the pristine specimen. In both materials, cracks were found to initiate at the filler–binder interface, and often correlated with microstructural features such as pores and thermal cracks. Crack bridging and deflections emerged as the primary toughening mechanisms in both unirradiated and irradiated NBG-17. However, compared with the pristine specimen, the cracks in the neutron-irradiated specimen were more likely to grow trans granularly, resulting in less deflected crack paths. The reduced micro-porosity and strengthened filler–binder boundaries were considered to be the cause of the observed differences in crack morphologies. In conclusion, this study provides a qualitative analysis of the fracture behavior of neutron-irradiated nuclear graphite in the absence of radiolytic oxidation.

36 - MATERIALS SCIENCE↗