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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Interaction of beryllium with 316H stainless steel in molten Li 2 BeF 4 (FLiBe)

The increased demand in renewable energy resources has led to a renewed interest in nuclear reactors including a new generation of molten salt reactors. One of the issues encountered in handling of molten chloride or fluoride salts is the inability of metals to form and retain a protective oxide surface layer, generally chromium oxide. The low chromium content, nickel-based alloy INOR-8, later marketed as Hastelloy N, was specifically developed to contain molten fluorides. This alloy had some deficiencies, but its composition resulted in it incurring far less chromium depletion by the molten fluoride salts than alloys like 316 stainless steel. For the next generation of molten salt reactors, an alloy with pressure vessel code approval and higher temperature capability than Hastelloy N is required. Consequently, several reactor designers have reportedly chosen to build the reactor containment vessel from 316H stainless steel. In order to minimize corrosion of the stainless steel by the molten fluoride salt (2LiF-BeF 2 ), it is proposed to add beryllium metal to the salt to react with impurities and provide a means to lower the oxidation potential thus making the salt less corrosive. Even so, there is a concern whether the beryllium would react with components of the stainless steel. To address this concern, static capsule tests were conducted in which selected amounts of beryllium were added to capsules containing 2LiF-BeF 2 (FLiBe) salt and 316H stainless steel and Hastelloy N samples. Following exposure, the samples were cleaned and examined using a variety of analysis techniques including, optical microscopy, SEM-EDS, XPS, LIBS, XRD and EPMA. Tensile samples were also exposed in capsules, and those were subjected to tensile testing. Notably, It was found that for the level of beryllium additions used in this study, intermetallic compounds were formed which could be detrimental to the long-term performance of the 316H stainless steel.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Possible impacts of Mo chemical banding and second phase impurities on the irradiation behavior of monolithic U-10Mo fuels

This study investigated the microstructural behavior of both full-size and mini-size monolithic U-10Mo fuel plates irradiated to high burnup with a focus on the evolution of the second phase impurities in monolithic U-Mo using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and wavelength dispersive spectroscopy (WDS). Key indicators of possible mechanical and thermal compromise include cracks, large fission gas porosity, and interconnection of fission gas pores. For the fission densities evaluated in this work (3.5 × 10 21 fissions/cm 3 -5.1 × 10 21 fissions/cm 3 ), fine porosity can develop along the UC phase boundary; however, the size of the fission gas pores is no more than those observed in the U-Mo fuel phase. Further, other inclusions such as Si-rich second-phase impurities found in the as-fabricated microstructure were difficult to resolve post-irradiation because they can become overshadowed by porosity development in the fuel phase. Additionally, the presence of a Fe-rich sublayer formed in the Zr diffusion layer during fabrication remained enriched in the Zr layer in the irradiated U-10Mo microstructures near the U-Mo/Zr interface; however, based on the burnup assessed in this study the identified impurities did not appear to contribute to notable microstructural degradation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Analysis of neptunium oxides produced through modified direct denitration

Production of neptunium-237 ( 237 Np) target materials for plutonium-238 ( 238 Pu) radioisotope thermoelectric generators (RTGs) for deep space exploration requires advanced chemistry and engineering development. Currently, the domestic Pu-238 Supply Program at Oak Ridge National Laboratory produces neptunium dioxide (NpO 2 ) for target material using a modified direct denitration (MDD) flowsheet. Although the chemistry, reaction mechanisms, and product characteristics of MDD are well understood for uranium, corresponding studies of the neptunium system are still needed to continue optimization of target material properties, production equipment design, and production flowsheets. Here, the objective of this work is to characterize crystalline phases, morphology, surface texture, and particle size of NpO 2 produced via MDD reactions. Solid-phase characterization techniques, including powder X-ray diffraction (pXRD) and scanning electron microscopy with energy-dispersive spectroscopy (SEM-EDS), were employed to achieve this objective. Subsequent data processing using the Morphological Analysis for Material Attribution (MAMA) software was performed to analyze particle morphology and size. Broadly, the powders were found to contain a mixture of NpO 2 and Np 2 O 5 after denitration with a variety of morphologies. After high-firing, the product was found to be NpO 2 with a typical polycrystalline oxide morphology and a grain size ranging from 0.72 to 0.94 µm. These analyses provide knowledge on the reaction pathway for a non-traditional NpO 2 synthesis method and offer additional unique insight into production-scale environments for transuranic materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Investigation of W-SiC compositionally graded films as a divertor material

W-SiC composite material is a promising plasma-facing material candidate alternative to pure W due to the low neutron activation, low impurity radiation, and low tritium diffusivity of SiC while leveraging the high erosion resistance of the W armor. Additionally, W and SiC have high thermomechanical compatibility given their similar thermal expansion rates. The present study addresses the synthesis and performance of compositionally graded W-SiC films fabricated by pulsed-DC magnetron sputtering. Compositional gradients were characterized using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy (EDS), and crystallographic information was obtained using electron diffraction and X-ray diffraction (XRD). Samples were exposed to L-mode deuterium plasma discharges in the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Post-mortem characterizations were performed using scanning electron microscopy (SEM) and XRD. Electron diffraction and XRD showed that the compositionally graded W-SiC films were composed of polycrystalline W and amorphous SiC with amorphous W+SiC interlayers. No macroscopic delamination or microstructural changes were observed under mild exposure conditions. Here, this study serves as a preliminary examination of W-SiC compositionally graded composites as a potential candidate divertor material in future tokamak devices.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Microstructural evolution and hardness changes in ion irradiated nickel-based Haynes 282 superalloy

This study investigated the crucial aspects of thermal and irradiation stability in precipitation hardened Haynes 282 Ni-based superalloy. The Haynes 282 Ni-based alloy was irradiated by 8 MeV Ni 3+ ions to assess its resistance to phase instabilities and mechanical property alterations. The mid-range doses (at a depth of ∼1 µm) were 1 and 10 displacements per atom (dpa) at temperatures of 600°C and 750°C. Nanoindentation tests provided insights into bulk equivalent hardness of the irradiated and pristine regions, while scanning transmission electron microscopy (STEM) and energy dispersive X-ray spectroscopy (EDS) were used to examine the microstructural evolution of irradiation-induced defects and γ′ -Ni 3 (Al, Ti) precipitates and defect structure under irradiation. These precipitates, with an average diameter of 29 nm and a number density of 5 x 10 21 /m 3 , acted as robust dispersion strengthening agents with high radiation point defect sink strength. Remarkably, irradiation did not significantly alter the size or number density of the γ′ precipitates, indicating exceptional thermal stability and radiation resistance of these precipitates at the examined conditions. Radiation-induced dislocation loops were observed at 600°C, albeit without a substantial impact on mechanical properties due to the dominance of γ′ precipitates on the overall alloy strength. The superior stability of γ′ precipitates observed in this study contrasts with several previous research findings on Ni-based alloys that reported poor precipitate stability. Plausible reasons for this difference are discussed. Moreover, this work explicitly outlines a physically grounded approach to ensure accurate microstructure-hardness correlations and clarifies the hardening model by addressing common misapplications of superposition in prior studies.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Characterization of in-situ and ex-situ ion-irradiated additively manufactured 316L and 316H stainless steels

Additively manufactured (AM) 316 stainless steel (SS) differs from its wrought counterpart in its unique dislocation cell structure and the presence of segregation and oxide particles at the cell walls. This work investigated the evolution of the microstructure in laser powder bed fusion (LPBF) 316L and 316H SS under in-situ 1 MeV Kr ion irradiation at 600 °C to 5 dpa, and ex-situ 4 MeV Ni ion irradiation at 300 °C and 600 °C from 0.2 dpa to 10 dpa, with a dose rate for all experiments of 10 -3 dpa/s. The results reveal that the dislocation cell structure results in heterogeneous formation of dislocation loops and voids, particularly at 600 °C, where loops tend to form within the cell interiors while voids form at the cell boundaries. LPBF 316H has a reduced level of swelling compared to LPBF 316L due to prolonged incubation. Energy Dispersive X-ray Spectroscopy (EDS) mapping indicates Ni and Si segregation at void surfaces due to radiation-induced segregation. At 300 °C, where voids are absent, the distribution of dislocation loops and stacking fault tetrahedra appears to be uniform. Dislocation cell structures mostly disappeared by 2 dpa for all conditions in this work. M 23 C 6 carbides were observed in LPBF 316H at 600 °C as early as 0.2 dpa, but not in LPBF 316L. Nanoindentation was performed to obtain the hardness of irradiated materials. In conclusion, this work illustrated the influence of additive manufacturing processes on microstructure evolution under irradiation, revealing the differences as well as the similarities as compared with wrought 316 SS, and the AM-related phenomenon that can potentially occur under neutron irradiation in nuclear reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Direct reuse of graphite and lithium nickel manganese cobalt oxide (NMC) recovered from ultrafast-laser ablation debris in Li-ion battery electrodes

Here we demonstrate that debris collected from the ultrafast-laser ablation of graphite anodes can be directly reused in a lithium-ion battery with little to no negative effects on electrochemical performance. Further, we show that while post-ablation NMC (LiNi 0.33 Mn 0.33 Co 0.33 O) cathode debris will require additional processing before being incorporated into an electrode, its critical materials (nickel, manganese, and cobalt) are not lost during ablation and can be recovered for recycling. Pre- and post-ablation materials are characterized with a suite of diagnostics, including SEM, TEM, X-ray CT, EDS, and XRD, to study changes in material morphology, composition and crystal structure. Graphite exhibited little to no morphological or compositional changes and a slight annealing of its crystal structure. NMC underwent profound morphological and crystallographic changes, but retained its elemental composition. Finally, post-ablation materials were re-manufactured into electrodes and cycled vs lithium metal. Graphite showed equal or better capacity and Coulombic efficiency compared to pre-ablation electrodes, while the post-ablation NMC exhibited severely reduced electrochemical performance. Furthermore, we discuss the outlook for application of this work to advanced manufacturing lines that produce next-generation, laser-patterned electrodes.

25 ENERGY STORAGE↗
DOE OSTI2026

Tetrahedral Zn 2+ doping LiNi 0.6 Mn 0.2 Co 0.2 O 2 improves discharge capacity retention by altering surface Ni valence during cycling and preventing oxygen evolution

5 mol-% Zn 2+ was added to the co-precipitation synthesis of LiNi 0.6 Mn 0.2 Co 0.2 O 2 (NMC622) to understand Zn's effect on the structural and electronic properties of NMC622. Zn is determined to successfully dope into the NMC crystal lattice as there is an increase in d-spacing along the [003], [104], and [110] directions and an increase in molar volume (100.75 cm 3 mol -1 for NMC622 and 101.00 cm 3 mol -1 for Zn-NMC). Through XPS, Zn is determined to dope on the tetrahedral 6c Wyckoff site in the NMC structure and TEM-EDS reveals that despite a slight ZnO impurity phase, 1.4 mol-% Zn successfully dopes into the crystal structure. Furthermore, cyclic voltammetry shows a suppressed anodic wave associated with oxygen evolution and EIS demonstrates that Zn-doped NMC decreases charge transfer resistance and increases Li diffusion by an order of magnitude. After the formation cycles, NMC has a higher 1 C accessible capacity, 141 mAh/g, over Zn-NMC (132 mAh/g), but after long term cycling Zn-NMC has a higher accessible capacity (117 mAh/g compared to 96 mAh/g for NMC) and increases discharge capacity retention by 18% leading to longer cycle life.

Electrochemistry↗
DOE OSTI2026

Unravelling fast-charging degradation in NMC/Gr pouch cells: Lithium plating and SEI properties

As fast-charging technology expands across the electric vehicle and emerging energy-storage applications, understanding its impact on battery performance and longevity is critical. In this study, 1.8 Ah LiNi 0.6 Mn 0.2 Co 0.2 O 2 /graphite pouch cells were charged at various charging rates (0.5C, 2C, 4C, and 6C) to investigate the degradation mechanisms. Our results showed that well-designed NMC/Gr pouch cells could reach over 1000 cycles with a 2C charging rate, while only reaching around 500 cycles with 4C and 6C charging rates. Fast-charging effects on NMC and graphite electrodes were obtained through a series of post-mortem characterizations, including electrochemical impedance spectroscopy (EIS), Raman spectroscopy, X-ray photoelectron spectroscopy (XPS), and scanning electron microscopy (SEM) with energy-dispersive X-ray spectroscopy (EDS). Although higher charging rates cause pulverization of NMC secondary particles, the dominant degradation mechanism driving the fading of fast-charging-related performance lies in the graphite anode, where lithium plating and LiF-rich solid electrolyte interphase (SEI) formation result in Li inventory loss and impedance growth. The postmortem results suggest that the formation of a LiF-rich SEI, which exacerbates anode impedance and some irreversible Li + ion loss, is likely driven by the substantial decomposition of PF 6 − during fast charging, an effect often overlooked in smaller laboratory-scale studies.

Luo, Mei [Argonne National Laboratory (ANL), Argon↗
DOE OSTI2020

Effect of aging on failure mechanism of Alloy 709 at various temperatures

Alloy 709 is a novel austenitic stainless steel with high temperature creep strength, weldability, and corrosion resistance. These properties make the material suitable for applications in the structure of next-generation nuclear power plants. Enduring high temperatures for an extended period of time in the harsh environments of a nuclear power plant results in thermal aging of the material. Therefore, it is imperative to study the effect of thermal aging on the microstructure and mechanical properties of Alloy 709 before its application in the next generation power plants. In this study, hot-processed (forged + rolled), annealed and quenched ingots of Alloy 709 are aged at 650 °C for 2000 h in air and then tested at various temperatures up to 850 °C in an in-situ heating-loading Scanning Electron Microscope (SEM) equipped with Energy Dispersive Spectroscopy (EDS) and Electron Backscatter Diffraction (EBSD). The effect of testing temperature and aging on microstructural evolutions during tensile testing of as-received and aged samples are studied. Furthermore, both as-received and aged samples displayed serrations and drop in ductility at elevated temperatures, due to the effect of dynamic strain aging (DSA). The occurrence of DSA activity was found within temperatures range of 500 °C–750 °C in as-received and 550 °C - 650 °C in aged samples. The aged samples showed less elongation accompanied by shallower dimples on the fractured surface, indicating less ductile failure mechanism compared to as-received samples. Failure mechanism observations on the fracture surface using SEM fractography are correlated to the observations made on the sample surface using in-situ SEM to achieve a complementary set of information to better understand the failure mechanism of this novel alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2021

A study on the creep behavior of alloy 709 using in-situ scanning electron microscopy

In this research, an experimental evaluation of creep properties of Alloy 709 in the temperature range of 750–850 °C was undertaken. Alloy 709 is a novel austenitic stainless steel with 20% Cr and 25% Ni by wt% that was developed for application in structural components of nuclear power plants. Creep rupture tests were conducted in an in-situ heating-loading and Scanning Electron Microscope (SEM) unit equipped with Electron Backscatter Diffraction (EBSD) detector and Energy Dispersive Spectroscopy (EDS). “Real-time” creep damage mechanisms of Alloy 709 at various stresses and temperatures using a flat, un-notched sample with continuously reducing cross-section is studied so that the failure and maximum creep damage occurred at the center of the sample where the in-situ SEM imaging could be focused. Accelerated creep tests at temperatures and stresses above service conditions were performed by employing multiple blocks of constant loads where the loads were increased once the sample attained constant creep rate, indicating a secondary creep regime. This technique ensures multiple data points can be obtained from the same test, saves the time required for an otherwise long-term creep test and usage of SEM. Further, Coincident Site Lattice (CSL) boundary maps were collected as control maps before testing, and the grain boundaries were observed during the creep test to understand the effect of grain boundary character on the creep damage mechanism. Void growth, grain boundary separation, and sliding were found to be the main creep mechanisms whose rate is dependent on stress and temperature. Failure mechanisms studied on the fracture surface using SEM fractography were correlated to the sample surface observations to create complementary information to better understand the underline creep mechanism of Alloy 709.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Chemistry effects on ODS steel consolidated via laser powder bed fusion from GARS powder

Oxide Dispersion Strengthened (ODS) steels are promising candidate alloys for structural and cladding applications in extreme environments. They contain a high density of nanoscale oxides for high temperature mechanical strength and radiation resistance. In this work, gas atomization reaction synthesis (GARS) was used to produce powders that were used for additive manufacturing (AM) Laser Powder Bed Fusion consolidation of ODS steels, in order to skip the traditional mechanical alloying of blended yttria and alloy powders. Powder containing iron, chromium, and tungsten with varying amounts of yttrium, titanium, oxygen and zirconium were used to produce ODS steel samples. AM consolidated specimens and powder samples were characterized with transmission electron microscopy. TEM imaging, diffraction patterns, and energy dispersive X-ray spectroscopy (EDS) was used to identify phases present before and after consolidation across chemistries. The effect of the controlled oxygen input (from GARS) and the oxide-forming additions (Y, Ti, Zr) on precipitate size distribution and composition is substantiated and discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2026

FIB-ToF-SIMS characterization of irradiated U-10Zr

Post-irradiation examination (PIE) is critical for the performance assessment and qualification of nuclear fuels. Secondary ion mass spectrometry (SIMS) is a powerful materials characterization technique that allows for elemental and isotopic mapping with a depth resolution greater than EDS and EPMA. However, it has not yet been applied to PIE of metallic nuclear fuel. Here, in this work, we characterize an fast neutron spectrum irradiated U-10Zr fuel sample using a time-of-flight SIMS (ToF-SIMS) system connected to a FIB/SEM system, which allows for flexible sample analysis compared to a dedicated ToF-SIMS instrument. Analysis of the resulting hyperspectral micrograph data was aided by the development of an unsupervised machine learning (ML) algorithm that iterates on existing methods to segment the 3D micrographic datasets based on the similarity of mass spectra. The results showed that the FIB-ToF-SIMS instrument was potentially capable of spatially resolving closed fission gas bubbles in 3D by continued ion sputtering of the analyzed volume. Additionally, the ML algorithm proved useful in revealing the chemical segregation of light fission products (those with an atomic mass between approximately 85–105 amu, such as ruthenium and rhodium) plus matrix zirconium, heavy fission products (those with an atomic mass between approximately 135–150 amu, such as the lanthanides) and uranium. Future studies are planned to conduct FIB-ToF-SIMS analysis on more irradiated U-Zr samples to study the constituent redistribution.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Catalysts by pyrolysis: Direct observation of transformations during re-pyrolysis of transition metal-nitrogen-carbon materials leading to state-of-the-art platinum group metal-free electrocatalyst

Transition metal-nitrogen-carbon (M-N-C) materials have been the focus of scientists’ efforts to address the rising need for earth-abundant materials solutions for energy technology and decarbonization of the economy. They are viewed as one of the most promising candidates to replace platinum group metal (PGM) catalysts in the fuel cell and energy conversion fields, including the application of oxygen reduction reaction, carbon dioxide reduction reaction, and nitrogen reduction reaction. In the effort to improve M-N-C materials properties and achieve atomic dispersity of the transition metal in the carbonaceous matrix, a re-pyrolysis process has been proposed. This secondary heat treatment process of already obtained primary pyrolysis-derived M-N-C materials has been widely reported to substantially improve the electrochemical performance and operational stability of the catalysts. Here, we report a systematic investigation of this process used on samples of templated M-N-C catalysts to obtain state-of-the-art catalysts via in situ heating X-ray photoelectron spectroscopy (XPS), scanning transmission electron microscopy (STEM), energy-dispersive X-ray spectroscopy (EDS), electron energy loss spectroscopy (EELS), X-ray diffraction (XRD), and X-ray computed tomography (CT) characterization methods. It is found that the re-pyrolysis of M-N-C materials could result in the partial amorphization of the carbonaceous substrate. It causes the rearrangement and transformation of multitudinous N moieties, leading to optimization of their morphological display and association with atomically dispersed transition metal dopants. Ultimately, the re-pyrolysis results in an increase in uniformity of the active Fe-Nx sites distribution without the formation of nano-crystalline phases (metallic or carbide) and with overall preservation of the morphology of the carbonaceous framework achieved during the first formative pyrolysis step of the templated synthesis. Finally, these observations provide confirmation that empirically established re-pyrolysis is recommended to be used on all M-N-C materials despite the different synthesis routes to obtain a practical advanced catalytic material.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Application of GO anchored mediator in a polymer electrolyte membrane for high-rate solid-state supercapacitors

Here, we synthesized a novel polymer electrolyte membrane by combining poly (vinylidene fluoride) (PVDF) and lithium bis(trifluoromethanesulfonyl)imide (LiTFSI) with graphene oxide (GO) nanosheets and a lithium salt of tungstosilicic acid (Li 4 SiW 12 O 40 , hereafter, referred to SiWLi). The impact of the addition of GO/SiWLi on the microstructure and morphology of the membrane were characterized using X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), Fourier transform infrared spectroscopy (FTIR), and X-ray photoelectron spectroscopy (XPS). We found that adding the GO/SiWLi to the PVDF/LiTFSI polymer electrolyte membrane significantly reduced the pore size. Furthermore, the addition of the GO/SiWLi resulted in not only an increase of the ionic conductivity from 0.87 $\times$ 10 –2 to 3.12 $\times$ 10 –2 Scm –1 but also an increase in the lithium-ion transference number from 0.52 to 0.87. The polymer electrolyte membranes with and without GO/SiWLi were utilized to fabricate solid-state supercapacitors. The supercapacitors fabricated with the membrane containing GO/SiWLi displayed 37.2% lower equivalent series resistance and 88.2% greater specific capacitance than those fabricated using the membrane without GO/SiWLi at 200mVs –1 .

42 ENGINEERING↗
DOE OSTI2023

Determination of film thicknesses of metal oxides prepared by atomic layer deposition on SBA-15

ZrO 2 and CeO 2 films were grown in SBA-15 by Atomic Layer Deposition (ALD) to loadings of 0.92 g oxide/g SBA-15. Scanning Transmission Electron Microscopy (STEM) with Energy Dispersive Spectra (EDS) showed that the oxides grew uniformly inside the mesopores. Film thicknesses were then analyzed as a function of the number of ALD cycles using three methods: 1) mass changes assuming bulk densities for the films; 2) changes in pore size from Barret-Joyner-Halenda (BJH) analysis with N 2 adsorption isotherms; and 3) small-angle X-ray scattering (SAXS). Film thicknesses assuming bulk densities were between 6 and 8 times smaller than those obtained from BJH analysis. SAXS analysis gave film thicknesses that were approximately twice that obtained from bulk densities. Here, possible explanations of the discrepancies between these methods are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A practical guide to characterizing irradiated nuclear fuels using FIB tomography

We report focused ion beam (FIB) tomography with combined electron backscatter diffraction (EBSD) and energy dispersive x-ray spectroscopy (EDS) is a technique capable of statistically characterizing the microstructure and spatial compositional variation of nuclear fuel in three-dimensions (3D). The 3D visualization from FIB tomography provides a comprehensive picture of the interconnected microstructural and compositional features that can impact fuel performance. While these features are often characterized with surface examination, the complexity and relationship of fission products and grain boundary networks may not fully be captured by these 2D methods. This work presents a practical guide to FIB tomography that is tailored to nuclear fuel characterization. The steps used to collect and process the data are provided along with the scripts used to process the data. Additionally, suggestions for future characterization efforts utilizing this approach are given.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗