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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Revealing the aging process of solid electrolyte interphase on SiOx anode

Abstract As one of the most promising alternatives to graphite negative electrodes, silicon oxide (SiO x ) has been hindered by its fast capacity fading. Solid electrolyte interphase (SEI) aging on silicon SiO x has been recognized as the most critical yet least understood facet. Herein, leveraging 3D focused ion beam-scanning electron microscopy (FIB-SEM) tomographic imaging, we reveal an exceptionally characteristic SEI microstructure with an incompact inner region and a dense outer region, which overturns the prevailing belief that SEIs are homogeneous structure and reveals the SEI evolution process. Through combining nanoprobe and electron energy loss spectroscopy (EELS), it is also discovered that the electronic conductivity of thick SEI relies on the percolation network within composed of conductive agents (e.g., carbon black particles), which are embedded into the SEI upon its growth. Therefore, the free growth of SEI will gradually attenuate this electron percolation network, thereby causing capacity decay of SiO x . Based on these findings, a proof-of-concept strategy is adopted to mechanically restrict the SEI growth via applying a confining layer on top of the electrode. Through shedding light on the fundamental understanding of SEI aging for SiO x anodes, this work could potentially inspire viable improving strategies in the future.

25 ENERGY STORAGE↗
DOE OSTI2020

Determination of preferential binder oxidation in HTGR matrix material subjected to high temperature steam

Graphitic matrix material encapsulates tristructural isotropic (TRISO) coated fuel particles and is a structural component for both pebble and cylindrical fuel compacts in high temperature gas reactor (HTGR) designs. In an off-normal event involving steam exposure, the matrix material is exposed to oxidants, leading to degradation of the fuel compact and subsequent exposure of the TRISO particle fuel to oxidizing species. This study presents microstructural characterization of matrix material after exposure to high temperature (1200°C), 48 kPa steam in which the evolved microstructure demonstrated nonuniform degradation. Subsequent Raman spectroscopy determined the nature of the remaining material post-exposure. Finally, the electron microscopy characterization suggests and Raman spectroscopic analysis confirms, under off-normal conditions, the carbonized phenolic resin binder is preferentially oxidized ahead of the graphite flake filler and is responsible for the nonuniform degradation and enhanced depth of attack.

36 MATERIALS SCIENCE↗
DOE OSTI2021

The distribution and selective decontamination of carbon-14 from nuclear graphite

This work investigates 14 C release behaviour from UK Magnox reactor sourced irradiated graphite under relatively low temperature oxidation conditions: a 1% oxygen in argon atmosphere at temperatures ranging from 600°C to 700°C, and durations from 4 to 120 hours. A method is used for construction of a detailed 14 C release profile with mass loss. The 14 C release profile is predictable between samples, and an empirical release profile is derived to predict releases to higher mass losses than in this work. An accelerated 14 C release rate at small mass losses is observed, indicative of a 14 C-enriched surface region with a depleting concentration gradient into the material, and the source terms and complexities associated with predicting where the 14 C arose from are discussed. Selective decontamination of a fraction of the 14 C is possible, with the limitation of a reducing efficiency with mass loss and time. This work demonstrates a method for determining the distribution of 14 C between near-surface regions and the bulk material in differing sources of irradiated graphite, and provides data in support of making an informed assessment for the adoption of low temperature thermal treatment and isotopic reduction of irradiated graphite wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Synthesis of Interface-Driven Tunable Bandgap Metal Oxides

Mixed bandgap and bandgap tunability in semiconductors is critical in expanding their use. Composition altera-tions through single-crystal epitaxial growth and formation of multi-layer tandem structures are often employed to achieve mixed bandgaps albeit with limited tunability. In this study, self-assembled one-dimensional coordination polymers provide a fac-ile synthon and template for graphitic C-doped mesoporous oxides, gC-β-Ga 2 O 3 or gC-In 2 O 3 via controlled oxidative ligand ablation. These materials have mixed bandgaps, and colors, depending on amount of gC present. The carbon-oxide interface leads to induced gap states, hence, a stoichiometrically tunable band structure. Structurally, a multi-scale porous network percolating throughout the material is realized. The nature of heat-treatment and top-down process allows for facile tunabil-ity and formation of mixed bandgap metal oxides through controlled carbon deposition. As a proof of concept, gC-β-Ga 2 O 3 was utilized as a photocatalyst for CO 2 reduction, which demonstrated excellent selectivity and conversion rates into CH 4 and CO.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Graphite Degradation Modeling and Analysis

A graphite component in a nuclear reactor core is subjected to variety of stresses and can experience degradation during normal and off-normal operation. Understanding how a graphite component will behave in service is essential to ensuring core structural stability and safe reactor operation. This report summarizes a graphite modeling tool currently under development at Idaho National Laboratory. The model incorporates several of the anticipated stresses during service and includes the effects of oxidation and irradiation prior to turnaround. This tool is intended to be used to help assess the design of graphite components by utilizing design code rules found in Section III, Division 5 of the American Society of Mechanical Engineering Boiler and Pressure Vessel Code. Specifically, the tool uses the methodologies found within the Full and Simplified assessments from Article HHA-3000 to verify that a graphite component has an acceptably low probability of failure.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Graphite Licensing (ASME) and Testing (ASTM) Technical Assistance

DOE ART Graphite R&D program is currently involved in developing the structural code for licensing the new High Temperature Reactor (HTR) designs through the ASME Boiler Pressure Vessel Code (BPVC). The basic structural design code for graphite core components has been written, approved, and within the BPVC since 2012. However, several areas within the graphite code remain to be improved, specifically the addition of material property changes resulting from environmental degradation (neutron flux and oxidation). The incorporation of this data, which can significantly affect the graphite components and alter the structural integrity of the core during operation, is proving to be an important activity. How this data will be incorporated within the code and how it will be used for license applications will be critical to a number of HTR applicants in the future. These modifications to the code must take into account the unique irradiation and oxidation response of the wide spectrum of available nuclear graphite grades currently under consideration for HTR applications. Additionally, the ASME BPVC requires material property testing within the expected operational conditions of the HTR core including neutron radiation flux, elevated temperatures, and after chronic/acute oxidation of the graphite components. Currently, no standardized graphite testing methods are approved for these environmental conditions. New American Society for Testing and Materials (ASTM) standardized tests for elevated temperature testing, small sample testing, and testing after oxidation must be developed to support the requirements within the ASME code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Probing basal planes and edge sites in polygranular nuclear graphite by gas adsorption: Estimation of active surface area

Polygranular nuclear graphite, manufactured at 2800–3000 °C from a carbonized filler and binder mix, has high graphitization degree, crystalline structure close to perfect graphite, and about 20% porosity. The pore surfaces expose large regions of rough, defective prismatic edges of graphite crystallites which are the locus of graphite materials surface sites active for oxidation, chemisorption, and electron transfer. However, we show that high-resolution N 2 and Kr first monolayer adsorption on polygranular graphite (P/P 0 < 0.015) occurs in many ways like adsorption on graphitized carbon blacks. This proves the presence of energetically homogeneous basal planes domains in graphite porosity, which was not fully acknowledged before. Using classical analysis methods (Langmuir, Hill-de Boer, adsorption potential distribution) we quantity the basal plane area (BPA) of several polygranular graphite types and correlate it with their microstructure. Finally, we propose that gas adsorption is uniquely positioned to reliably estimate the active surface area of graphite by subtracting BPA from the BET total surface area. Direct estimation based on adsorption is preferable to indirect calculations based on microstructural information.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Review of Graphite Properties Relevant to Micro-Reactor Design

Microreactors are advanced reactors that are designed to be small, scalable, and portable with up to 20 MW power. These reactors could operate independently, or as part of a microgrid, without need of a frequent refueling. Micro-reactors are designed to be deployable to provide electric power in remote locations, therefore providing greater flexibility. Several micro-reactors are under development in the US. These reactors are based on advanced gas-cooled or heat-pipe concepts. While Graphite can be found in various applications, some high-temperature reactors plan to employ Graphite in their core components due to its role as a moderator and its ability to maintain its structural integrity when subjected to elevated temperatures and high neutron fluence. Although there is significant experience with use of Graphite in nuclear applications, there are still unknowns that should be addressed. Graphite properties depend on the graphite type and affected by the fabrication procedures. Material properties are functions of the temperature and irradiation dose. The fundamental properties also affected by the oxidation environment. Therefore, in-depth understanding of graphite properties, including the effect of irradiation, oxidation and other influencing parameters are required to evaluate the design accurately. This draft aims to provide a comprehensive review of material properties for Graphite relevant to micro reactor design. For this work, two distinct grades of graphite were considered, namely IG-110 and Mersen-2114. Physical, thermal, and mechanical properties for these grades were comparatively evaluated. The properties that are discussed in this draft aim to serve as a basis for design and safety analysis.

42 ENGINEERING↗
DOE OSTI2023

Degradation and Modeling of Large-Format Commercial Lithium-Ion Cells as a Function of Chemistry, Design, and Aging Conditions

Demand for large-format (>10 Ah) lithium-ion batteries has increased substantially in recent years, due to the growth of both electric vehicle and stationary energy storage markets. The economics of these applications is sensitive to the lifetime of the batteries, and end-of-life can either be due to energy or power limitations. Despite this, there is little information from cell manufacturers on the sensitivity of cell degradation to environmental conditions or battery use. This work reports accelerated aging test data from four commercial large-format lithium-ion batteries from three manufacturers, with varying design (thickness, casings, ...), chemistry (lithium-iron-phosphate (LFP) or lithium-nickel-manganese-cobalt-oxide positive electrodes (NMC), with graphite (Gr) negative electrodes), and capacity (50 to 250 Amp hours). The tested LFP|Gr cell is found to be relatively insensitive to cycling conditions like temperature or voltage window, while NMC|Gr cells have varying sensitivity. Degradation trends are further investigated by training predictive models: simple polynomial trend lines, a semi-empirical reduced-order model, and an empirical reduced-order model identified using machine-learning based on symbolic regression. Calendar and cycle life are simulated over a variety of conditions to directly compare the various batteries. Cell size and thickness are found to substantially impact sensitivity to temperature during cycle aging, while electrode chemistry impacts depth-of-discharge sensitivity. Real-world battery lifetime is evaluated by simulating residential energy storage and commercial frequency containment reserve systems in several U.S. climate regions. Predicted lifetime across cell types varies from 7 years to 20+ years, though all cells are predicted to have at least 10 year life in certain conditions.

battery lifetime↗
DOE OSTI2024

Viability of Additively Manufactured Electrodes for Lithium-Ion Batteries

As the global economy becomes increasingly electrified, the demand for batteries and energy storage is expected to rise significantly, particularly in the transportation and electricity sectors. Lithium-ion batteries (LIBs) are currently the most advanced and widely used technology in this field. Traditionally, LIBs are manufactured using simple 2D planar geometries to maximize production efficiency and minimize costs. However, this approach limits energy density due to the restricted design flexibility of the electrodes. Additive manufacturing (AM) offers a promising solution to enhance the energy density and efficiency of LIBs by enabling the design of architectures that reduce diffusive losses and allow for a greater amount of active material to be incorporated within the same device footprint, thereby minimizing the use of inactive materials. Different AM techniques come with their own set of limitations, including printing speed, material compatibility, and scale, which must be considered when designing electrodes. Scalable and cost-effective methods are particularly important for electric vehicle batteries, while achieving higher energy densities in microbatteries is crucial for the miniaturization of wearable electronics and medical devices. Here, in this study, we simulate various 3D porous electrode designs for LIBs using graphite and nickel manganese cobalt oxide (NMC) electrodes. These designs are selected to represent structures that could be produced using different AM techniques, such as direct ink writing, fused deposition modeling, and stereolithography. Our results indicate that at higher charging rates and increased areal mass loading, 3D structures can outperform traditional 2D electrodes, although the benefits may diminish with more complex designs that are harder to manufacture. The observed gains in energy density are attributed to improved electrode utilization and reduced diffusive energy losses. This comprehensive analysis of structure–performance relationships will provide valuable insights to guide future research on 3D designs, material selection, and AM techniques for additively manufactured battery electrodes.

25 ENERGY STORAGE↗
DOE OSTI2021

Electrochemical ion insertion from the atomic to the device scale

Electrochemical ion insertion involves coupled ion–electron transfer reactions, transport of guest species and redox of the host. The hosts are typically anisotropic solids with 2D conduction planes but can also be materials with 1D or isotropic transport pathways. These insertion compounds have traditionally been studied in the context of energy storage but also find extensive applications in electrocatalysis, optoelectronics and computing. Recent developments in operando, ultrafast and high-resolution characterization methods, as well as accurate theoretical simulation methods, have led to a renaissance in the understanding of ion-insertion compounds. In this Review, we present a unified framework for understanding insertion compounds across timescales and length scales ranging from atomic to device levels. Using graphite, transition metal dichalcogenides, layered oxides, oxyhydroxides and olivines as examples, we explore commonalities in these materials in terms of point defects, interfacial reactions and phase transformations. We illustrate similarities in the operating principles of various ion-insertion devices, ranging from batteries and electrocatalysts to electrochromics and thermal transistors, with the goal of unifying research across disciplinary boundaries.

36 MATERIALS SCIENCE↗
DOE OSTI2025

TRISO Particle Isolation and Graphite Removal Using SRNL Vapor Digestion Technology

Tri-structural isotropic fueled reactors are planned to come online in the next decade, but there is not currently a widely agreed upon disposition pathway for this new fuel stream. For long term used nuclear fuel storage, there would be a significant benefit if the waste volume could be reduced or mitigated. Most of the volume from used TRISO fuel contains graphite. SRNL currently has an active patent which describes one pathway to digest the graphite. One benefit in the utilization of this pathway is that multiple cylindrical pieces can be stacked end-to-end in a reaction tube, then nitric acid and water vapor can be flowed through the tube, and the weight change of the individual cores can provide both a reaction profile and overall oxidant use efficiency. The reaction of nitric acid and graphite passes through a series of intermediate products – NO 2 , NO, and N 2 O – from reaction and decomposition to eventually form N2. The data shows that NO2 grows in with increasing temperature between 500-600°C and then decreases by way of either reaction with graphite or decomposition. Nitric oxide, a reaction and thermal decomposition product of NO 2 , exhibits a consistent decline as a function of temperature, which is consistent with the literature. Similarly, the concentrations of N 2 O and N 2 increase as a function of temperature as their reactions with graphite become more favorable.

Advanced Reactors↗
DOE OSTI2025

Upcycling Polyethylene Waste into Hybrid Graphitic Porous Carbon Materials Used in High‐Performance Zinc‐Ion Hybrid Capacitors

Polyethylene (PE) waste is a challenge to upcycle into useful materials because this plastic tends to decompose into volatile compounds when heated at relatively low temperatures. In this work, mixtures of PE wastes into a hybrid graphitic porous carbon (HGPC) by a thermal oxidation pretreatment step, with assistance of an inert solid additive (KCl), to functionalize, crosslink, and stabilize the PE waste followed by carbonization and catalytic graphitization steps with a potassium carbonate catalyst, are upcycled. The PE waste‐derived HGPC (PW‐HGPC) has a hybrid structure composed of graphene‐like carbon nanosheets grown on the surface of carbon particles, high porosity with specific surface area, up to 1,763 m 2 g −1 , and good graphitic degree with average Raman I 2D / I G ratios of 0.53. When used as cathode material for zinc‐ion hybrid capacitors, this PW‐HGPC exhibits an excellent specific capacity, up to 126.7 mAh g −1 , at high mass loading of 10 mg cm −2 . Moreover, PW‐HGPC exhibits remarkable cycling stability with capacity retention of >94% after 10 000 cycles. Additionally, the KCl is recycled and reused over five times. This method provides a new solution for upcycling PE wastes into high value‐added carbon materials, not only for zinc‐ion hybrid capacitors but also for other electrochemical energy storage device applications.

hybrid graphitic porous carbon↗
DOE OSTI2022

Perspective on “code qualifying” new graphite grades for use in advanced nuclear reactors*

The American Society of Mechanical Engineers (ASME) publishes the Boiler and Pressure Vessel (BPV) Code, which include guidance for the safe development, construction, and operation of boilers and pressure vessels. ASME BPV Code Section III “Rules for Construction of Nuclear Facility Components” Division 5 focuses on “High Temperature Reactors”. Subsection HH, subpart A lists the different materials properties that are to be measured and how those properties change due to different environmental conditions (oxidation and irradiation damage) for a graphite to be accepted for use in a high temperature reactor core (i.e., “Code Qualified”). Currently there are no nuclear graphite grades that are “Code Qualified” (i.e., a reactor designer can select a graphite grade and build their reactor without any additional testing), which is due in part to development of new graphite grades in the last 20 years and the lack of comprehensive programs needed to produce the data for the code cases. This perspective is going to discuss the requirements, as called out in the ASME BPV Code, that are necessary to “code qualify” a nuclear graphite grade but will primarily focus on the practical and technical challenges associated with irradiation-induced property changes and how to address these to assist with getting graphite ready for use in advanced nuclear reactors. These same technical challenges can be expected to arise for other materials being developed for advanced reactor concepts.

advanced nuclear reactors↗
DOE OSTI2026

Mechanistic insights into low-temperature oxidation of carbon fibers: Influence of hydrogen defects and crystallite size

Although oxidation mechanisms have been exhaustively studied for graphite, similar analyses of carbon fibers are comparatively sparse. Most prior work has focused on quantifying weight loss or assessing protective surface coatings designed to slow oxidation. The use of optical spectroscopic techniques for oxidation analyses is comparatively unexplored, but such techniques could provide an early indicator of fiber oxidation that would undermine carbon fiber performance. Here, in this work, we applied Raman spectroscopy to study oxidation-induced spectral alterations in 16 carbon fiber types from 7 manufacturers oxidized at 300 °C for 72 h, 400 °C for 8 h, and 500 °C for 1 h. We connect these results with structural properties of the carbon fibers obtained through wide-angle X-ray scattering, identifying a linear dependence between the reactivity of carbon fibers and the crystallite size of the unperturbed fibers. We then demonstrate that substituted hydrogen defects are likely removed from the fiber surface during oxidation and use the relative defect concentration to predict the Raman spectral change as a function of temperature and time, assuming Arrhenius behavior.

Carbon fiber↗
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↗