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Achieving Highly Reproducible Results in Graphite-based Li-ion Full Coin Cells
Electrode materials, cathodes or anodes, are mostly evaluated in half coin cells where lithium (Li) metal is used. However, due to the instability nature of Li metal, the cycling performances of the half cells is more likely reflecting Li anode stability instead of the coupled electrodes, questioning the conclusions. Full coin cells utilizing graphite as the anode effectively assesses materials. But the key factors of preparing full coin cells with reproducible performances have rarely been discussed and overlooked leading to many inconsistent results. This work discusses the scientific gap between cell preparation and observed electrochemical performances to unveil the key parameters for achieving consistent and reliable results in full coin cells, critical for effective evaluation of any new materials tested under relevant conditions. The knowledge from this work is broadly applicable in other full cells systems such as Si-based full cells or Na-ion batteries.
Core–Shell Covalently Linked Graphitic Carbon Nitride–Melamine–Resorcinol–Formaldehyde Microsphere Polymers for Efficient Photocatalytic CO 2 Reduction to Methanol
Photocatalytic reduction of CO 2 with light and H 2 O to form CH 3 OH is a promising route to mitigate carbon emissions and climate changes. Although semiconducting metal oxides are potential photocatalysts for this reaction, low photon efficiency and leaching of environmental unfriendly, toxic metals limit their applicability. Here, we report a metal-free, core-shell photocatalysts consisting of graphitic carbon nitride (g-C 3 N 4 , CN) covalently linked to melamine-resorcinol-formaldehyde (MRF) microsphere polymers for this reaction. Cova-lent linkage enabled efficient separation of photo-generated carriers and photocatalysis. Using 100 mg of a photocatalyst containing 15 wt.% CN, a CH 3 OH yield of 0.99 μmol·h -1 was achieved at a reaction temperature of 80 °C and 0.5 MPa with external quantum efficiencies ranging from 5.5% at 380 nm to 1.7% at 550 nm. The yield was about 20 and 10 times higher than that of its components CN and MRF, respectively. Characterization with XPS, TEM, and bulk and surface elemental analyses supported a core-shell structure and charge transfer at C-N bond at the CN-MRF interface between the methoxy group in the 2,4-trishydroxylmethyl-1,3-diphenol part of MRF and the terminal amino groups in the CN. This enhanced ligand-to-ligand charge transfer resulted in 67% of photo-excited internal charge transferred from CN to hy-droxymethylamino group in MRF, whose amino group was the catalytic site for CO 2 photocatalytic reduction to CH 3 OH. Furthermore, this study provides a series of new metal-free photocatalyst designs and insights into the molecular-level structure-mediated photocatalytic response
Enabling Natural Graphite in High‐Voltage Aqueous Graphite || Zn Metal Dual‐Ion Batteries
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Curve fitting complex X-ray photoelectron spectra of graphite-supported copper nanoparticles using informed line shapes
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The impact of neutron irradiation, graphite oxidation and fluorination on tritium uptake into and desorption from graphite in molten salt environments
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The impact of fast spectrum graphite moderated systems on thermal spectrum graphite moderated reactors through the resonance physics of n+12C
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Roadmap of Graphite Moderator and Graphite-Matrix TRISO Fuel Management Options
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Assessing criticality implications of porosity misrepresentation in ENDF/B-VIII.1 graphite TSLs and small angle neutron scattering in nuclear graphite
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Examining Graphite Degradation in Molten Salt Environments: A Chemical, Physical, and Material Analysis
Molten-salt reactors (MSRs) are Generation IV nuclear reactors that use liquid salt as a coolant and/or fuel. In several MSR designs, graphite serves as a moderator and/or reflector. However, due to limited experimental data and operational experience, our understanding of graphite behavior in molten salt environments remains incomplete. This report aims to identify the degradation mechanisms of nuclear graphite in MSRs, detail the mechanisms of each factor, and provide an initial assessment of their impact on the structural integrity of graphite components. This assessment is based on an extensive literature review and insights from subject matter experts. Furthermore, given the limited data, a modeling strategy using existing Grizzly software is proposed for a more thorough analysis where appropriate. Additionally, it presents mitigation strategies where applicable. The report covers physical degradation mechanisms such as infiltration, erosion, and abrasion, as well as chemical degradation mechanisms including fluorination, intercalation, corrosion, and oxidation. Molten salt can infiltrate the porous structure of graphite, leading to several detrimental effects. Entrapment of fissile products within the graphite pores can cause radiation damage and could pose challenges in the handling and disposal of contaminated components. The differential thermal expansion between the infiltrated salt and graphite, along with internal stress from pressurized molten salt and volumetric heating, can compromise the structural integrity of graphite. To mitigate these effects, employing ultra-fine graphite grades and applying sealants and coatings are effective strategies. A computational model based on coupled solid mechanics and heat transfer phenomena could be used to predict the internal stresses using Grizzly software. In pebble-bed MSRs, graphite fuel pebbles can cause abrasion against reactor components due to friction and wear. The severity of wear is influenced by various factors such as temperature, environment, and the presence of lubricants. Tribological studies reveal that higher temperatures and molten salt environments, such as FLiBe, significantly reduce wear rates compared to dry conditions. Additionally, the chemical composition of the salt can further optimize graphite's tribological performance. Long-term wear effects can be modeled by incorporating surface defects into the geometry and predict stresses under thermal and radiation effects using Grizzly software. Chemical degradation of graphite in a molten salt environment can occur through fluorination and intercalation. Fluorination can occur via replacement of hydrogen or oxygen atoms, or at the active sites, but does not cause structural degradation. Intercalation, on the other hand, can lead to exfoliation, where layers of graphite separate and peel away, damaging the graphite. Protective coatings can enhance graphite's resistance to intercalation. Graphite generally exhibits good chemical stability in molten salt environments, though it can corrode under specific conditions, particularly in the presence of impurities or oxidants. Studies have shown that protective coatings, such as plasma-sprayed partially stabilized zirconia (PSZ), can effectively prevent such degradation. Corrosion behavior varies significantly with different graphite grades and coating applications, underscoring the need for detailed studies on uncoated and coated graphite to understand and mitigate corrosion mechanisms in MSRs. Research indicates that the presence of oxidants and impurities can accelerate graphite degradation in molten salts, making it essential to explore acceptable impurity limits. Oxidation is another critical degradation mechanism, leading to weight loss and structural damage due to the formation of CO and CO 2 from the reaction of carbon atoms with oxygen. This process creates new porosity and compromises graphite's integrity. While extensive research on graphite oxidation has been conducted for gas-cooled reactors, studies specific to MSRs are limited. Findings from the coal industry suggest that molten alkali metal salts can significantly accelerate graphite oxidation, a hypothesis worth exploring for fluoride salts in MSRs. Understanding oxidation behavior in MSRs is vital for developing protective measures. The analysis of post-irradiated graphite from the MSRE experiment demonstrated exceptional chemical compatibility with molten fluoride salt, suggesting that the extent of chemical attack on graphite largely depends on the salt's infiltration capability. Therefore, the use of ultra-fine grade graphite could help mitigate chemical degradation effects. Existing oxidation modeling capabilities in Grizzly, which use reaction-diffusion equations to model graphite-air interactions, could be adapted to simulate the chemical degradation effects of graphite in molten salt environments.
Computational Modeling of Graphite Degradation due to Molten Salt Infiltration and Wear
Molten-salt reactors (MSRs) represent a promising next-generation reactor design, with graphite serving as a moderator and/or reflector in several designs. However, due to limited experimental data and operational experience, a technical understanding of the structural integrity of graphite in molten salt environments remains incomplete. This report presents a modeling-based evaluation of graphite degradation in MSR environments, focusing on the effects of salt infiltration in fuel salt-based designs and surface wear in pebble bed reactor designs. The objective of this study is to enhance understanding of the structural integrity challenges posed by these degradation mechanisms and to provide a framework for assessing graphite behavior in MSRs. The first part of the report investigates the phenomenon of molten salt infiltration into graphite. This infiltration occurs when molten salt permeates the interconnected pore structure of the graphite moderator, driven by factors such as pressure differentials and the physical properties of both the salt and graphite. The infiltration process is influenced by characteristics of the pore structure, viscosity of the molten salt, and the interfacial energies between the graphite, salt, and the atmosphere within the graphite pore. Utilizing a coupled multiphysics modeling approach with Grizzly software, the study evaluates the stress induced by internal heat sources due to infiltration, which can lead to structural concerns. This evaluation is crucial for understanding how infiltration affects the mechanical integrity of graphite components in MSRs. The study considers the Molten-Salt Reactor Experiment (MSRE) graphite stringer geometry due to the availability of relevant data. Through detailed finite element analysis, the study examines stress distributions at varying infiltration percentages, revealing that stress levels increase with higher amounts of infiltration. Rare-event simulations, using the parallel subset simulation (PSS) framework, further quantify the failure probabilities under input uncertainties, with a user-specified failure metric. The PSS framework also identifies critical input parameters that significantly affect the stress values, including infiltration amount, thermal conductivity, and power density. Additionally, considering realistic reactor scenarios, the analysis was performed to account for the combined effects of radiation and infiltration, and modeling strategies on how to analyze new reactor designs or new graphite grades are discussed. The second part of the report focuses on wear mechanisms in pebble bed-based MSRs. As graphite fuel pebbles interact with the graphite reflector block, wear can result in material loss and the formation of surface defects, which may act as stress concentrators. A similar multiphysics modeling framework is employed to assess the impact of wear on the structural integrity of graphite components. This study considers a generic fluoride-cooled high-temperature reactor (gFHR) design due to the availability of comprehensive data. Worst-case scenario dimensions of the reflector blocks were analyzed under thermal and radiation conditions. Subsequently, wear in the form of idealized pits and grooves is modeled on the inner surface of the graphite block, with the maximum stress from previous simulations. The simulations show that groove-type defects are more detrimental than pits, leading to higher stress concentrations. Considering worst-case simulation scenarios and experimental wear rates, it was determined that the formation of a surface defect critical enough to affect the stress may not be possible in a gFHR design. Overall, the findings of this research contribute to the development of robust modeling tools for predicting graphite behavior under various operational conditions in MSRs.
Progress Report on Graphite-Salt Intrusion Studies
The document reports on the design, construction, testing and initial operation of the ORNL graphite intrusion system with pressurized molten salts. This technical memorandum is responsive to work package AT-20OR04060203, “Graphite-Salt Studies” —ORNL”, and fulfills Milestone M3AT20OR040602035 “Report on graphite-salt intrusion studies”.The report outlines the procedure of graphite intrusion experiments, which follows the ASTM standard guidance, and the procedures of other characterization measurements of graphite physical properties, also based on ASTM standard methods. A set of 12 graphite grades of various origin and properties went through preliminary characterizations, and six of them were further selected for intrusion experiments with molten FLiNaK. The results are discussed in the context of graphite physical properties and are compared with a several results from literature. The correlation between molten salt impregnation and mercury intrusion, first proposed by MSRE scientists, is discussed in detail. Pending collection of more data, the mercury intrusion porosimetry technique may possibly be used as a rapid method for screening and down selection of graphite grades with extremely low salt intrusion, as required for operation in molten salt reactors. This new system will generate data non only on intrusion behavior, but will also generate a wide range of salt- impregnated graphite samples that will be further analyzed with advanced characterization techniques for better understanding of the effect of salt intrusion on the properties of nuclear graphite. It expected that these results and future results obtained with this system will help advance our understanding of the factors controlling the extent of salt penetration in graphite and of the relationship with intrinsic structural properties of various grades. Together, correlating direct measurements results and graphite structural properties should allow development of a predictable model that could be used for selection and qualification of optimal graphite grades, tailored to design operation conditions and molten salt properties.
Oxidation Behavior and Property Degradation of Nuclear Graphites
During its multidecade operation in the core of nuclear reactors, graphite components are subjected to aggressive and continuous exposure to a high field of ionizing and neutron irradiation, high temperature, and various types of present and postulated chemical attacks. High density, high crystallinity polygranular synthetic graphite is unique among other materials for its extraordinary capacity of resisting and adapting to the aggression inflicted by high temperature, high energy neutron bombardment and ionizing gamma radiation. But, as a carbonaceous material, even though of very high purity, graphite is reactive towards common oxidizing agents: oxygen, carbon dioxide, water. Safe operation of HTGRs relies, among other aspects, on engineered safeguard systems for efficient and continuous protection of graphite components against oxidation. Graphite oxidation behavior was, and continues to be, an important direction of theoretical and experimental research, engineering analyses, models and simulations, and design and safety regulations. The avalanche of publications, reports, experimental data, computer codes, and regulatory documents related to oxidation behavior of nuclear graphite is now accelerating to new levels, prompted by the increased interest for nuclear energy as a clean, carbon-free energy source. Even though public’s perception of nuclear energy advantages may still be influenced by the memories of past accidents of nuclear reactors from generations II and III, the community of informed scientists and engineers, regulators and statemen knows that generation IV of nuclear reactors is designed at very high safety standards, doubled by great advances of scientific knowledge and technological progress. One of routes of these recent advances is directed at better understanding of graphite oxidation behavior, its relationship with graphite manufacturing and microstructural properties, along with the effects of various environmental factors and process variables. Together, the recent progress in manufacturing, properties characterization, and modeling of intricated physical and chemical processes that concur to the oxidation behavior led to development of powerful simulation codes able to analyze various scenarios of normal operation and hypothetical off-normal events, and thus to clearly specify the allowable parameters envelopes for the designers, constructors, and operators of current and future modular HTGRs. This review begins with an introduction on manufacturing methods, structure, and properties of nuclear graphite, including basic requirements that this specialty graphite type must satisfy for nuclear use. It continues with a chapter on environmental effects on nuclear graphite, where emphasis is placed less on irradiation and much more on oxidation phenomena, their safety implications, and the basic traits of chronic and acute oxidation by air (oxygen) and water (humidity, steam). Particular attention is placed on the three graphite grades of interest for this document (IG-110, NBG-18, PCEA). A chapter on properties degradation induced by oxidation follows, with focus on density, dimensional, and mechanical properties changes. The next chapter is intended as a brief review of various approaches used for modeling of graphite oxidation behavior. It summarizes the progress of oxidation models, from the early attempts to complex computational approaches interfaced with specialized computer codes designed for nuclear reactor simulations. Last, a list is presented of knowledge gaps where more research is needed. A short summary concludes the review.
Determining the oxidation behavior of matrix graphite
This work presents the oxidation behavior of matrix graphite in air. Matrix graphite, graphite powder/flakes bonded by a small amount of non-graphitic carbon, surrounds coated fuel particles in order to form cylindrical fuel compacts (in prismatic core designs) or spheres (in pebble-bed reactor designs). This work focuses on oxidation tests conducted on two matrix graphite materials, one provided by Kairos Power and the other A3 matrix graphite. Some of the tests followed American Society for Testing and Materials (ASTM) oxidation testing standards using a vertical furnace system and others were performed in a thermogravimetric analyzer (TGA). It was determined that, at temperatures of 450 °C–700 °C, the oxidation rate of the Kairos matrix graphite follows the Arrhenius equation. In comparison with A3 matrix graphite, the Kairos matrix graphite shows better oxidation resistance at high temperatures (≥550 °C), but also a higher oxidation rate at low temperatures. Both the A3 matrix graphite and the Kairos matrix graphite materials may experience preferential oxidation of the partially graphitized binder. An oxygen penetration gradient was also observed when using the three characterization methods (i.e., optical microscope, x-ray tomography [XCT], and density profile by the lathe) enlisted in this research. In conclusion, the oxygen penetration depth increases with decreasing isothermal oxidation temperature, while the center of the oxidized samples (10% weight loss) remains almost untouched even at 500 °C.
GIF-VHTR Materials (Graphite) Project Plan: Progress since September 2021
¦1: Data, Design Methodology and Construction 1.1: Graphite Selection and Acquisition Strategy (Original tasks completed) 1.2: Graphite Properties (Virgin Materials) 1.3: Graphite Fracture Behaviour(Original tasks completed) 1.4: Graphite Oxidation Behaviour(Original tasks completed) 1.5: Graphite Component Testing 1.6: Graphite Irradiation Effects 1.7: Graphite Irradiation Induced Creep 1.8: Graphite Codes and Standards Development (ASME & ASTM) 1.9: Graphite BehaviourModels Development 1.10: Links to Existing Graphite Irradiation BehaviourDatabases 1.11: Materials Database ¦2: Operation and Inspection 2.1: Methodologies for determining requirements for inspection of graphite cores 2.2: Methodologies for ISI of graphite cores ¦3: Decommissioning and Disposal (New task added 2021) 3.1: Technical data for decommissioning and disposal issues
Application of the Multi-Species, Multi-Reaction Model to Coal-Derived Graphite for Lithium-Ion Batteries
Graphite is a critical material used as the negative electrode in lithium-ion batteries. Both natural and synthetic graphites are utilized, with the latter obtained from a range of carbon raw materials. In this paper, efforts to synthesize graphite from coal as a domestic feedstock for synthetic graphite are reported. Domestic coal-derived graphite could address national security and energy issues by standing up domestic supply chains for battery critical materials. The performance in lithium-ion coin cells of this coal derived graphite is compared to a commercial battery-grade graphite. For the first time, a multi-species, multi-reaction (MSMR) modeling technique is applied to synthetic graphite derived from coal. Key thermodynamic, transport, and kinetic parameters are obtained for the coal derived graphite and compared to the same parameters for commercial battery-grade graphite. Modeling of synthetic graphites will allow for virtual evaluation of these materials toward production of domestically sourced graphite.
Characterization of Urbix Resources Natural Flake Graphite Material for Nuclear Applications (CRADA NFE-18-07340 Final Report)
Urbix Resources is a company specializing in graphite powder processing and purification. Urbix is looking to become the first domestic supplier of nuclear-grade graphite in the United States. This has led to a Cooperative Research and Development Agreement (CRADA) being established between the Oak Ridge National Laboratory (ORNL) and Urbix Resources, with the goal of characterizing a purified graphite feedstock supplied by Urbix Resources. The resulting data will then be used to determine the viability of a large-scale nuclear graphite production and qualification campaign. These characterization methods have been performed on the as-received graphite feedstock, a heat-treated batch of the feedstock, a milled batch of the feedstock, and a nuclear-grade graphite powder historically used in fuel production. The data from the heat-treated and milled feedstock will determine how different processing methods affect the powder’s physical properties, whereas the data from the nuclear grade graphite will provide information about the characteristics of a graphite previously accepted for nuclear applications. All of these powders have been characterized individually and as part of a resin-graphite blend (herein referred to as matrix) that has been used in past fuel experiments. The results of these experiments show that the Urbix powder and nuclear-grade graphite powder display notably different properties from one another in many cases, but this discrepancy may have been the result of the typical particle size of each powder. This hypothesis is supported by tests conducted on the milled powder, which showed a closer resemblance in properties to the nuclear-grade graphite than the as-received powder. The method for pressing the graphite matrix also had a noticeable effect on the measured properties.