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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 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Report on Initial Tribological Studies of Graphite in Dry Argon and Molten Salt Environment

This report documents completion of the Advanced Reactor Technologies (ART) Level 2 Milestone M2TG-24OR0501081: “Complete report on initial tribological studies within molten salt environment”, due January 31, 2024. This milestone is a carryover milestone from FY23 M2TG-23OR0501091. The report summarizes the initial studies on tribological characterization of the graphite–graphite sliding interface in a dry argon and molten FLiNaK salt environment. The experiments were conducted on a high-temperature multifunctional tribometer placed in a glovebox. This configuration enables probing the wear and friction properties at high temperatures in an inert environment. The worn surfaces were analyzed using high-resolution microscopy and white-light profilometry to determine the wear modes and wear rates. This report discusses the previous wear and friction studies on 316H stainless steel (SS) sliding against graphite in molten salt for different temperatures, sliding speeds, and salt amount. Moreover, key tribological parameters of pebble-on-pebble interaction in gas-cooled and molten salt reactors (MSRs) are analyzed and used to design an experimental test matrix for future studies. The outcomes of this project could provide key information about the tribological behavior of graphite pebbles in gas-cooled and MSRs, thereby contributing to safer operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

NDMAS System and Process Description

The U. S. Department of Energy (DOE) has made a significant investment in research to develop the next generation of reactor technologies as well as to improve the performance and lengthen the life cycle of existing nuclear reactors. Data collected to demonstrate new concepts may also be used in the future to support licensing of these technologies. Provenance of these data must be preserved. The Nuclear Data Management and Analysis System (NDMAS) was established to manage and preserve data collected by fuels and materials research conducted by the high-temperature, gas-cooled reactor program. The scope of NDMAS is expanding to include other nuclear research programs that have the shared need to preserve the provenance of research data. Nuclear research funded by DOE is conducted by Idaho National Laboratory (INL), universities, other national laboratories, foreign research partners, and private companies. This research will generate a large amount of data from a variety of sources over a period of many years. Managing the data generated by the research and development projects presents a significant challenge for retaining data integrity and availability.

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DOE OSTI2022

Advanced Graphite Creep Uncertainty Analysis

Radiation damage estimation is an important component of the post irradiation analysis of the Advanced Graphite Creep (AGC) experiment. It depends primarily on the fast fluence, which is determined using well established methods of spectral adjustment. These are based on best estimates from models such as Monte Carlo N-Particle (MCNP), input cross-sections, and measured activities from flux wires in the experiment. Each of these parameters can propagate uncertainties which will affect the uncertainty in the calculated dose levels for AGC, or any experiment irradiated within a reactor. While the methods of propagating uncertainty are well-established, the final uncertainty estimates they provide are only as good as the estimates of uncertainty in the inputs on which they are based. The purpose of this work is to outline some deficiencies in the ways these input uncertainties are presently estimated, and to outline a methodology by which they can be improved. The fast fluence and radiation damage received by graphite specimens irradiated in the Advanced Graphite Creep (AGC) experiments is presently estimated using spectral adjustment methods that are based on both flux wire activity measurements, and MCNP model predictions. This work describes an ongoing effort to quantify and propagate uncertainties in inputs to the spectral adjustment process, and thereby quantify the resultant error in radiation damage (dpa) estimates. The effort is multi-faceted, and we consider the impacts of both the set of flux wires selected, and the counting process. An expanded set of flux wires is identified that provides a more comprehensive data set on the fast spectrum. To address the counting process itself, a series of round-robin measurements in several reactor metrology laboratories across the Department of Energy (DOE) complex and nuclear industry are being undertaken to refine the American Society for Testing and Materials (ASTM) standards for flux wire measurements. To address the contribution of uncertainty in the MCNP model predictions, an uncertainty quantification (UQ) tool has been developed that statistically samples the model input parameters, runs a series of cases, and assimilates the results to provide an overall uncertainty. The impact of the MCNP UQ tool results is demonstrated by re-analyzing previous AGC flux wire and irradiation data. While the expanded flux wire set obviously cannot be added to these experiments retroactively, plans for future graphite irradiations are outlined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

SAM Code Development for Source Term Modeling in Fluoride-salt-cooled High-temperature Reactors

The SAM code is under development and supported by DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. These advanced reactor concepts incorporate novel and improved approaches to achieve safety and economic feasibility. Many concepts leverage higher operating temperatures for improved efficiency with a variety of coolants and structural materials to support those needs. Such design choices may also pose unique radiological source term risks. There are continuous efforts to incorporate new physics and phenomena relevant to advanced reactor concepts, including capabilities to enable system-level source term modeling. The Pebble-Bed Fluoride-salt-cooled High-temperature Reactor (PB-FHR) is a promising candidate among advanced nuclear reactor concepts with its improved passive safety characteristics and high thermal efficiency. In addition to past efforts to support the development and utilization of SAM for PB-FHR safety analysis, the species transport modeling capabilities in SAM have been extended to simulate source term phenomena in concepts like the PB-FHR. The PB-FHR concept utilizes pebble-form TRISO fuel and FLiBe salt coolant to provide robust barriers to the release of almost all fission products and radiological source terms. However, tritium poses a unique radiological risk due to its significant production from neutron interactions with 6 Li and 9 Be in the FLiBe as well as its high mobility at elevated temperatures where it can permeate through structural metals. Secondly, the use of graphite at elevated temperatures poses the risk of oxidation damage to any structural components or fuel pebbles that are exposed to accidental air ingress. This report summarizes progress made in modeling these phenomena in SAM, which leverages the system-level multi-physics thermal hydraulic simulation to support an effective engineering-scale source term modeling capability. A tritium transport model is developed to simulate the various transport pathways in FHR and MSR concepts and is presented through verification and validation examples as well as demonstrations simulating experimental test loops and reference plant FHR models. A graphite oxidation model is also introduced to calculate local oxidation rates and is presented with initial validation results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Nuclear Data Impact Assessment for the HTR-10 Pebble-Bed Reactor Using SCALE

The HTR-10 was used as a representative pebble-bed high-temperature gas-cooled reactor in this assessment of nuclear data’s impact on important reactor and spent fuel metrics, including safety-related quantities such as the effective multiplication factor (k eff ), temperature reactivity feedback, spent fuel inventory, and decay heat. Using the SCALE code system tools and ENDF/B-VII.1 nuclear data libraries, we quantify the effect of nuclear data uncertainties on these key performance metrics for both fresh fuel and equilibrium core configurations. For reactor core key parameters, important contributors to uncertainty include reactions of 235 U [$\bar{v}$, fission, (n, γ)], 238 U [elastic, (n, γ)], and graphite [elastic, (n, γ)]. Additional important contributors for the equilibrium core include reactions of higher actinides ( 239 Pu, 240 Pu) and fission products ( 135 Xe, 149 Sm). For spent fuel analysis, most nuclide inventory uncertainties remain below 5%. Higher uncertainties up to 11% are being observed for minor actinides like 243 Am and 244 Cm. Additionally, fission product uncertainties in 155 Eu and 155 Gd, of 25% and 23% respectively, are also significant and have implications for burnup credit applications. 110m Ag also shows high uncertainty of up to 11%, mainly due to fission product yield uncertainties. Decay heat relative uncertainties remain below 0.6% up to 10 years’ cooling time after fuel discharge. The highest relative uncertainty of 1.5% occurs at 500 years of cooling; however, because the decay heat value is very low at that time, the absolute uncertainty is not significant. This work demonstrates that extending assessments beyond fresh fuel k eff to include irradiated cores, nuclide inventories, and decay heat is essential in understanding the behavior of uncertainties as a function of fuel burnup and can support improvements of safety margins and spent fuel management.

Nuclear data impact↗
DOE OSTI2024

Operando NMR characterization of cycled and calendar aged nanoparticulate silicon anodes for Li-ion batteries

Replacing graphite anodes with Si anodes can greatly increase the energy of current Li-ion batteries. Detailed characterization of Si lithiation reactions, SEI formation, and reversibility are therefore active areas of research. Solid-state 7 Li nuclear magnetic resonance (NMR) spectroscopy is useful for characterizing different lithium local environments within Si anodes. Here, we developed an operando NMR methodology to characterize aging of carbon-coated nanoparticulate Si anodes in pouch cells paired with Ni-rich cathodes. We observed a new lithiation mechanism in the Si nanoparticles: direct formation of over-lithiated Li 15+x Si 4 (x<0.6) phase. Furthermore, our novel operando cells maintained good performance with long-term cycle and calendar aging. Here we identified trapped lithium silicides as a major contributor to capacity fade with aging. Finally, we determined that the addition of Mg (TFSI) 2 to the electrolyte decreased the amount of trapped lithium silicides and therefore increased the capacity and capacity retention for the nanoparticulate Si used.

25 ENERGY STORAGE↗
DOE OSTI2020

Carburization of metals by a chemical mechanism of carbon transport through molten fluoride salts

Carburization of structural alloys used in Fluoride Cooled High Temperature Reactors (FHR) can result from the interaction between the metal and graphite components and cause changes in alloy properties. However, carburization requires transport of carbon from graphite to alloy surfaces through the molten fluoride coolant. This work introduces an alternative to previously proposed carbon transport mechanisms. The new mechanism is based on the generation of carbonate ions (CO 3 2- ) by oxide impurities in the salt reacting with graphite. CO 3 2- is then reduced on metal surfaces to deposit carbon. This work demonstrates the reduction of dissolved CO 3 2- on pure Fe, Ni, Cr, Mo, and W surfaces without an applied potential in molten FLiNaK at 700 °C. Finally, we also demonstrate that the interaction between oxide and graphite in molten FLiNaK media produces CO 3 2- .

36 MATERIALS SCIENCE↗
DOE OSTI2024

Characterization of fluidized bed chemical vapor deposition ZrC coatings on PyC/YSZ kernels deposited under differing conditions

In this study, coated fuel particle architectures with ZrC coatings are candidate fuels for advanced power reactors and space nuclear propulsion (SNP) concepts. Owing to its relevance to SNP, the composition, microstructure, and mechanical properties of eight ZrC coatings prepared by fluidized bed chemical vapor deposition were evaluated. Evaluation by SEM and EBSD showed that all grains were columnar. Across the various examined samples, minor axis diameters varied between 0.3 and 1.1 μm, and major axis diameters varied between 0.4 and 2.3 μm. Major and minor diameters increased with thickness particularly at higher deposition temperatures in which the major grain axis (from an ellipse fit to the grain shape) increased by 2.5 μm over the entire coating. Coatings with higher reactive gas flows and Zr/C concentrations closer to 1 were observed to contain nanocrystalline graphite deposits. Reactive gas flow doubling led to increases in coating thickness from around 10–15 μm to around 22–27 μm.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

INL Report: Multiphysics Modeling and Simulation of Deimos, an Advanced Reactor Experiment

When designing a novel reactor, nuclear experiments are essential to validate the predictive capability of modeling/simulation tools and to justify the investment of a full-scale prototype. Being able to accurately predict temperature reactivity coefficients is of high importance to reactor designers as it impacts the safety and performance of the system. Deimos is a proposed graphite-moderated, beryllium-reflected, high assay low-enriched uranium (HALEU) tri-structural isotropic (TRISO) fueled experiment for NCERC. Deimos is a valuable experiment for validating the predictive performance of modeling and simulation tools since it comprises a relatively thermal energy spectrum with 84% thermal neutrons causing fission, 15% epithermal neutrons causing fission, and 1% fast neutrons causing fission. Additionally, by electrically heating the experiment, reactivity can be validated at various temperatures. Deimos will serve as a testbed for future advanced reactor concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Implementation of TSL Evaluations Beyond the Incoherent Approximation [Slides]

This presentation discusses the motivation behind the project, which is to facilitate the production of TSL libraries that contain high fidelity physics including coherent inelastic effects, such as graphite and beryllium. High fidelity TSL libraries are of interest in various applications, for example in relation to neutron scattering instrument response analysis and directional biases (reflection) in neutronic analysis. In summation, the removal of typical approximations, in terms of incoherent, cubic, and atom site are examined. The application of generalized equations to crystalline graphite and beryllium TSL evaluations and the validation of data and data formats is also discussed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Probing the Reactivity of the Active Material of a Li-Ion Silicon Anode with Common Battery Solvents

Calculations and modeling have shown that replacing the traditional graphite anode with silicon can greatly improve the energy density of lithium-ion batteries. However, the large volume change of silicon particles and high reactivity of lithiated silicon when in contact with the electrolyte lead to rapid capacity fading during charging/discharging processes. In this report, we use specific lithium silicides (LS) as model compounds to systematically study the reaction between lithiated Si and different electrolyte solvents, which provides a powerful platform to deconvolute and evaluate the degradation of various organic solvents in contact with the active lithiated Si-electrode surface after lithiation. Nuclear Magnetic Resonance (NMR) characterization results show that a cyclic carbonate such as ethylene carbonate is chemically less stable than a linear carbonate such as ethylmethyl carbonate, fluoroethylene carbonate, and triglyme as they are found to be more stable when mixed with LS model compounds. Furthermore, guided by the experimental results, two ethylene carbonate (EC)-free electrolytes are studied, and the electrochemical results show improvements with graphite-free Si electrodes relative to the traditional ethylene-carbonate-based electrolytes. More importantly, the study contributes to our understanding of the significant fundamental chemical and electrochemical stability differences between silicon and traditional graphite lithium-ion battery (LIB) anodes and suggests a focused development of electrolytes with specific chemical stability vs lithiated silicon which can passivate the surface more effectively.

25 ENERGY STORAGE↗
DOE OSTI2020

Generation IV Benchmarking of TRISO Fuel Performance Models Under Accident Conditions Final Report

The Generation IV International Forum (GIF) is a co-operative international endeavor of fourteen members organized to carry out the research and development needed to establish the feasibility and performance capabilities of the next generation nuclear energy systems. GIF selected six reactor technologies, amongst which is the Very High Temperature Reactor (VHTR) that is primarily dedicated to the cogeneration of electricity and hydrogen. The technical basis for VHTR is the tristructural isotropic (TRISO)-coated particle fuel, the graphite as the core structure, helium coolant, as well as the dedicated core layout and lower power density to removal decay heat in a natural way. At the heart of safety features of the VHTR concept lie the TRISO fuel particles that are designed to keep their structural integrity and retain fission products at temperatures up to 1600°C. As part as the design and future operation of VHTRs, a key aspect is the accurate prediction of fuel performance under irradiation and accident conditions. Modeling and simulation allow prediction of TRISO fuel behavior when subject to neutron flux and in high temperature accident scenarios. The refinement of the fuel performance models and codes is performed by comparison to in-pile and out-of-pile experimental data that reproduce the expected irradiation conditions in high temperature gas-cooled reactors (HTGRs). Historically, the International Atomic Energy Agency (IAEA) developed a benchmark dedicated to the validation of predictive methods for fuel and fission product behavior through the Coordinated Research Program CRP-2 (IAEA, 1997). CRP-2 was later updated to cover fuel fabrication, quality assurance, irradiation performance, safety testing, and spent fuel. The scope of the resulting CRP-6 benchmarks focused on HTGR fuel performance and fission product release (IAEA, 2012). Taking advantage of additional TRISO fuel fabrication, irradiation, and safety testing campaigns, GIF launched a Generation IV Benchmarking of TRISO Fuel Performance Models under Accident Conditions in late 2015. This GIF benchmark is a three-year program steered by Idaho National Laboratory (INL, USA). The other participants include the Japan Atomic Energy Agency (JAEA, Japan) and the Korea Atomic Energy Research Institute (KAERI, Korea). The objectives of the benchmark are to: follow on the IAEA CRP benchmarks, (2) model fission product release under accident conditions, (3) compare results obtained by the fuel performance modeling codes of the benchmark participants, and (4) compare these code predictions to experimental data. Safety tests chosen for modeling include the first and second experiments of the Advanced Gas Reactor program (AGR-1 and AGR-2) and the High Flux Reactor (HFR) EU1bis experiment. This report presents the results obtained by the three research institutions using their respective fuel performance modeling codes. Comparisons of the corresponding fission product release predictions are made with experimental data from AGR-1, AGR-2, and HFR-EU1bis. The benchmark results show good agreements between all participants but also show a general trend of over-prediction of the experimental release data, which is mainly attributed to the use of over-estimated diffusion coefficients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Measurement of the mass-changing, charge-changing, and production cross sections of 11 C, 11 B, and 10 B nuclei in 12 C + p interactions at 13.5 GeV/𝑐 per nucleon

A good knowledge of nuclear fragmentation cross sections is important to interpret the fluxes of secondary cosmic rays from the Galaxy. Here we report new measurements of nuclear fragmentation with the NA61/SHINE experiment at the CERN SPS. The specific focus is on cross sections important for the production of boron in the Galaxy from the interactions of 12 C nuclei with hydrogen in the interstellar medium, including the contribution from the decay of the short-lived 11 C fragments. The data were taken with the secondary 12 C beam at beam momentum of 13.5 GeV/𝑐per nucleon and two fixed targets, polyethylene (CH 2 ) and graphite (C), from which we derive the cross sections of carbon on hydrogen. We present the measurement of the fragmentation cross sections of 11 C, 11 B, and 10 B as well as the mass- and charge-changing cross sections.

nuclear fragmentation↗
DOE OSTI2022

Materials Scorecards, Phase 2: Advanced Materials and Manufacturing Technology

This report presents updated material score cards with detailed justification of the proposed rating provided and quantifies the readiness of the advanced manufacturing of materials for nuclear applications. Based on an expanded literature review of available data, the scores in this report (Phase 2) have either been modified or remained constant as compared with Phase 1 scoring. The rating justification are provided for each scoring criteria whereas the maturity quantification of additively manufactured alloys and the knowledge gaps are discussed with reference traceability. This report focuses on the detailed justification and traceability for the adoption of additively manufactured 316SS, SS304, Alloy 800H, Graphite C/C, Alloy N, Silicon Carbide, HT9, Alloy 617, and Alloy 718. Significant variability in data and knowledge gaps exist among the identified materials. The scorecards are therefore based on published literature data, industry response to a survey, stakeholder input collected at workshops, and expert opinion. The interpretation and assessment were somewhat challenged as experimental process parameters and methodology varied and best judgment was exercised for score quantification. This assessment reflects the overall level of technical maturity for AM materials and their deployment in Gen-VI nuclear systems. These Phase 2 scorecards are likely to be revised based on additional input from stakeholders and as new research results become available.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Valorizing the carbon byproduct of methane pyrolysis in batteries

While low-cost natural gas remains abundant, the energy content of this fuel can be utilized without greenhouse gas emissions through the production of molecular hydrogen and solid carbon via methane pyrolysis. In the absence of a carbon tax, methane pyrolysis is not economically competitive with current hydrogen production methods unless the carbon byproducts can be valorized. In this work, we assess the viability of the carbon byproduct produced from methane pyrolysis in molten salts as high-value-added anode or conductive additive for secondary Li-ion and Na-ion batteries. Raman characterization and electrochemical differential capacity analysis demonstrate that the use of molten salt mixtures with catalytically-active FeCl 3 - or MnCl 2 result in more graphitic carbon co-products. These graphitic carbons exhibit the best electrochemical performance (up to 272 mAh/g of reversible capacity) when used as Li-ion anodes. For all carbon samples studied here, disordered carbon domains and retained salt species trapped and/or intercalated into the carbon structure were identified by X-ray photoelectron and multinuclear solid-state nuclear magnetic resonance spectroscopy. The latter lead to reduced electrochemical activity and reversibility, and poorer rate performance compared to commercial carbon anodes. The electronic conductivity of the pyrolyzed carbons is found to be highly dependent on their purity, with the purest carbon exhibiting an electronic conductivity nearly on par with that of commercial carbon additives. These findings suggest that more effective removal of the salt catalyst could enable applications of these carbons in secondary batteries, providing a financial incentive for the large-scale implementation of methane pyrolysis for “low-carbon” hydrogen production.

08 HYDROGEN↗
DOE OSTI2026

Experiments towards a neutron target for measurements in inverse kinematics

Neutron-induced reactions play an important role in fundamental nuclear physics, nuclear astrophysics, and applications. In the case of reactions on rare isotopes, there are limited options for direct experimental measurements. The Neutron Target Demonstrator project at Los Alamos National Laboratory seeks to test the feasibility of moderating spallation neutrons within a 1 m 3 graphite cube to create a standing neutron target for neutron-induced reaction measurements in inverse kinematics. This paper presents the results of experimental neutron flux distribution tests using neutron sources (ranging from 1 keV to 50 MeV) created by accelerators at the University of Notre Dame and Texas A&M University. Measurements were made with both the full graphite cube as well as a ”half cube” setup in which half of the graphite cube was removed. The measured distributions agree with simulated distributions in the case of the full cube moderator, although there remain discrepancies in certain cases for the half cube moderator. The results shown here will provide useful information for an upcoming experimental campaign to test the neutron target proof-of-principle.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗