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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 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
NASA NTRS1985

Dual-ion-beam deposition of carbon films with diamond-like properties

A single and dual ion beam system was used to generate amorphous carbon films with diamond like properties. A methane/argon mixture at a molar ratio of 0.28 was ionized in the low pressure discharge chamber of a 30-cm-diameter ion source. A second ion source, 8 cm in diameter was used to direct a beam of 600 eV Argon ions on the substrates (fused silica or silicon) while the deposition from the 30-cm ion source was taking place. Nuclear reaction and combustion analysis indicate H/C ratios for the films to be 1.00. This high value of H/C, it is felt, allowed the films to have good transmittance. The films were impervious to reagents which dissolve graphitic and polymeric carbon structures. Although the measured density of the films was approximately 1.8 gm/cu cm, a value lower than diamond, the films exhibited other properties that were relatively close to diamond. These films were compared with diamond like films generated by sputtering a graphite target.

Mirtich, M. J.↗
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↗
NASA NTRS1993

Precision linear shaped charge severance of graphite-epoxy materials

This paper presents Precision Linear Shaped Charge (PLSC) components designed to sever a variety of target materials. Recent data for the severance of graphite-epoxy panels or targets with PLSC's are presented. A brief history of the requirement to originate the development of PLSC's for weapon components at Sandia National Laboratories is presented. The Department of Energy's (DOE) nuclear weapon systems have continually decreased in size. Today's relatively small weapons require the design of much more efficient, lighter, and smaller explosive components because fragments, air shocks, and pyro-shocks associated with the function of these components can damage electrical and other sensitive components located nearby. The DOE requirements for PLSC's are listed. Therefore, linear shaped charge (LSC) components for weapon systems can no longer be empirically or experimentally designed for a given application. Many of today's designs require severing concentric cylinders, for example, where the LSC jet is designed to sever only one of the two cylinders as was the case for the B90/Nuclear Depth Strike Bomb. Therefore, code modeling and simulation technology must be utilized to obtain a better understanding of the LSC jet hydrodynamic penetration, fracture, shear, and spall mechanisms associated with the severance of metallic as well as composite targets.

Vigil, Manuel G.↗
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↗
NASA NTRS1980

Positron annihilation studies of moisture in graphite-reinforced composites

The positron lifetime technique of monitoring absorbed moisture is applied to several composites, including graphite/polymides which are candidates for high-temperature (over 260 C) applications. The experimental setup is a conventional fast-slow coincidence system wherein the positron lifetime is measured with respect to a reference time determined by the detection of a nuclear gamma ray emitted simultaneously with the positron. From the experiments, a rate of change of positron mean lifetime per unit mass of water can be determined for each type of specimen. Positron lifetime spectra are presented for a graphite/polyimide composite and for a pure polyimide.

Singh, J. J.↗
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↗
DOE OSTI2020

ASME Code Activities

U.S. Nuclear Regulatory Commission (NRC) is currently assessing ASME Section III, Division 5 (2017 Edition) and accompanying Code Cases for endorsement Draft Regulatory Guide (DG-1380) and draft NUREG (2245) on the endorsement of Division 5 were issued and public comment phase completed DG-1380 places conditions in the areas of general requirements, mechanical design, metallic materials allowable stresses, and graphite materials/design of Division 5 DG-1380 is a revision to Regulatory Guide 1.87, Revision 1, 1975 (which endorsed the 159X series of ASME Code Cases, with conditions) NRC staff is currently reviewing the low-temperature and elevated temperature Alloy 617 code cases for endorsement Results of the review will be incorporated into DG-1380 and changes will be announced publicly The formal endorsement of Section III, Division 5 and accompanying code cases, with conditions, will be made through Regulatory Guide 1.87, Revision 2

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Concepts for Actinide Recovery from TRISO Used Nuclear Fuel

The work described in this report has developed a preliminary conceptual flowsheet based on limited published literature for the head-end of a TRISO oxycarbide UNF reprocessing plant. The essential function of the head-end is to prepare the UNF for subsequent separation of actinides. The main steps of the conceptual TRISO UNF head-end reprocessing flowsheet are: 1. Fragmentation of the graphite moderator to expose the fuel particles. 2. Separation of fuel particles from fragmented graphite moderator by sieving and fluidization. 3. Fragmentation of the fuel particles to expose the fuel kernel. 4. Dissolution of the fuel kernel in nitric acid. Noteworthy at this step is the formation of organic acids (mainly oxalic and mellitic acids) from the carbon associated with the fuel and TRISO coatings. These acids can interfere with actinide separations and solvent extraction hydraulic performance and carbon dioxide gas generation. 5. Clarification of the dissolved fuel solution to separate the coating fragments. 6. Dissolver off-gas treatment with specific emphasis on managing generation of carbon dioxide gas containing carbon-14. The quantity of carbon dioxide is projected to be 100 to 1000 times greater than that arising from dissolving LWR UNF. Pre or post treatment of the exposed fuel kernels could be undertaken to mitigate the formation of organic acids but there still exists a significant off-gas treatment challenge. The literature on reprocessing uranium nitride UNF was reviewed but a flowsheet was not specifically developed since its scope and uncertainties are expected to be encompassed by the oxycarbide flowsheet. Overall, the predominant feature of processing TRISO UNF is carbon management to reduce waste volume and mitigate the formation of carbon dioxide and organic acids. This report recommends broad areas of research that will help to further define the head-end flowsheet.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Nuclear fuels for transient test reactors

Transient test reactors with the ability to test fissile specimens under extreme conditions have been crucial tools in the development of nuclear technologies. Less than 10 unique facility designs have ever been constructed, most of which remain operational today and still use the original nuclear fuel constructed for them more than 40 years ago. Historic fuel systems for transient test reactors vary in significant ways which have marked influences on reactor capabilities. Eventually, new fuel will be needed to support the longevity of transient test reactor missions. This paper reviews precedent transient reactor fuel systems in the context of their unique requirements. A few key conclusions are illustrated by comparing and contrasting these transient test reactors. Fuel composites which are mostly graphite can enable transient reactors with very high neutron fluence capability (>2E16 n/cm 2 ) and are amenable to longer “shaped” transients but cannot achieve pulses <10 ms in duration. Reducing the graphite-to-uranium ratio can yield a very narrow pulse capability but delivers less fluence and requires cores with considerably more fissile material. Designs based on uranium dioxide (UO 2 ) make use of readily available materials to create compact cores with narrow pulse width capabilities but with moderate neutron fluence capabilities (~2E15 n/cm 2 ). Uranium zirconium hydride (U-ZrHx) is a well-established fuel system for pulsing reactors which has been intermittently manufactured throughout the decades. U-ZrHx offers similar capabilities to UO 2 designs in terms of nuclear kinetics, but with about half the fluence capability (~1E15 n/cm 2 ). An evolution of the UO 2 system, termed “ternary ceramic” fuel, shows that dispersing UO 2 in zirconium oxide and calcium oxide can increase fluence capability greatly (~7E15 n/cm 2 ), but is not presently a commonly available fuel form. A unique composite of UO 2 and beryllium oxide (UO 2 -BeO) can be used to create a core with similar kinetics and compact core geometry as U-ZrHx designs, but with significantly higher fluence capability (~6E15 n/cm 2 ). Like ternary ceramic fuel, newly fabricated UO 2 -BeO would require reestablishing its historic manufacturing process which would be further complicated by the health hazards associated with beryllium. In conclusion, like most engineering problems, there is no perfect solution, but this paper outlines the advantages and disadvantages of candidate fuel options to help guide detailed evaluations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Design Rules for Carboborothermic Reduction Synthesis of High Uranium Density UB 4 –UBC Composites

Uranium borides are promising candidate fuel forms for use in advanced nuclear reactors due to their high thermal conductivity and potential for dual use as both fuel and burnable absorber. In this work, uranium tetraboride () and uranium monoboroncarbide (UBC) composite were synthesized by using industrially scalable carboborothermic reduction method. The final uranium boride phase composition is sensitive to the sample holding crucibles ( and graphite) such that graphite supply excess carbon, promoting the formation of a predominant UBC phase. The high‐temperature in situ synchrotron X‐ray diffraction of pristine –UBC show persistence , UBC, and phases while preoxidized –UBC leads to predominant and formation due to progressive oxidation and boron loss at high temperature. The oxidation behavior was further characterized using thermogravimetric analysis, allowing direct comparison with other potential accident tolerant fuels such as , , UC, and UN. The –UBC shows higher uranium loading than monolithic and demonstrates promising oxidation behavior at high temperature, pointing to its potential as an improved uranium boride‐based fuel form.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗