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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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All sourcesNASA NTRSDOE OSTI
NASA NTRS2015

Low Earth Orbit Environmental Durability of Recently Developed Thermal Control Coatings

The Materials International Space Station Experiment provided a means to expose materials and devices to the low Earth orbit environment on the exterior of the International Space Station. By returning the specimens to Earth after flight, the specimens could be evaluated by comparison with pre-flight measurements. One area of continuing interest is thermal control paints and coatings that are applied to exterior surfaces of spacecraft. Though traditional radiator coatings have been available for decades, recent work has focused on new coatings that offer custom deposition or custom optical properties. The custom deposition of interest is plasma spraying and one type of coating recently developed as part of a Small Business Innovative Research effort was designed to be plasma sprayed onto radiator surfaces. The custom optical properties of interest are opposite to those of a typical radiator coating, having a combination of high solar absorptance and low infrared emittance for solar absorber applications, and achieved in practice via a cermet coating. Selected specimens of the plasma sprayed coatings and the solar absorber coating were flown on Materials International Space Station Experiment 7, and were recently returned to Earth for post-flight analyses. For the plasma sprayed coatings in the ram direction, one specimen increased in solar absorptance and one specimen decreased in solar absorptance, while the plasma sprayed coatings in the wake direction changed very little in solar absorptance. For the cermet coating deployed in both the ram and wake directions, the solar absorptance increased. Interestingly, all coatings showed little change in infrared emittance.

Optical Properties↗
NASA NTRS2015

Affordable Development and Demonstration of a Small NTR Engine and Stage: How Small is Big Enough?

In FY11, NASA formulated a plan for Nuclear Thermal Propulsion (NTP) development that included Foundational Technology Development followed by system-level Technology Demonstrations The ongoing NTP project, funded by NASAs Advanced Exploration Systems (AES) program, is focused on Foundational Technology Development and includes 5 key task activities:(1) Fuel element fabrication and non-nuclear validation testing of heritage fuel options;(2) Engine conceptual design;(3) Mission analysis and engine requirements definition;(4) Identification of affordable options for ground testing; and(5) Formulation of an affordable and sustainable NTP development program Performance parameters for Point of Departure designs for a small criticality-limited and full size 25 klbf-class engine were developed during FYs 13-14 using heritage fuel element designs for both RoverNERVA Graphite Composite (GC) and Ceramic Metal (Cermet) fuel forms To focus the fuel development effort and maximize use of its resources, the AES program decided, in FY14, that a leader-follower down selection between GC and cermet fuel was required An Independent Review Panel (IRP) was convened by NASA and tasked with reviewing the available fuel data and making a recommendation to NASA. In February 2015, the IRP recommended and the AES program endorsed GC as the leader fuel In FY14, a preliminary development schedule DDTE plan was produced by GRC, DOE industry for the AES program. Assumptions, considerations and key task activities are presented here Two small (7.5 and 16.5 klbf) engine sizes were considered for ground and flight technology demonstration within a 10-year timeframe; their ability to support future human exploration missions was also examined and a recommendation on a preferred size is provided.

spacecraft design↗
NASA NTRS2015

Affordable Development and Demonstration of a Small NTR Engine and Stage: A Preliminary NASA, DOE, and Industry Assessment

In FY11, NASA formulated a plan for Nuclear Thermal Propulsion (NTP) development that included Foundational Technology Development followed by system-level Technology Demonstrations. The ongoing NTP project, funded by NASA's Advanced Exploration Systems (AES) program, is focused on Foundational Technology Development and includes 5 key task activities: (1) Fuel element fabrication and non-nuclear validation testing of heritage fuel options; (2) Engine conceptual design; (3) Mission analysis and engine requirements definition; (4) Identification of affordable options for ground testing; and (5) Formulation of an affordable and sustainable NTP development program. Performance parameters for Point of Departure designs for a small criticality-limited and full size 25 klb (sub f) (kip, or 1000 pound-force)-class engine were developed during FYs 13-14 using heritage fuel element designs for both Rover NERVA (Nuclear Engine for Rocket Vehicle Applications) Graphite Composite (GC) and Ceramic Metal (Cermet) fuel forms. To focus the fuel development effort and maximize use of its resources, the AES program decided, in FY14, that a leader-follower down-selection between GC and cermet fuel was required. An Independent Review Panel (IRP) was convened by NASA and tasked with reviewing the available fuel data and making a recommendation to NASA. In February 2015, the IRP recommended and the AES program endorsed GC as the leader fuel In FY14, a preliminary development schedule DDTE (Design Development Test and Evaluation) plan was produced by GRC (Glenn Research Center), DOE (Department of Energy) and industry for the AES program. Assumptions, considerations and key task activities are presented here. At the direction of NASA HQ (32515), NASA and DOE are to work together to formulate a detailed development plan and schedule allowing the affordable development of a small (7.5-16.5 klbf) GC engine for possible flight technology demonstration (FTD) mission within a 10-year timeframe.

Lunar Flyby Mission↗
DOE OSTI2025

Microscopic analysis of irradiated UN–Mo–W for space nuclear propulsion

Nuclear propulsion is an effective method of providing thrust in deep space, allowing for shorter mission duration for long-term surface missions to Mars. The fuel for nuclear propulsion systems must exhibit superior mechanical and thermal properties, to decrease the possibility of fuel failure and to increase reactor efficiency. Here, the objective of the Sirius project is to fabricate and irradiate cermet fuels at the Idaho National Laboratory Transient Reactor Facility to determine the properties of the fuel after irradiation at prototypical nuclear propulsion start-up conditions. The Sirius project featured 3 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A and -2B featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). The fuel samples from the Sirius-2A and -2B capsules were examined after irradiation using optical microscopy, scanning electron microscopy, and X-ray diffraction analyses. The mechanical properties of the fuel from Sirius-2A and -2B are discussed in a previous work. In this study, the results of the optical microscopy, scanning electron microscopy, and X-ray diffraction analyses are discussed.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Comparison of Nuclear Thermal Propulsion Reactor Cores Using Different Fuel Forms, Neutron Spectra, and Uranium Enrichments

A comparison of 300 MWt HEU and HALEU NTP cores using different fuel forms is presented to demonstrate trends and potentially inform fuel development. Compared to HEU, HALEU cores require additional in-core moderation. Beryllium was chosen as the reference in-core moderator material; metal hydrides may also be considered. Typical NTP fuel forms such as carbide and carbonitride solid solutions, carbide-graphite composites, CERMETs and CERCERs were considered together with less typical fuel forms such as UC and U metal. Be-moderated HALEU cores are heavier than HEU cores (1.9-2.7 and 0.9-2.0 metric tons, respectively). Furthermore, HALEU loading is very sensitive to the amount of in-core moderation which translates into tight fabrication tolerances. HALEU cores requires more reactivity control mechanisms than HEU cores to compensate for temperature feed-back. Using enrichments only slightly above HALEU (e.g., 25%) may provide a compromise between the administrative preference for using HALEU and the engineering preference for simplifying the reactor systems by, for example, not using in-core moderators.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Infiltrated electrodes for metal supported solid oxide electrolysis cells

Metal-supported solid oxide cells (MSOCs) are an alternative to conventional solid oxide cells (SOCs) based on ceramic cermets, offering lower material costs and higher operational flexibility. In this study symmetric MSOCs with infiltrated electrodes are explored for steam electrolysis operation to understand the underlying operation and degradation principles and suggest a direction for future MSOCs development. Two different fuel electrode backbones are used: an electronically-conductive lanthanum strontium co-doped iron nickel titanate (LSFNT) infiltrated with cerium-gadolinium oxide (CGO), or an ionic conductive zirconia based backbone (10ScYSZ) infiltrated with Ni:CGO. At the oxygen side, the backbone is 10ScYSZ, which is infiltrated with lanthanum-strontium co-doped cobalt oxide (LSC), or praseodymium oxide as cobalt-free alternative for comparison. This study suggests that the backbone electronic conductivity is key for good electrochemical performance as well as for boosting cell durability. Highly electronically conductive nanoparticles, especially nickel, were observed to irreversibly agglomerate driven by thermal conditions, whereas CGO proved to be a very stable electrocatalyst. At the fuel side, CGO (LSFNT) electrode showed lower ASR and degradation rate than Ni:CGO(ScYSZ) configuration with measured values of 0.50 Ω cm2 and 11 %/1000 h (at 0.60 A/cm2), and 0.70 Ω cm2 and 26 %/1000 h (at 0.50 A/cm2) at 1.30 V, respectively (700 °C, 50 % steam in hydrogen at the fuel side and air at the oxygen electrode side, LSC(ScYSZ) oxygen electrode).

25 ENERGY STORAGE↗
DOE OSTI2025

Effects of a refractory metal protective barrier on uranium carbonitride ceramic–metallic composites exposed to a high temperature hydrogen environment

Uranium nitride has been under consideration at NASA and the Department of Energy for nuclear thermal propulsion applications intended for deep space exploration vehicles. It is critical to the success of the technology development that the behavior of the fuel system is properly characterized in operational environments. This body of work fabricated ceramic metal composites consisting of uranium ceramic fuel in a molybdenum and molybdenum–tungsten metal matrix, both with and without an exterior molybdenum barrier. Specimens were introduced to a high temperature hydrogen environment and characterized through x-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy to identify characteristic changes. A portion of the specimens was fabricated with a 1.27 mm thick molybdenum barrier in order to isolate the material from the hydrogen environment. Instability of the uranium nitride occurred as low as 2073 K, leading to aggressive instability approaching 2658 K. Carbon reactions with the metal matrix were observed as well as alloying between the uranium and metal matrix constituents. The liquid uranium resulting from the instability of the uranium nitride was observed to attack the grain boundaries of the refractory metal, resulting in the formation of a liquid uranium molybdenum eutectic melt. Specimens encased in a barrier experienced similar results. The authors conclude this is a significant challenge to applications of this cermet at these temperatures. Alternate configurations with resilient hermetic seals or lower temperature applications such as power reactors may have more success.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Mass Optimization of a Multilayered Shield for Transportable Microreactors

The ability to easily transport microreactors is a major selling point for deploying microreactors to remote areas. However, this creates a unique shielding challenge, especially when the microreactor is being shipped after irradiation. A traditional reactor configuration utilizes a separate biological shield and pressure vessel to meet radiological shielding and pressure needs. The limited space available for transportable microreactors for both shielding and pressure vessels requires a revised assessment of separating out the biological shield and pressure vessel. To address these concerns, we examine a nuclear-grade sandwich composite (NGSC) that combines the reactor pressure vessel and biological shielding functions into a single component. Through a series of optimization problems for both transportation and operational use cases, the NGSC is able to minimize dose, minimize the vessel cost, and ensure that weight requirements are met for transportation. Initial results show that using a tungsten-tetraboride cermet in the first two layers of a six-layer NGSC provides adequate shielding for both use cases. These results show promise that an NGSC has enough overlap between operational and transportation cases to help reduce the design space for future analysis and assessment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Evaluation of Radiation Transport Through a Nuclear-Grade Sandwich Composite

The development and deployment of microreactors are often synonymous with their transportability; however, designing a shield which is appropriate for both transportation and operations is challenging. A recent approach is the use of composite style shield which combines the reactor pressure vessel and biological shielding into a single unit to enhance the shields effectiveness without degrading the structural properties. To assess the viability of a nuclear grade sandwich composite (NGSC), and understanding of how radiation transports through the shield needs to be understood to select the appropriate materials. This work examines a two-layered NGSC of varying thicknesses with two primary shielding materials (tungsten-tetraboride cermet and boron carbide). To assess the NGSC, quantities of interest including neutron attenuation, photon attenuation, energy deposition, displacements per atom, helium generation, and dose are examined. Altogether, it is found that a combination of both tungsten-tetraboride and boron carbide are necessary to act as an appropriate shield, where the order in which these materials are place in the composite are important to their effectiveness.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

COMSOL Results for the Nominal Steady-State Operation of the Proposed 95-MW LEU Silicide Core for HFIR Conversion

Engineering design studies are being performed to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel at Oak Ridge National Laboratory. This activity is sponsored by the Office of Reactor Conversion and Uranium Supply (ORCUS) under the auspices of the US Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization. HFIR is a very high flux, pressurized, light water–cooled and moderated, flux trap–type research reactor with a core made of involute shaped U 3 O 8 /Al cermet fuel plates and coolant channels. HFIR currently operates at a thermal power of 85 MW and supports key national and international missions in neutron scattering, isotope production, materials/fuels irradiation, neutron activation analysis, gamma irradiation, and neutrino research. Advanced multiphysics computational fluid dynamics models have been developed in the COMSOL Multiphysics software to simulate the steady-state operating conditions for the proposed low-and high-density LEU U 3 Si 2 -Al (uranium silicide dispersion) fuel designs. The COMSOL models for HFIR inner and outer fuel element models incorporate various essential inputs and physics such as spatially dependent nuclear heat deposition, multilayer heat conduction, conjugate heat transfer, turbulent flows (using Reynolds-averaged Navier Stokes turbulence models), structural mechanics (thermal–structural interactions and fuel swelling), and oxide layer build-up. This report presents the best-estimate thermal hydraulics results for the low- and high-density optimized silicide LEU core designs at 95 MW steady-state nominal operation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1956

Coke Deposition and Smoke Formation in Turbojet Engines

In the early development of jet engines, it was occasionally found that excessive amounts of coke or other carbonaceous deposits were formed in the combustion chamber. Sometimes a considerable amount of smoke was noted in the-exhaust gases. Excessive coke deposits may adversely affect jet-engine performance in several ways. The formation of excessive amounts of coke on or just downstream of a fuel nozzle (figs. 116(a) and (b)) changes the fuel-spray pattern and possibly affects combustor life and performance. Similar effects on performance can result from the deposition of coke on primary-air entry ports (fig. 116(c)). Sea-level or altitude starting may be impaired by the deposition of coke on spark-plug electrodes (fig. 116(b)), deposits either grounding the electrodes completely or causing the spark to occur at positions other than the intended gap. For some time it was thought that large deposits of coke in turbojet combustion chambers (fig. 116(a)) might break away and damage turbine blades; however, experience has indicated that for metal blades this problem is insignificant. (Cermet turbine blades may be damaged by loose coke deposits.) Finally, the deposition of coke may cause high-temperature areas, which promote liner warping and cracking (fig. 116(d)) from excessive temperature gradients and variations in thermal-expansion rates. Smoke in the exhaust gases does not generally impair engine performance but may be undesirable from a tactical or a nuisance standpoint. Appendix B of reference 1 and references 2 to 4 present data obtained from full-scale engines operated on test stands and from flight tests that indicate some effects on performance caused by coke deposits and smoke. Some information about the mechanism of coke formation is given in reference 5 and chapter IX. The data indicate that (1) high-boiling fuel residuals and partly polymerized products may be mixed with a large amount of smoke formed in the gas phase to account for the consistency, structure, and chemical composition of the soft coke in the dome and (2) the hard deposits on the liner are similar to petroleum coke and may result from the liquid-phase thermal cracking of the fuel. During the early development period of jet engines, it was noted that the excessive coke deposits and exhaust smoke were generally obtained when fuel-oil-type fuels were used. Engines using gasoline-type fuels were relatively free from the deposits and smoke. These results indicated that some type of quality control would be needed in fuel specifications. Also noted was the effect of engine operating conditions on coke deposition. It is possible that, even with a clean-burning fuel, an excessive amount of coke could be formed at some operating conditions. In this case, combustor redesign could possibly reduce the coke to a tolerable level. This chapter is a summary of the various coke-deposition and exhaust-smoke problems connected- with the turbojet combustor. Included are (1) the effect of coke deposition on combustor life or durability and performance; (2) the effect of combustor design, operating conditions, inlet variables, and fuel characteristics on coke deposition; (3) elimination of coke deposits; (4) the effect of operating conditions and fuel characteristics on formation of exhaust smoke; and (5) various bench test methods proposed for determining and controlling fuel quality.

Hibbard, R. R.↗
NASA NTRS1971

Dynamics of heat-pipe reactors

A split-core heat pipe reactor, fueled with either U(233)C or U(235)C in a tungsten cermet and cooled by 7-Li-W heat pipes, was examined for the effects of the heat pipes on reactor while trying to safely absorb large reactivity inputs through inherent shutdown mechanisms. Limits on ramp reactivity inputs due to fuel melting temperature and heat pipe wall heat flux were mapped for the reactor in both startup and at-power operating modes.

Niederauer, G. F.↗
NASA NTRS1972

Radiation from plutonium 238 used in space applications

The principal mode of the nuclear decay of plutonium 238 is by alpha particle emission at a rate of 17 curies per gram. Gamma radiation also present in nuclear fuels arises primarily from the nuclear de-excitation of daughter nuclei as a result of the alpha decay of plutonium 238 and reactor-produced impurities. Plutonium 238 has a spontaneous fission half life of 4.8 x 10 to the 10th power years. Neutrons associated with this spontaneous fission are emitted at a rate of 28,000 neutrons per second per gram. Since the space fuel form of plutonium 238 is the oxide pressed into a cermet with molybdenum, a contribution to the neutron emission rate arises from (alpha, n) reactions with 0-17 and 0-18 which occur in natural oxygen.

Keenan, T. K.↗
NASA NTRS1972

Development and fabrication of insulator seals for thermionic diodes

Eight different types of cermet seals for thermionic diodes were investigated: (1) 1 micron Al2O3 with Nb spheres; (2) 200 A Al2O3 with Nb spheres; (3) 1 micron Al2O3 with Nb 1% Zr spheres; (4) 200 A Al2O3 with Nb 1% Zr spheres; (5) Pure Y2O3 with Nb 1% Zr spheres; (6) Y2O3 3% ZrO2 with Nb 1% Zr spheres; (7) Y2O3 10% ZrO2 with Nb 1% Zr spheres; and (8) ZrO2 12% Y2O3 with Nb 1% Zr spheres. Investigations were made to determine the most favorable fabrication techniques and the effect of the bonding cycle, (length of bonding time and shutdown sequences). The analysis of the seals included tensile test, vacuum test, electrical test and metallurgical examination. At the conclusion of the development phase, 36 seals were fabricated for delivery for evaluation.

Poirier, V. L.↗