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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Advanced Fuel Cycle Cost Basis Report: Module D1-4 and Module D1-5 Ceramic Pelletized Sodium-Cooled Fast Reactor (SFR) Fuel Fabrication Ceramic Vibrocompacted Fuel Fabrication

This is a cost module that is part of the Advanced Fuel Cycle -- Cost Basis Report. Module D1-4: Nature of this module update (Rev 1) from previous advanced fuel cycle cost basis reports (AFCCBRs): new life cycle cost data on U,Pu SFR mixed oxide (MOX) fuels is derived from the Nonproliferation Assessment Systems Analysis Program (NASAP) conducted in the late 1970s. Highassay low-enriched uranium (HALEU) ceramic fuel is also discussed in more detail compared to earlier AFC-CBRs, since some advanced SFR concepts currently under development will require this HALEU fuel type for startup. Module D1-5: Nature of this FY-21 module update from previous AFC-CBRs: this module includes a few new references and a somewhat expanded discussion of vibrocompaction fuel fabrication technology. The WIT values for VIPAC are pegged directly to those for conventional LWR and SFR pelletized U,Pu MOX (note that this module now includes VIPAC MOX fuel for LWRs in addition to SFRs). Based on information from Russia, where VIPAC has been studied extensively, the unit costs are expressed as a percentage of those in the new updated Module D1-2 (pelletized U,Pu MOX) and new updated Module D1-4 (pelletized U,Pu MOX) for large NOAK fabrication facilities of the same production capacity. Since both ceramic pelletized MOX Modules D1-2 and D1-4 benefitted from analysis of 1970s NASAP data, by extension this Module D1-5 also benefits.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Advanced Fuel Cycle Cost Basis Report: Module D1-1: Uranium-based Ceramic LWR Fuel Fabrication (Rev.2)

Cost module on LWR fuel fabrication to accompany the Advanced Fuel Cycle Cost Basis Report. In this update, we are adding detailed life cycle cost data and a calculated, levelized fabrication cost derived from a non-proprietary bottom-up estimate prepared in 1978 by Oak Ridge National Laboratory (ORNL) (Judkins and Olsen 1978a). In 2018, the 1978 ORNL estimate was escalated by SA&I authors to 2017 USD using factors that represent inflation, escalation above inflation typical of nuclear projects, and the effects of more stringent safety and environmental regulations. In this FY-21 document, the $/kgU results in 2017 USD from the unpublished 2018 interim study (Williams and Ganda 2018) can be escalated to 2020 USD using a factor of 1.052. The literature-based unit cost (or price) data from previous (2004–2017) AFC-CBRs are escalated to 2020 USD using factors from Chapter 8 of the main FY-21 AFC-CBR document. This data, in addition to the results of the updated bottom-up estimate, are used to define the “what-it-takes” (WIT) range for the unit fabrication costs for conventional ceramic UOX light-water reactor (LWR) fuel. This FY-23 document also includes calculated unit costs for accident-tolerant LWR fuels (ATFs) of three different types. Some of these fuels constitute a ceramic pelletized form with fuel meat uranium compounds other than UO2 (a.k.a.,. UOX), thus the change in the title of this module in which the word “UO2” is changed to “Uranium-based Ceramic.”

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2017

Nuclear Rocket Ceramic Metal Fuel Fabrication Using Tungsten Powder Coating and Spark Plasma Sintering

Nuclear thermal propulsion is an enabling technology for crewed Mars missions. An investigation was conducted to evaluate spark plasma sintering (SPS) as a method to produce tungsten-depleted uranium dioxide (W-dUO2) fuel material when employing fuel particles that were tungsten powder coated. Ceramic metal fuel wafers were produced from a blend of W-60vol% dUO2 powder that was sintered via SPS. The maximum sintering temperatures were varied from 1,600 to 1,850 C while applying a 50-MPa axial load. Wafers exhibited high density (>95% of theoretical) and a uniform microstructure (fuel particles uniformly dispersed throughout tungsten matrix).

Barnes, M. W.↗
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 OSTI2021

Chemically Inert Hydrocarbon-Based Slurries for Rapid Laser Sintering of Thin Proton-Conducting Ceramics

The process of rapid laser sintering of thin BaZrO 3 –BaCeO 3 -based proton-conducting electrolytes is being developed for easy fabrication of ceramic fuel cells and electrolyzers. However, cracks on the electrolytes caused by volume change due to chemical reactions between the basic ceramic constituents and the polar solvents during wet processing has been problematic. In order to address this issue, the use of chemically inert saturated-hydrocarbon-based slurries comprised of hexadecane, polybutene, and a long-chain saturated fatty acid were investigated in this work. By optimizing slurry composition and laser sintering conditions, a 20 mm long, 4 mm wide, 13.5-μm-thick and 97%-dense BaCe 0.7 Zr 0.1 Y 0.07 Sm 0.13 O 3-d membrane showing proton conductivity on the order of 10 –4 S•cm –1 at 600 °C was successfully prepared in just three seconds by laser sintering. As a result, the use of saturated-hydrocarbon-based slurries will facilitate wet processing and rapid laser sintering of proton-conducting ceramic electrolytes.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Fracture toughness measurement of ceramics with fuel pellet geometry

The crack initiation of nuclear fuel pellets, driven by extreme temperature gradients, is a critical factor impacting reactor performance. Although this fact is widely acknowledged, there is limited fracture toughness data for specimens with representative geometry and fabricated using relevant processing methods. Here, to address this, a new chevron-notched short-rod specimen configuration with diameter and length both equal to 10 mm, representative of a conventional fuel pellet and tested in tension, was developed in an attempt to validly measure Mode I fracture toughness, K Ic . Test calibration was carried out by testing polycrystalline ZrO 2 , SiC, and Si 3 N 4 ceramics fabricated with conventional fuel pellet geometry and known fracture toughnesses. The results suggest that the method described in this paper produces valid K Ic measurements using conventional ceramic fuel pellets.

Ceramics↗
NASA NTRS1990

Multilayer ceramic oxide solid electrolyte for fuel cells and electrolysis cells and method for fabrication thereof

An unitary layered ceramic structure is disclosed which comprises co-sintered layers. The co-sintered structure comprises a sintered central layer of yttria stabilized zirconia (YSZ) which is about 8 mole percent yttria and having a density of at least about 95% of theoretical, and sintered outer layers of strontium lanthanum manganite (LSM) having the approximate molecular composition La.sub.0.8 Sr.sub.0.2 MnO.sub.3, having a density from about 50 to about 60% of theoretical, and having interconnected porosity from about 40 to 50% with an interconnected pore diameter from about one micron to about five microns. The sintered central layer is sandwiched by and bonded and sintered to the outer layers and is essentially free of significant amounts of manganese. A process for making the unitary composition-of-matter is also disclosed which involves tape casting a LSM tape and then on top thereof casting a YSZ tape. The process comprises presintering LSM powder at 1250.degree. F., crushing the presintered commercially available LSM powder, forming a slurry with the crushed LSM, a binder and solvent, tape casting the slurry and allowing the slurry to air dry. A mixture of commercially available submicron size particle YSZ powder is milled with a dispersant and solvent to disperse the YSZ particles thereby forming a dispersed YSZ slurry. The YSZ slurry is then tape cast on the dried LSM tape. If desired, a third layer of LSM can be cast on top of the dried YSZ layer. After drying the composite LSM/YSZ and LSM/YSZ/LSM tapes are fired at 1300.degree. C. No migration of manganese into the YSZ layer was observed with scanning electron microscope/edax in the sintered multilayer tape.

Schroeder, James E.↗
NASA NTRS2009

Regenerative Performance of the NASA Symmetrical Solid Oxide Fuel Cell Design

The NASA Glenn Research Center is developing both a novel cell design (BSC) and a novel ceramic fabrication technique to produce fuel cells predicted to exceed a specific power density of 1.0 kW/kg. The NASA Glenn cell design has taken a completely different approach among planar designs by removing the metal interconnect and returning to the use of a thin, doped LaCrO3 interconnect. The cell is structurally symmetrical. Both electrodes support the thin electrolyte and contain micro-channels for gas flow-- a geometry referred to as a bi-electrode supported cell or BSC. The cell characteristics have been demonstrated under both SOFC and SOE conditions. Electrolysis tests verify that this cell design operates at very high electrochemical voltage efficiencies (EVE) and high H2O conversion percentages, even at the low flow rates predicted for closed loop systems encountered in unmanned aerial vehicle (UAV) applications. For UAVs the volume, weight and the efficiency are critical as they determine the size of the water tank, the solar panel size, and other system requirements. For UAVs, regenerative solid oxide fuel cell stacks (RSOFC) use solar panels during daylight to generate power for electrolysis and then operate in fuel cell mode during the night to power the UAV and electronics. Recent studies, performed by NASA for a more electric commercial aircraft, evaluated SOFCs for auxiliary power units (APUs). System studies were also conducted for regenerative RSOFC systems. One common requirement for aerospace SOFCs and RSOFCs, determined independently in each application study, was the need for high specific power density and volume density, on the order of 1.0 kW/kg and greater than 1.0 kW/L. Until recently the best reported performance for SOFCs was 0.2 kW/kg or less for stacks. NASA Glenn is working to prototype the light weight, low volume BSC design for such high specific power aerospace applications.

Cable, Thomas L.↗
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↗
NASA NTRS2005

Fabrication of High Temperature Cermet Materials for Nuclear Thermal Propulsion

Processing techniques are being developed to fabricate refractory metal and ceramic cermet materials for Nuclear Thermal Propulsion (NTP). Significant advances have been made in the area of high-temperature cermet fuel processing since RoverNERVA. Cermet materials offer several advantages such as retention of fission products and fuels, thermal shock resistance, hydrogen compatibility, high conductivity, and high strength. Recent NASA h d e d research has demonstrated the net shape fabrication of W-Re-HfC and other refractory metal and ceramic components that are similar to UN/W-Re cermet fuels. This effort is focused on basic research and characterization to identify the most promising compositions and processing techniques. A particular emphasis is being placed on low cost processes to fabricate near net shape parts of practical size. Several processing methods including Vacuum Plasma Spray (VPS) and conventional PM processes are being evaluated to fabricate material property samples and components. Surrogate W-Re/ZrN cermet fuel materials are being used to develop processing techniques for both coated and uncoated ceramic particles. After process optimization, depleted uranium-based cermets will be fabricated and tested to evaluate mechanical, thermal, and hot H2 erosion properties. This paper provides details on the current results of the project.

Hickman, Robert↗
DOE OSTI2020

Fabrication of UN-Mo CERMET Nuclear Fuel Using Advanced Manufacturing Techniques

Ceramic-metallic nuclear fuels are a candidate fuel for nuclear thermal propulsion systems due to their high heat transport properties, which are necessary in very high-temperature environments. The conventional fabrication of uranium nitride–molybdenum fuel has been thoroughly studied in the past, but modern manufacturing techniques have presented a unique opportunity for further development within this field. This work demonstrates the use of advanced manufacturing techniques to produce nuclear fuel pellets composed of uranium nitride microspheres encased in a molybdenum matrix. Binder jetting is used to print molybdenum disks that are filled with uranium nitride microspheres and afterward sintered using spark plasma sintering. Two fuel pellets were fabricated to demonstrate the methodology and to provide a baseline analysis of the effects of temperature and pressure processing conditions. Characterization of the sintered fuel pellets includes detailed microstructural analysis and thermal conductivity measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2007

Freeze Tape Casting of Functionally Graded Porous Ceramics

Freeze tape casting is a means of making preforms of ceramic sheets that, upon subsequent completion of fabrication processing, can have anisotropic and/or functionally graded properties that notably include aligned and graded porosity. Freeze tape casting was developed to enable optimization of the microstructures of porous ceramic components for use as solid oxide electrodes in fuel cells: Through alignment and grading of pores, one can tailor surface areas and diffusion channels for flows of gas and liquid species involved in fuel-cell reactions. Freeze tape casting offers similar benefits for fabrication of optimally porous ceramics for use as catalysts, gas sensors, and filters.

Sofie, Stephen W.↗
DOE OSTI2024

Post-irradiation examination of UN-Mo-W fuels for space nuclear propulsion

The National Aeronautics and Space Administration's return to space nuclear propulsion stems from the need for a more efficient method of space travel. Nuclear thermal propulsion systems have been shown to be two times more efficient than chemical propulsion. NASA's Sirius program was created to fabricate and test fuels for space nuclear propulsion, specifically to determine their performance under prototypical startup conditions. The Sirius project featured 4 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A, -2B, and -3 featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). This study discusses the Sirius-2A and -2B irradiation experiments at the Idaho National Laboratory, specifically their performance under irradiation at the Transient Reactor Test Facility. It was found that the fuel samples overall did not exhibit significant cracking, though the Sirius-2A fuel did have one large crack on the surface of the fuel. There was minimal hydrogen absorption in the samples, though it is unknown if the absorption occurred during irradiation or during fabrication. Mechanical testing indicated that the UN fuel demonstrated ceramic behavior as expected, and the Mo/W matrix demonstrated linear elastic behavior to failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗