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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
NASA NTRS1971

Compatibility of buffered uranium carbides with tungsten.

Results of compatibility tests between tungsten and hyperstoichiometric uranium carbide alloys run at 1800 C for 1000 and 2500 hours. These tests compared tungsten-buffered uranium carbide with tungsten-buffered uranium-zirconium carbide. The zirconium carbide addition appeared to widen the homogeneity range of the uranium carbide, making additional carbon available for reaction. Reaction layers could be formed by either of two diffusion paths, one producing UWC2, while the second resulted in the formation of W2C. UWC2 acts as a diffusion barrier for carbon and slows the growth of the reaction layer with time, while carbon diffusion is relatively rapid in W2C, allowing equilibrium to be reached in less than 2500 hours at a temperature of 1800 C.

Phillips, W. M.↗
NASA NTRS1972

Some observations on uranium carbide alloy/tungsten compatibility

Chemical compatibility between both pure and thoriated tungsten and uranium carbide alloys was studied at 1800 C for up to 3300 hours. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, dependent upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. The presence of a thermal gradient had no effect on the reactions observed nor did the presence of thoria in the tungsten clad.

Phillips, W. M.↗
NASA NTRS1972

Some observations on uranium carbide alloy/tungsten compatibility.

Results of chemical compatibility tests between both pure tungsten and thoriated tungsten run at 1800 C for up to 3300 hours with uranium carbide alloys. Alloying with zirconium carbide appeared to widen the homogeneity range of uranium carbide, making additional carbon available for reaction with the tungsten. Reaction layers were formed both by vapor phase reaction and by physical contact, producing either or both UWC2 and W2C, depending upon the phases present in the starting fuel alloy. Formation of UWC2 results in slow growth of the reaction layer with time, while W2C reaction layers grow rapidly, allowing equilibrium to be reached in less than 2500 hours at 1800 C. Neither the presence of a thermal gradient nor the presence of thoria in the tungsten clad affect the reactions observed.

Phillips, W. M.↗
NASA NTRS1974

Out-of-pile creep behavior of uranium carbide

Compression creep tests were investigated on various UC-based fuel materials having a variation in both density and composition. Specimens were prepared by casting and by hot pressing. Steady-state creep rates were measured under vacuum at 1400 to 1800 C in the stress range 500-4000 psi.

Wright, T. R.↗
NASA NTRS2026

Thermodynamic Properties of Uranium-Zirconium Carbide Solid Solutions with Oxygen

Uranium-zirconium carbide solid solutions are potential candidates for fuel in space nuclear reactors. To support development of advanced robust fuels for space applications, (U y Zr 1-y )C x O z solid solutions were synthesized via carbothermic reduction with uranium, carbon, and oxygen contents raging from y=0.05 – 0.3, x=0.95 – 0.98, and z = 0.02 – 0.07. These solid solutions were subsequently characterized by X-ray diffraction and elemental combustion analysis. High-temperature oxidative drop solution calorimetry in molten sodium molybdate solvent at 1073 K was employed to determine the formation, oxidation, and mixing enthalpies of (U y Zr 1-y )C x O z solid solutions, and thereby for evaluating their thermodynamic stability. The solid solutions are enthalpically less stable than the UC and ZrC end members. They also become less stable and more reactive towards oxidation with increasing uranium content. Incorporation of oxygen into solid solutions increases their solid solution-stability and improves their oxidation resistance.

Calorimetry↗
NASA NTRS1975

Creep behavior of uranium carbide-based alloys

The present work gives the results of experiments on the influence of zirconium carbide and tungsten on the creep properties of uranium carbide. The creep behavior of high-density UC samples follows the classical time-dependence pattern of (1) an instantaneous deformation, (2) a primary creep region, and (3) a period of steady-state creep. Creep rates for unalloyed UC-1.01 and UC-1.05 are several orders of magnitude greater than those measured for carbide alloys containing a Zr-C and/or W dispersoid. The difference in creep strength between alloyed and unalloyed materials varies with temperature and applied stress.

Seltzer, M. S.↗
NASA NTRS1973

Irradiation test of tungsten clad uranium carbide-zirconium carbide ((U,Zr)C) specimens for thermionic reactor application at conditions conductive to long-term performance

Uranium carbide fueled, thermionic emitter configurations were encapsulated and irradiated. One capsule contained a specimen clad with fluoride derived chemically vapor deposited (CVD) tungsten. The other capsule used a duplex clad specimen consisting of chloride derived on floride derived CVD tungsten. Both fuel pins were 16 millimeters in diameter and contained a 45.7-millimeter length of fuel.

Creagh, J. W. R.↗
NASA NTRS1972

Material development for thermionic fuel-cladding systems.

Fuel-cladding systems capable of maintaining stable dimension and electron emission characteristics at high temperatures under irradiation are essential to the successful operation of nuclear thermionic power sources. Studies regarding two possible systems, involving tungsten clad uranium carbide and uranium oxide, are considered. Out-of-pile developmental efforts and the in-pile evaluation results for the two systems are described. Prototypical emitters have been tested for 8000 hours for the carbide system and 4000 hours for the oxide system.

Yang, L.↗
NASA NTRSDate not supplied

Additive Manufacturing of ZrC for Nuclear Thermal Propulsion Applications

Looking into the future, we are seeing that nuclear is becoming a more and more viable fuel source for space applications due to it being one of the safest energy forms, low greenhouse gas emissions, and inexpensive maintenance. National Aeronautics and Space Administration (NASA) and a few others are looking at nuclear fuels in relation to space travel, specifically in the form of Nuclear Thermal Propulsion (NTP) [1]. NTP provides the answer for space travel to Mars where chemical rockets would not complete the job. A concept art of what these spacecrafts could look like is shown (Figure 1). Uranium Carbide fuel has been chosen to heat a liquid propellant for propulsion. The complex fuel geometries required by space reactors can be achieved by novel techniques fairly easily, cheaply, and safely. In this specific study, we are looking at using Zirconium Carbide (ZrC) to use as a surrogate for Uranium Carbide (UC) due to their similar behaviors under high temperatures. An ink made with ZrC will be made and will be extruded into unique, complex geometries. This is a great starting point to test additive manufacturing on these ceramics.

Additive Manufacturing↗
NASA NTRS2012

Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

Valentine, Peter G.↗
NASA NTRSDate not supplied

Centrifugal Nuclear Thermal Rocket Challenges and Potential

The Centrifugal Nuclear Thermal Rocket (CNTR) is a liquid fueled fission propulsion concept designed to heat propellant to 5000 K prior to expansion through a nozzle. A specific impulse up to 1800 s may be achieved using hydrogen propellant, and a specific impulse up to 1000 s may be achieved using more storable propellants such as methane, ammonia, or propane. The high uranium density of the liquid metallic uranium or liquid uranium carbide fuel will help enable compact engines suitable for missions such as fast (<15 month) round trip human Mars missions or high delta-V missions in cislunar space. Long term applications of the CNTR could include the advanced exploration and utilization of the solar system through direct use of in-situ volatiles as propellant. Challenges associated with the CNTR are numerous. Centrifugal force is used to retain the liquid fuel in rotating fuel cylinders, and rotational velocities up to 5000 rpm may be required. Propellant flow must be directed such that all structures and moderators in the core are adequately cooled prior to the propellant entering the liquid fuel and being heated to 5000 K. The rotating fuel cylinder wall (RFCW) must have an inner surface designed to be compatible with liquid uranium metal or uranium carbide fuel up to at least 1500 K, and that inner surface may also need to be textured to help maintain acceptable wall temperatures. Propellant must flow radially inward through the RFCW while fuel is simultaneously contained. The RFCW should ideal-ly be made from a material with low neutron absorption to help minimize engine mass and facilitate the use of High Assay Low Enriched Uranium (HALEU) fuel in the system. The drive system for the rotating fuel cylinders must support all phases of operation. This paper will discuss computational and experimental research being conducted to address some of the challenges associated with the CNTR, and will also note potential mission benefits from the CNTR.

Michael G. Houts↗