Mass-retention measurements in a binary compressible vortex flow
Binary vortex flow fields used as gaseous-fuel- retention devices for gas-core nuclear reactors
REACTOR↗
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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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Computer program for calculating isothermal, turbulent jet mixing of two gases
Computer program for calculating isothermal, turbulent jet mixing flow of two gases in coaxial flow, gas gore nuclear reactors
Donovan, L. F.↗
A mini-cavity probe reactor.
The mini-cavity reactor is a rocket engine concept which combines the high specific impulse from a central gaseous fueled cavity (0.6 m diam) and NERVA type fuel elements in a driver region that is external to a moderator-reflector zone to produce a compact light weight reactor. The overall dimension including a pressure vessel that is located outside of the spherical reactor is approximately 1.21 m in diameter. Specific impulses up to 2000 sec are obtainable for 220 to 890 N of thrust with pressures less than 1000 atm. Powerplant weights including a radiator for disposing of the power in the driver region are between 4600 and 32,000 kg - less than payloads of the shuttle. This reactor could also be used as a test reactor for gas-core, MHD, breeding and materials research.
Hyland, R. E.↗
Benchmark gas core critical experiment.
A critical experiment with spherical symmetry has been conducted on the gas core nuclear reactor concept. The nonspherical perturbations in the experiment were evaluated experimentally and produce corrections to the observed eigenvalue of approximately 1% delta k. The reactor consisted of a low density, central uranium hexafluoride gaseous core, surrounded by an annulus of void or low density hydrocarbon, which in turn was surrounded with a 97-cm-thick heavy water reflector.
Kunze, J. F.↗
Instabilities in fissioning plasmas as applied to the gas-core nuclear rocket-engine
The compressional wave spectrum excited in a fissioning uranium plasma confined in a cavity such as a gas cored nuclear reactor, is studied. Computer results are presented that solve the fluid equations for this problem including the effects of spatial gradients, nonlinearities, and neutron density gradients in the reactor. Typically the asymptotic fluctuation level for the plasma pressure is of order 1 percent.
Source record↗
Nuclear Analysis
Exploratory calculations were performed for several gas core breeder reactor configurations. The computational method involved the use of the MACH-1 one dimensional diffusion theory code and the THERMOS integral transport theory code for thermal cross sections. Computations were performed to analyze thermal breeder concepts and nonbreeder concepts. Analysis of breeders was restricted to the (U-233)-Th breeding cycle, and computations were performed to examine a range of parameters. These parameters include U-233 to hydrogen atom ratio in the gaseous cavity, carbon to thorium atom ratio in the breeding blanket, cavity size, and blanket size.
Clement, J. D.↗
Investigation B: Laminar confined coaxial entrance flow with heat generation
The results of a parametric study on the entrance flow region in a gas core nuclear reactor are presented. The physical system is modeled as laminar confined, coaxial flow with heat generation in the inner fluid. The governing equations include the boundary layer approximations and the assumptions of only radial radiative transport of energy represented as an energy diffusion term. The Von Mises transformation and a zeta transformation are used to transform the equations into nonlinear nonhomogeneous convective-diffusion equations. A unique combination of forward and backward difference equations which yields accurate results at moderate computational times, is used in the numerical method. Results show that the rapidly accelerating, heat generating inner stream actually shrinks in radius as it expands axially.
Bobba, G. K. M.↗
Helium heater design for the helium direct cycle component test facility
The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.
Larson, V. R.↗
Production of hydrogen by direct gasification of coal with steam using nuclear heat
Problems related to: (1) high helium outlet temperature of the reactor, and (2) gas generator design used in hydrogen production are studied. Special attention was given to the use of Oklahoma coal in the gasification process. Plant performance, operation, and environmental considerations are covered.
Source record↗
Mechanical design
Design concepts for a 1000 mw thermal stationary power plant employing the UF6 fueled gas core breeder reactor are examined. Three design combinations-gaseous UF6 core with a solid matrix blanket, gaseous UF6 core with a liquid blanket, and gaseous UF6 core with a circulating blanket were considered. Results show the gaseous UF6 core with a circulating blanket was best suited to the power plant concept.
Source record↗
Efficiency and cost advantages of an advanced-technology nuclear electrolytic hydrogen-energy production facility
The concept of an advanced-technology (viz., 1985 technology) nuclear-electrolytic water electrolysis facility was assessed for hydrogen production cost and efficiency expectations. The facility integrates (1) a high-temperature gas-cooled nuclear reactor (HTGR) operating a binary work cycle, (2) direct-current (d-c) electricity generation via acyclic generators, and (3) high-current-density, high-pressure electrolyzers using a solid polymer electrolyte (SPE). All subsystems are close-coupled and optimally interfaced for hydrogen production alone (i.e., without separate production of electrical power). Pipeline-pressure hydrogen and oxygen are produced at 6900 kPa (1000 psi). We found that this advanced facility would produce hydrogen at costs that were approximately half those associated with contemporary-technology nuclear electrolysis: $5.36 versus $10.86/million Btu, respectively. The nuclear-heat-to-hydrogen-energy conversion efficiency for the advanced system was estimated as 43%, versus 25% for the contemporary system.
Donakowski, T. D.↗
Comparative description of coal feeding systems for fixed bed pressure gasification
Coal feeding systems are discussed which are capable of feeding 20-100 T/H and the range of pressure is up to 100 bar. Most emphasis is placed on dry feeding systems. The systems outlined are subdivided into continuous and intermittent and the influence of each system on lock gas losses and reactor design is shown. Finally a cost estimate is presented which indicates the areas of preferred application and permits conclusions to be drawn regarding the economics of the various systems.
Funk, E.↗
Potentials of fissioning plasmas
Successful experiments with the nuclear pumping of lasers have demonstrated that in a gaseous medium the kinetic energy of fission fragments can be converted directly into nonequilibrium optical radiation. This confirms the concept that the fissioning medium in a gas-phase nuclear reactor shows an internal structure such as a plasma in near thermal equilibrium varying up to a state of extreme nonequilibrium. During 20 years of research under NASA support major elements of the fissioning plasma reactor were demonstrated in theory and experiment, culminating in a proof-of-principle reactor test conducted at the Los Alamos Scientific Laboratory. It is concluded that the construction of a gaseous fuel reactor power plant is within the reach of present technology.
Thom, K.↗
Fluidized bed desulfurization
High sulfur content carbonaceous material, such as coal is desulfurized by continuous fluidized suspension in a reactor with chlorine gas, inert dechlorinating gas and hydrogen gas. A source of chlorine gas, a source of inert gas and a source of hydrogen gas are connected to the bottom inlet through a manifold and a heater. A flow controler operates servos in a manner to continuously and sequentially suspend coal in the three gases. The sulfur content is reduced at least 50% by the treatment.
Ravindram, M.↗
Development of an Advanced Computational Model for OMCVD of Indium Nitride
An advanced computational model is being developed to predict the formation of indium nitride (InN) film from the reaction of trimethylindium (In(CH3)3) with ammonia (NH3). The components are introduced into the reactor in the gas phase within a background of molecular nitrogen (N2). Organometallic chemical vapor deposition occurs on a heated sapphire surface. The model simulates heat and mass transport with gas and surface chemistry under steady state and pulsed conditions. The development and validation of an accurate model for the interactions between the diffusion of gas phase species and surface kinetics is essential to enable the regulation of the process in order to produce a low defect material. The validation of the model will be performed in concert with a NASA-North Carolina State University project.
Cardelino, Carlos A.↗
Hardware Progress Made in the Early Flight Fission Test Facilities (EFF-TF) To Support Near-Term Space Fission Systems
The EFF-TF provides a facility to experimentally evaluate thermal hydraulic issues through the use of highly effective non-nuclear testing. These techniques provide a rapid, more cost effective method of evaluating designs and support development risk mitigation when concerns are associated with non-nuclear aspects of space nuclear systems. For many systems, electrical resistance thermal simulators can be used to closely mimic the heat deposition of the fission process, providing axial and radial profiles. A number of experimental and design programs were underway in 2004. Initial evaluation of the SAFE-100a (19 module stainless steel/sodium heat pipe reactor with integral gas neat exchanger) was performed with tests up to 17.5 kW of input power at core temperatures of 1000 K. A stainless steel sodium SAFE-100 heat pipe module was placed through repeated freeze/thaw cyclic testing accumulating over 200 restarts to a temperature of 1000 K. Additionally, the design of a 37-fuel pin stainless steel pumped sodium/potassium (NaK) loop was finalized and components procured. Ongoing testing at the EFF-TF is geared towards facilitating both research and development necessary to field a near term space nuclear system. Efforts are coordinated with DOE laboratories, industry, universities, and other NASA centers. This paper describes some of the 2004 efforts.
VanDyke, Melissa↗
RESOLVE 2010 Field Test
This slide presentation reviews the field tests conducted in 2010 of the Regolith Environment Science & Oxygen & Lunar Volatile Extraction (RESOLVE). The Resolve program consist of several mechanism: (1) Excavation and Bulk Regolith Characterization (EBRC) which is designed to act as a drill and crusher, (2) Regolith Volatiles Characterization (RVC) which is a reactor and does gas analysis,(3) Lunar Water Resources Demonstration (LWRD) which is a fluid system, water and hydrogen capture device and (4) the Rover. The scientific goal of this test is to demonstrate evolution of low levels of hydrogen and water as a function of temperature. The Engineering goals of this test are to demonstrate:(1) Integration onto new rover (2) Miniaturization of electronics rack (3) Operation from battery packs (elimination of generator) (4) Remote command/control and (5) Operation while roving. Views of the 2008 and the 2010 mechanisms, a overhead view of the mission path, a view of the terrain, the two drill sites, and a graphic of the Master Events Controller Graphical User Interface (MEC GUI) are shown. There are descriptions of the Gas chromatography (GC), the operational procedure, water and hydrogen doping of tephra. There is also a review of some of the results, and future direction for research and tests.
Captain, J.↗
Large-Scale Weibull Analysis of H-451 Nuclear- Grade Graphite Specimen Rupture Data
A Weibull analysis was performed of the strength distribution and size effects for 2000 specimens of H-451 nuclear-grade graphite. The data, generated elsewhere, measured the tensile and four-point-flexure room-temperature rupture strength of specimens excised from a single extruded graphite log. Strength variation was compared with specimen location, size, and orientation relative to the parent body. In our study, data were progressively and extensively pooled into larger data sets to discriminate overall trends from local variations and to investigate the strength distribution. The CARES/Life and WeibPar codes were used to investigate issues regarding the size effect, Weibull parameter consistency, and nonlinear stress-strain response. Overall, the Weibull distribution described the behavior of the pooled data very well. However, the issue regarding the smaller-than-expected size effect remained. This exercise illustrated that a conservative approach using a two-parameter Weibull distribution is best for designing graphite components with low probability of failure for the in-core structures in the proposed Generation IV (Gen IV) high-temperature gas-cooled nuclear reactors. This exercise also demonstrated the continuing need to better understand the mechanisms driving stochastic strength response. Extensive appendixes are provided with this report to show all aspects of the rupture data and analytical results.
Nemeth, Noel N.↗