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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 NTRS1976

Thermal energy storage material thermophysical property measurement and heat transfer impact

The thermophysical properties of salts having potential for thermal energy storage to provide peaking energy in conventional electric utility power plants were investigated. The power plants studied were the pressurized water reactor, boiling water reactor, supercritical steam reactor, and high temperature gas reactor. The salts considered were LiNO3, 63LiOH/37 LiCl eutectic, LiOH, and Na2B4O7. The thermal conductivity, specific heat (including latent heat of fusion), and density of each salt were measured for a temperature range of at least + or - 100 K of the measured melting point. Measurements were made with both reagent and commercial grades of each salt.

Tye, R. P.↗
NASA NTRS2007

Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is expected within the next 30 to 50 years, as predicted by the Hubbert model and confirmed by other global energy consumption prognoses. Having invested national resources into the development of NGNP, the technology and experience accumulated during the project needs to be documented clearly and in sufficient detail for young engineers coming on-board at both DOE and NASA to acquire it. Hands on training on reactor operation, test rigs of turbomachinery, and heat exchanger components, as well as computational tools will be needed. Senior scientist/engineers involved with the development of NGNP should also be encouraged to participate as lecturers, instructors, or adjunct professors at local universities having engineering (mechanical, electrical, nuclear/chemical, and/or materials) as one of their fields of study.

Juhasz, Albert J.↗
NASA NTRS1999

Power System Optimization Codes Modified

A major modification of and addition to existing Closed Brayton Cycle (CBC) space power system optimization codes was completed. These modifications relate to the global minimum mass search driver programs containing three nested iteration loops comprising iterations on cycle temperature ratio, and three separate pressure ratio iteration loops--one loop for maximizing thermodynamic efficiency, one for minimizing radiator area, and a final loop for minimizing overall power system mass. Using the method of steepest ascent, the code sweeps through the pressure ratio space repeatedly, each time with smaller iteration step sizes, so that the three optimum pressure ratios can be obtained to any desired accuracy for each of the objective functions referred to above (i.e., maximum thermodynamic efficiency, minimum radiator area, and minimum system mass). Two separate options for the power system heat source are available: 1. A nuclear fission reactor can be used. It is provided with a radiation shield 1. (composed of a lithium hydride (LiH) neutron shield and tungsten (W) gamma shield). Suboptions can be used to select the type of reactor (i.e., fast spectrum liquid metal cooled or epithermal high-temperature gas reactor (HTGR)). 2. A solar heat source can be used. This option includes a parabolic concentrator and heat receiver for raising the temperature of the recirculating working fluid. A useful feature of the code modifications is that key cycle parameters are displayed, including the overall system specific mass in kilograms per kilowatt and the system specific power in watts per kilogram, as the results for each temperature ratio are computed. As the minimum mass temperature ratio is encountered, a message is printed out. Several levels of detailed information on cycle state points, subsystem mass results, and radiator temperature profiles are stored for this temperature ratio condition and can be displayed or printed by users.

Juhasz, Albert J.↗
NASA NTRS1992

Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications

Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

Juhasz, Albert J.↗
NASA NTRS1998

Impact of Radiation Hardness and Operating Temperatures of Silicon Carbide Electronics on Space Power System Mass

The effect of silicon carbide (SiC) electronics operating temperatures on Power Management and Distribution (PMAD), or Power Conditioning (PC), subsystem radiator size and mass requirements was evaluated for three power output levels (100 kW(e) , 1 MW(e), and 10 MW(e)) for near term technology ( i.e. 1500 K turbine inlet temperature) Closed Cycle Gas Turbine (CCGT) power systems with a High Temperature Gas Reactor (HTGR) heat source. The study was conducted for assumed PC radiator temperatures ranging from 370 to 845 K and for three scenarios of electrical energy to heat conversion levels which needed to be rejected to space by means of the PC radiator. In addition, during part of the study the radiation hardness of the PC electronics was varied at a fixed separation distance to estimate its effect on the mass of the instrument rated reactor shadow shield. With both the PC radiator and the conical shadow shield representing major components of the overall power system the influence of the above on total power system mass was also determined. As expected, results show that the greatest actual mass savings achieved by the use of SiC electronics occur with high capacity power systems. Moreover, raising the PC radiator temperature above 600 K yields only small additional system mass savings. The effect of increased radiation hardness on total system mass is to reduce system mass by virtue of lowering the shield mass.

Juhasz, Albert J.↗
NASA NTRS2015

Molecular Tagging Velocimetry Development for In-situ Measurement in High-Temperature Test Facility

The High Temperature Test Facility, HTTF, at Oregon State University (OSU) is an integral-effect test facility designed to model the behavior of a Very High Temperature Gas Reactor (VHTR) during a Depressurized Conduction Cooldown (DCC) event. It also has the ability to conduct limited investigations into the progression of a Pressurized Conduction Cooldown (PCC) event in addition to phenomena occurring during normal operations. Both of these phenomena will be studied with in-situ velocity field measurements. Experimental measurements of velocity are critical to provide proper boundary conditions to validate CFD codes, as well as developing correlations for system level codes, such as RELAP5 (http://www4vip.inl.gov/relap5/). Such data will be the first acquired in the HTTF and will introduce a diagnostic with numerous other applications to the field of nuclear thermal hydraulics. A laser-based optical diagnostic under development at The George Washington University (GWU) is presented; the technique is demonstrated with velocity data obtained in ambient temperature air, and adaptation to high-pressure, high-temperature flow is discussed.

Andre, Matthieu A.↗
NASA NTRS1991

Nuclear fuels status

A discussion on coated particle fuel performance from a modular High Temperature Gas Reactor (HTGR) is presented along with experimental results. The following topics are covered: (1) the coated particle fuel concept; (2) the functional requirements; (3) performance limiting mechanisms; (4) fuel performance; and (5) methods/techniques for characterizing performance.

Kania, Michael↗
NASA NTRS1983

Energetic closed-cycle gas core reactors for orbit raising

Closed-cycle gas core reactor power plants can be of two types. In the 'mixed flow' type, the gaseous nuclear fuel is intimately mixed with the working gas in the cavity. In the 'light bulb' type the fissioning plasma is enclosed in a transparent tube, and energy transfer to the separate working gas occurs by thermal radiation. The potentials of high temperature gas core reactors in terrestrial electric power generator applications have been considered, and a number of civilian power-beaming applications for gaseous fuel nuclear-MHD power plants in space have been suggested. Major conclusions of investigations related to the design of space power systems are discussed. Attention is given to options for conversion cycles, the power system specific mass, and research and technology issues.

Rosa, R. J.↗
NASA NTRS1975

Studies of the use of high-temperature nuclear heat from an HTGR for hydrogen production

The results of a study which surveyed various methods of hydrogen production using nuclear and fossil energy are presented. A description of these methods is provided, and efficiencies are calculated for each case. The process designs of systems that utilize the heat from a general atomic high temperature gas cooled reactor with a steam methane reformer and feed the reformer with substitute natural gas manufactured from coal, using reforming temperatures, are presented. The capital costs for these systems and the resultant hydrogen production price for these cases are discussed along with a research and development program.

Peterman, D. D.↗
NASA NTRS1976

Energy Conversion Alternatives Study (ECAS), Westinghouse phase 1. Volume 9: Closed-cycle MHD

A closed-cycle MHD system for an electric power plant was studied. It consists of 3 interlocking loops, an external heating loop, a closed-cycle cesium seeded argon nonequilibrium ionization MHD loop, and a steam bottomer. A MHD duct maximum temperature of 2366 K (3800 F), a pressure of 0.939 MPa (9.27 atm) and a Mach number of 0.9 are found to give a topping cycle efficiency of 59.3%; however when combined with an integrated gasifier and optimistic steam bottomer the coal to bus bar efficiency drops to 45.5%. A 1978 K (3100 F) cycle has an efficiency of 55.1% and a power plant efficiency of 42.2%. The high cost of the external heating loop components results in a cost of electricity of 21.41 mills/MJ (77.07 mills/kWh) for the high temperature system and 19.0 mills/MJ (68.5 mills/kWh) for the lower temperature system. It is, therefore, thought that this cycle may be more applicable to internally heated systems such as some futuristic high temperature gas cooled reactor.

Tsu, T. C.↗
NASA NTRS1993

Scoping calculations of power sources for NEP

Three Nuclear Electric Propulsion (NEP) systems are evaluated in the context of the two following criteria: (1) power levels (P)--10-50 Mw; and (2) core life (D)--2-10 yrs. The three types of reactors are as follows: (1) high temperature gas-cooled reactors of the NERVA derivative type; (2) lithium-cooled advanced fuel pin--one-phase flow; and (3) lithium-cooled Cermet--one-phase flow. The discussion is presented in vugraph form.

Difilippo, Felix C.↗
NASA NTRS1979

Dedicated nuclear facilities for electrolytic hydrogen production

An advanced technology, fully dedicated nuclear-electrolytic hydrogen production facility is presented. This plant will produce hydrogen and oxygen only and no electrical power will be generated for off-plant use. The conceptual design was based on hydrogen production to fill a pipeline at 1000 psi and a 3000 MW nuclear base, and the base-line facility nuclear-to-shaftpower and shaftpower-to-electricity subsystems, the water treatment subsystem, electricity-to-hydrogen subsystem, hydrogen compression, efficiency, and hydrogen production cost are discussed. The final conceptual design integrates a 3000 MWth high-temperature gas-cooled reactor operating at 980 C helium reactor-out temperature, direct dc electricity generation via acyclic generators, and high-current density, high-pressure electrolyzers based on the solid polymer electrolyte approach. All subsystems are close-coupled and optimally interfaced and pipeline hydrogen is produced at 1000 psi. Hydrogen costs were about half of the conventional nuclear electrolysis process.

Foh, S. E.↗
NASA NTRS1975

A preliminary systems-engineering study of an advanced nuclear-electrolytic hydrogen-production facility

An advanced nuclear-electrolytic hydrogen-production facility concept was synthesized at a conceptual level with the objective of minimizing estimated hydrogen-production costs. The concept is a closely-integrated, fully-dedicated (only hydrogen energy is produced) system whose components and subsystems are predicted on ''1985 technology.'' The principal components are: (1) a high-temperature gas-cooled reactor (HTGR) operating a helium-Brayton/ammonia-Rankine binary cycle with a helium reactor-core exit temperature of 980 C, (2) acyclic d-c generators, (3) high-pressure, high-current-density electrolyzers based on solid-polymer electrolyte technology. Based on an assumed 3,000 MWt HTGR the facility is capable of producing 8.7 million std cu m/day of hydrogen at pipeline conditions, 6,900 kPa. Coproduct oxygen is also available at pipeline conditions at one-half this volume. It has further been shown that the incorporation of advanced technology provides an overall efficiency of about 43 percent, as compared with 25 percent for a contemporary nuclear-electric plant powering close-coupled contemporary industrial electrolyzers.

Escher, W. J. D.↗
NASA NTRS1976

Hydrogen production from coal using a nuclear heat source

A strong candidate for hydrogen production in the intermediate time frame of 1985 to 1995 is a coal-based process using a high-temperature gas-cooled reactor (HTGR) as a heat source. Expected process efficiencies in the range of 60 to 70% are considerably higher than all other hydrogen production processes except steam reforming of a natural gas. The process involves the preparation of a coal liquid, hydrogasification of that liquid, and steam reforming of the resulting gaseous or light liquid product. A study showing process efficiency and cost of hydrogen vs nuclear reactor core outlet temperature has been completed, and shows diminishing returns at process temperatures above about 1500 F. A possible scenario combining the relatively abundant and low-cost Western coal deposits with the Gulf Coast hydrogen users is presented which provides high-energy density transportation utilizing coal liquids and uranium.

Quade, R. N.↗
NASA NTRS1977

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.↗
NASA NTRS2012

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.↗
NASA NTRS1996

A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high efficiency in the gas core reactors. The model is also used to predict the convective and radiation heat fluxes for the gas core reactors. The maximum value of heat flux occurs at the exit of the reactor core. Radiation heat flux increases with higher wall temperature. This behavior is due to the fact that the radiative heat flux is strongly dependent on wall temperature. This study also found that at temperature close to 3500 K the radiative heat flux is comparable with the convective heat flux in a uranium fluoride failed gas core reactor.

Anghaie, S.↗
NASA NTRS2005

Modeling and Testing of Non-Nuclear, Highpower Simulated Nuclear Thermal Rocket Reactor Elements

When the President offered his new vision for space exploration in January of 2004, he said, "Our third goal is to return to the moon by 2020, as the launching point for missions beyond," and, "With the experience and knowledge gained on the moon, we will then be ready to take the next steps of space exploration: human missions to Mars and to worlds beyond." A human mission to Mars implies the need to move large payloads as rapidly as possible, in an efficient and cost-effective manner. Furthermore, with the scientific advancements possible with Project Prometheus and its Jupiter Icy Moons Orbiter (JIMO), (these use electric propulsion), there is a renewed interest in deep space exploration propulsion systems. According to many mission analyses, nuclear thermal propulsion (NTP), with its relatively high thrust and high specific impulse, is a serious candidate for such missions. Nuclear rockets utilize fission energy to heat a reactor core to very high temperatures. Hydrogen gas flowing through the core then becomes superheated and exits the engine at very high exhaust velocities. The combination of temperature and low molecular weight results in an engine with specific impulses above 900 seconds. This is almost twice the performance of the LOX/LH2 space shuttle engines, and the impact of this performance would be to reduce the trip time of a manned Mars mission from the 2.5 years, possible with chemical engines, to about 12-14 months.

Kirk, Daniel R.↗