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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
NASA NTRS1985

Alkali Metal Thermoelectric Conversion (AMTEC) for space nuclear power systems

Performance parameters of the Alkali Metal Thermoelectric Converter (AMTEC) for a 100 kW electric power system have been calculated at four technological levels assuming a heat pipe-cooled nuclear reactor heat source. The most advanced level considered would operate between 1180 K converter temperature and 711 K radiator temperature at 16 percent efficiency, and would weigh 1850 kg with a radiator area of 43 sq m. In addition, electrode research studies for the AMTEC systems have been conducted utilizing an experimental test cell of Bankston et al. (1983) and Mo and several Mo-Ti electrodes. It was found that the Mo-Ti electrodes offered no improvement in lifetime characteristics over the pure Mo electrodes, however, oxygen treatment of a degraded Mo electrode restored its specific power output to 90 percent of its original specific power and maintained this level for 60 hr, thus offering a potential for lifetime stability.

Bankston, C. P.↗
NASA NTRS1972

Hot-gas-side heat transfer with and without film cooling on a simulated nuclear rocket thrust chamber using H2-O2

Heat-transfer coefficients were obtained on a thrust chamber which simulated the geometry of the NERVA nuclear rocket. The tests were performed with and without peripheral film cooling over a chamber pressure range of 1.05 million to 5.84 million newtons per square meter (153 to 847 psia). With no film cooling, the overall axial variation in the value of the correlation coefficient C of the equation (Stanton)* (Prandtl)* to the 0.7ths power = C(Reynolds)* to the -0.2ths power, where * indicates the reference enthalpy condition, was reduced 66 percent when the local diameter in the Reynolds number was replaced by the axial distance from the injector face. The average peak values of C were reduced 25 percent with 2 and 3.75 percent cooling and 50 percent with 7.5 percent cooling.

Quentmeyer, R. J.↗
DOE OSTI2021

Stochastically estimated covariance matrices for independent and cumulative fission yields in the ENDF/B-VIII.0 and JEFF-3.3 evaluations

A Monte-Carlo method for the generation of correlation and covariance matrices for independent and cumulative fission yields has been developed. The method uses a constrained Monte-Carlo resampling structure in order to vary evaluated fission yield libraries in a way that meets basic conservation principles. This results in the generation of correlation/covariance matrices with limited model bias and uncertainty; the matrices are primarily reflective of the evaluated fission yield uncertainties and correlations that arise from the evaluation process. This method has been applied to generate correlation and covariance matrices for all of the fissioning systems of the ENDF/B-VIII.0 and JEFF-3.3 evaluations, marking the first time such matrices have been generated for all of these systems. These covariance matrices have been published online for immediate public use. These correlation and covariance matrices can be used to improve uncertainty estimation in calculations of reactor antineutrino emission rates, decay heat problems, and nuclear forensics.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Competitiveness Assessment of Decarbonizing Electricity and Process Heat Supply to a Campus with a Small Nuclear Reactor

This paper analyzes the competitiveness of siting a small nuclear reactor to support decarbonization of sites requiring tens of MW of electricity and/or process heat to support centralized heating and cooling system. This paper focuses on campuses as representative of sites with collections of buildings and research facilities with decarbonization needs represented by buildings heating, and electricity consumption by electrical loads which may include cooling via chilled water (e.g., for air conditioning and to cool down computer clusters). A nuclear reactor can be considered to decarbonize a site’s high-temperature steam generation used mostly for building heating needs, climate control, and hot water, by supplying process heat capabilities, while electricity decarbonization would be achieved mostly by the grid. However, a secondary application can be considered to maximize reactor utilization and avoid ramping down the reactor if the steam demand varies significantly throughout the year. Chilled water generation through steam-driven systems was identified as an attractive secondary option for the site analyzed, due to potential for plant design simplification, while electricity generation could be considered as well to reduce electricity purchases for a wider range of site applications. For a campus with peak 60MW thermal power demand, a small nuclear reactor with similar thermal power rating would almost eliminate CO2 emissions from steam generation and reduce electricity imports for chilled water production. A preliminary techno-economic feasibility study shows that a small nuclear reactor design that is optimized to support process heat can represent an economically feasible option when compared with other decarbonization alternatives.

Stauff, Nicolas E.↗
DOE OSTI2022

Controlling a power output of a nuclear reaction without control rods

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above the reactor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Callaway, Allyson↗
DOE OSTI2024

Controlling a power output of a nuclear reactor without using control rods

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above there actor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Callaway, Allyson↗
DOE OSTI2022

Controlling a power output of a nuclear reaction without control rods

A nuclear power system includes a reactor vessel that includes a reactor core mounted therein. The reactor core includes nuclear fuel assemblies configured to generate a nuclear fission reaction. The reaction vessel does not include any control rod assemblies therein. The nuclear power system further includes a riser positioned above the reactor core, a primary coolant flow path, a primary coolant that circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the received heat to generate electric power in a power generation, and a control system communicably coupled to the power generation system and configured to control a power output of the nuclear fission reaction independent of any control rod assemblies.

Callaway, Allyson↗
NASA NTRS1985

Numerical simulation of comet nuclei. I - Water-ice comets

A one-dimensional numerical model of pure water-ice cometary nuclei is presented, and the influence of the nuclear interior as a heat reservoir on the behavior of the nuclear surface is examined. It is shown that a number of effects, including the thermal inertia due to heat stored in the core and the release of latent heat, which goes entirely into heating the adjacent layers or into sublimation on passing through a phase transition from amorphous to crystalline ice, can help to explain such characteristics as the asymmetrical lightcurve of Comet Halley. Results are given for the cases of Comet Schwassmann-Wachmann 1 and Comet Encke. Consideration is also given to the insulating effect of an evolving dust mantle. The role of this mantle in determining the surface temperature of the ice core is studied as a function of the mass fraction of the dust in the ice-dust mixture and the thermal conductivity of the nucleus. The loose-lattice model of Mendis and Brin (1977) indicates that both high dust to ice ratios and high-core conductivities inhibit mantle blowoff.

Herman, G.↗
DOE OSTI2023

Thermal Analysis of the SIRIUS3 Nuclear Propulsion Fuel Calibration Experiment

NASA is considering Nuclear Thermal Propulsion (NTP) for long range extraterrestrial missions; these engines eject hot hydrogen gas heated by a nuclear reactor for rocket thrust. To economize the use of hydrogen to the greatest extent possible, the NTP engines will be expected to, in a very short time (i.e., on the order of a minute or less), go from warm$(\sim 300 K)$ zero power conditions to full operational power, with a coolant outlet temperature on the order of 2700–3000 K \cite{en15176181}. These conditions will introduce significant thermomechanical stresses on the NTP fuel. The objective of the SIRIUS series of experiments is to examine the performance of candidate NTP fuel materials when subjected to temperature ramp rates that are prototypical of NTP system startup and operation. The SIRIUS experiments are a series of experiments that will be irradiated in the TREAT reactor and are subjected to power ramps and cycles that are prototypical for NTP operation. The experiments will be accomplished by executing a series of shaped transients on the SIRIUS specimens while collecting in situ specimen temperature data. These tests will determine whether operational startup ramps and peak temperatures will result in detrimental fuel performance phenomena (i.e. fuel deformation, fragmentation and cracking) To this end, INL has been evaluating a number of SIRIUS experiments, and these evaluations include thermal analysis of the SIRIUS experiments. Thermal analysis were performed for the calibration irradiation of the SIRIUS-3 experiment, and the focus of this memo is to document the results from the calibration irradiation thermal analysis. The thermal analysis reveals that radiation heat emission of the outer fuel elements and conduction to different metal components such as the molybdenum element tubes remove significant quantities of heat from the fuel element specimen, and future experiment designs need to consider these heat transfer mechanics. This paper begins with a brief experiment overview with a discussion of the fuel sample and experiment configuration. The experiment and model description section is followed by a set of results with a brief discussion, and finally the memo concludes with suggestion for future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

3D modelling of MHD mixed convection flow in a vertical duct with transverse magnetic field and volumetric or surface heating

Understanding phenomena associated with the multiple effects/interactions of the fusion nuclear environment on liquid metal flow is required to correctly design liquid metal (LM) blankets for fusion facilities. These effects are investigated in the present work by numerically simulating 3D LM MHD flow. The simulated geometry consists of a straight, vertical duct which runs perpendicular to a strong, fringing applied magnetic field. There is also a region of applied heating as the primary goal is to explore buoyancy effects in MHD duct flows. Results are presented for both buoyancy assisted (upwards) and buoyancy opposed (downwards) flows in conducting and insulating ducts for a range of Hartmann numbers (Ha) up to 100, Reynolds numbers (Re) from 10 3 to 10 4 and Grashof (Gr) numbers from 10 7 to 10 8 . While increasing Gr or decreasing Re increases buoyancy effects, increasing Ha was shown to increase maximum temperature through turbulence reduction. The extent to which the MHD mixed convection flows are quasi-2D is analyzed and buoyant effects, in competition with electromagnetic forces, are shown to bring about 3D flow features not seen in purely MHD flows. Volumetric nuclear heating with steep gradients is applied to the vertical MHD flows for comparison to flows with surface heating only. Finally, surface heating generates stronger buoyancy effects than volumetric heating of the same total power; however, many of the same phenomena occur. Therefore, surface heating, the only option for lab experiments, can provide indication of the effects of volumetric heating in MHD flows.

42 ENGINEERING↗
NASA NTRS1976

System Design for a Nuclear Electric Spacecraft Utilizing Out-of-core Thermionic Conversion

Basic guidelines are presented for a nuclear space power system which utilizes heat pipes to transport thermal power from a fast nuclear reactor to an out of core thermionic converter array. Design parameters are discussed for the nuclear reactor, heat pipes, thermionic converters, shields (neutron and gamma), waste heat rejection systems, and the electrical bus bar-cable system required to transport the high current/low voltage power to the processing equipment. Dimensions are compatible with shuttle payload bay constraints.

Estabrook, W. C.↗
NASA NTRS1971

External fuel thermionic reactor system.

Thermionic reactors are prime candidates for nuclear electric propulsion. The national thermionic reactor effort is concentrated on the flashlight concept with the external-fuel concept as the backup. The external-fuel concept is very adaptable to a completely modular power subsystem which is attractive for highly reliable long-life applications. The 20- to 25-cm long, externally-fueled converters have been designed, fabricated, and successfully tested with many thermal cycles by electrical heating. However, difficulties have been encountered during encapsulation for nuclear heated tests and none have been started to date. These nuclear tests are required to demonstrate the concept feasibility.

Mondt, J. F.↗
DOE OSTI2025

Smaller and faster: a review of conventional and nanocalorimetry techniques for determining thermophysical properties of nuclear materials

Thermal analysis of nuclear materials is critical for the advancement of nuclear technology. The heat effects associated with heat capacity, phase transformation, and radiation damage can be measured with conventional calorimeters. However, conventional calorimetric techniques are often restricted in terms of heating rate and sample mass, especially when studying the limited amounts of materials subject to extreme conditions. In this review, we summarize conventional calorimetric studies of critical thermophysical and thermochemical properties of pure actinide metals (U, Np, Am, Pu), fast reactor metallic fuel alloy systems (U–Zr, U–Pu–Zr, Pu–U, Pu–Zr), and actinide oxides that are primary constituents or transmutation products in light water reactor fuel rods (U–O, Np–O, Am–O, Pu–O, Pu–U–O). Adiabatic and drop calorimetry have been the primary techniques used for these studies, however the development of fast scanning calorimetry using micro-electro-mechanical-based systems allows determination of thermodynamic properties from smaller sample masses. We report recent investigations that leverage the fast heating rates of nanocalorimetry by itself or combined with other characterization techniques. Furthermore, we then discuss opportunities for nanocalorimetry to provide solutions to some of the technical challenges inherent in thermal analysis of nuclear materials, namely a reduction in sample activity, emulating heating transients, investigation of phase evolution in irradiated samples, and characterization of radiation damage evolution. Nanocalorimetry has the potential to significantly advance the understanding of thermophysical properties in nuclear materials and thus accelerate the development of nuclear technology.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗