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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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13 records
All sourcesNASA NTRSDOE OSTI
NASA NTRS1991

Graphite Fiber/Copper Matrix Composites for Space Power Heat Pipe Fin Applications

High specific thermal conductivity (thermal conductivity divided by density) is a major design criterion for minimizing system mass for space power systems. For nuclear source power systems, graphite fiber reinforced copper matrix (Gr/Cu) composites offer good potential as a radiator fin material operating at service temperatures above 500 K. Specific thermal conductivity in the longitudinal direction is better than beryllium and almost twice that of copper. The high specific thermal conductivity of Gr/Cu offers the potential of reducing radiator mass by as much as 30 percent. Gr/Cu composites also offer the designer a range of available properties for various missions and applications. The properties of Gr/Cu are highly anisotropic. Longitudinal elastic modulus is comparable to beryllium and about three times that of copper. Thermal expansion in the longitudinal direction is near zero, while it exceeds that of copper in the transverse direction.

McDanels, David L.↗
NASA NTRS2003

ATIC Experiment: Elemental Spectra from the Flight in 2000

The Advanced Thin Ionization Calorimeter (ATIC) had successful Long Duration Balloon flights from McMurdo, Antarctica in both 2000 and 2002. The instrument consists of a silicon matrix charge detector, a 0.75 nuclear interaction length graphite target, 3 scintillator strip hodoscopes, and an 18 radiation length thick BGO calorimeter to measure the cosmic ray composition and energy spectra from approximately 30 GeV to near 100 TeV. In this paper, we present preliminary results from the first flight, which was a test flight that lasted for 16 days, starting on 12/28/00.

Ahn, H. S.↗
NASA NTRS2004

The Energy Spectra of Proton and Helium Measured from the ATIC Experiment

The Advanced Thin Ionization Calorimeter (ATIC) balloon experiment is designed to investigate the composition and energy spectra of cosmic rays at the highest energies currently accessible from direct measurements, the region up to 100 TeV. The instrument consists of a silicon matrix for charge measurement, a graphite target (0.75 nuclear interaction length) to induce hadronic fragmentation, 3 scintillator strip hodoscopes for triggering and helping reconstruct trajectory, and a BGO calorimeter (18 radiation lengths) to measure the energy of incident particles. ATIC had two successful Long Duration Balloon (LDB) flights from McMurdo, Antarctica: from 12/28/00 to 01/13/01 and from 12/29/02 to 01/18/03. We present the energy spectra of proton and helium extracted from the ATIC flights, over the energy range from 100 GeV to 100 TeV, and compare them with the results from other experiments at both the lower and higher energy ends.

Ahn, H. S.↗
NASA NTRS2003

ATIC Experiment: Preliminary Results from the Flight in 2002

Abstract The Advanced Thin Ionization Calorimeter (ATIC) had successful Long Duration Balloon flights from McMurdo, Antarctica in both 2000 and 2002. The instrument consists of a Silicon matrix for charge measurement, a flared graphite target to induce nuclear interactions, scintillator strip hodoscopes for triggering and helping reconstruct trajectory, and a BGO calorimeter to measure the energy of incident particles. In this paper, we discuss the second flight, which lasted 20 days, starting on 12/29/02. Preliminary results from the on-going analysis of the data including the proton and helium spectra are reported.

Ahn, H. S.↗
NASA NTRS1966

Nuclear Rocket Technology Conference

The Lewis Research Center has a strong interest in nuclear rocket propulsion and provides active support of the graphite reactor program in such nonnuclear areas as cryogenics, two-phase flow, propellant heating, fluid systems, heat transfer, nozzle cooling, nozzle design, pumps, turbines, and startup and control problems. A parallel effort has also been expended to evaluate the engineering feasibility of a nuclear rocket reactor using tungsten-matrix fuel elements and water as the moderator. Both of these efforts have resulted in significant contributions to nuclear rocket technology. Many successful static firings of nuclear rockets have been made with graphite-core reactors. Sufficient information has also been accumulated to permit a reasonable Judgment as to the feasibility of the tungsten water-moderated reactor concept. We therefore consider that this technoIogy conference on the nuclear rocket work that has been sponsored by the Lewis Research Center is timely. The conference has been prepared by NASA personnel, but the information presented includes substantial contributions from both NASA and AEC contractors. The conference excludes from consideration the many possible mission requirements for nuclear rockets. Also excluded is the direct comparison of nuclear rocket types with each other or with other modes of propulsion. The graphite reactor support work presented on the first day of the conference was partly inspired through a close cooperative effort between the Cleveland extension of the Space Nuclear Propulsion Office (headed by Robert W. Schroeder) and the Lewis Research Center. Much of this effort was supervised by Mr. John C. Sanders, chairman for the first day of the conference, and by Mr. Hugh M. Henneberry. The tungsten water-moderated reactor concept was initiated at Lewis by Mr. Frank E. Rom and his coworkers. The supervision of the recent engineering studies has been shared by Mr. Samuel J. Kaufman, chairman for the second day of the conference, and Mr. Roy V. Humble. Dr. John C. Eward served as general chairman for the conference.

Source record↗
NASA NTRS2021

Microstructural Observations on Advanced Multilayer Protective Coatings for Nuclear Thermal Propulsion

The present research discusses the microstructures of a new multilayered coating concept proposed to solve the “midrange corrosion” problem observed in NbC and ZrC-coated graphite (Gr) nuclear fuels during the NERVA/Rover programs. The concept envisions designing a compliant multilayered coatings architecture to accommodate the thermal expansion mismatch strains between the ZrC outer coating and the Gr/(U,Zr)C fuel matrix along with using a diffusion barrier to minimize carbon and hydrogen diffusion, respectively. Multilayered coatings were deposited on Gr substrates by chemical vapor deposition (CVD). Graphite disks, as well as the inner channels of a 19-hole hexagonal rod 152.4 mm long, were coated by CVD in proof-of concept process demonstration studies. Coated disk specimens were thermally-cycled between ambient and 1900 K for as long as 20 cycles with a 1h hold time at the temperature using both stepped and non-interrupted thermal cycling methods. Detailed cross-sectional microstructural analyses were conducted on the thermally-cycled coated disks. These results showed that the coatings were intact after 20 thermal cycles at 1900 K although the bonding of the overlay Mo coating appeared to be influenced by the oxygen content in the layer. Seven transverse sections were cut along the length from the coated hexagonal rod for microstructural studies. All the nineteen channels were observed to have been coated in all the sections. This paper reports the results of detailed microstructural analyses, compositional line scans and x-ray dot maps conducted on three coated channels from a mid-section of the rod. The measured values of carbon (C) were observed to be exceptionally high in all the specimens, which were attributed to hydrocarbon contamination of the specimen surfaces. It was observed that the oxygen (O) levels were high in all the coating layers even after approximately correcting for the observed high values of C. The source of O contamination of the CVD reactor system was attributed to either the presence of oxychloride impurities in the precursor salts or a pin hole leak in the coating system. It is concluded that high values of O result in poor bonding of the Mo overlay to the ZrC layer. These trials have established that it is possible to deposit multilayered coatings by CVD in 19-hole hexagonal rods 152.4 mm long similar to the NERVA fuel rod design although the quality of the coatings and the uniformity of their thicknesses require further improvement.

Nuclear thermal propulsion, multilayer coatings, m↗
NASA NTRSDate not supplied

Development of Multilayered Coatings for Nerva-Type Graphite Fuel Elements- An Overview

“Mid-passage corrosion” (MPC) was observed in NbC and ZrC-coated graphite (Gr)-based nuclear fuels during the NERVA/Rover programs. A new multilayered coating concept was proposed to solve this problem. The concept envisions designing a compliant multilayered coatings architecture to accommodate the thermal expansion mismatch strains between the ZrC outer coating and the Gr/(U,Zr)C fuel matrix. Additionally, the coating architecture uses diffusion barriers to minimize carbon diffusion from the substrate, and hydrogen diffusion from the flowing gas stream, into the multilayer coating. The present paper provides a brief overview of the development of multilayered coatings deposited on Gr disks, as well as within the inner channels of a19-hole hexagonal rod 152.4 mm long, by chemical vapor deposition (CVD) in proof-of concept studies. The cross-sectional microstructural observations are discussed. These trials have established that it is possible to deposit multilayered coatings by CVD in 19-hole hexagonal rods similar to the NERVA fuel rod design although the quality of the coatings and the uniformity of their thicknesses require further improvement.

Multilayer coatings, graphite, nuclear thermal pro↗
NASA NTRS2013

Unit-Sphere Multiaxial Stochastic-Strength Model Applied to Anisotropic and Composite Materials

Models that predict the failure probability of brittle materials under multiaxial loading have been developed by authors such as Batdorf, Evans, and Matsuo. These "unit-sphere" models assume that the strength-controlling flaws are randomly oriented, noninteracting planar microcracks of specified geometry but of variable size. This methodology has been extended to predict the multiaxial strength response of transversely isotropic brittle materials, including polymer matrix composites (PMCs), by considering (1) flaw-orientation anisotropy, whereby a preexisting microcrack has a higher likelihood of being oriented in one direction over another direction, and (2) critical strength, or K (sub Ic) orientation anisotropy, whereby the level of critical strength or fracture toughness for mode I crack propagation, K (sub Ic), changes with regard to the orientation of the microstructure. In this report, results from finite element analysis of a fiber-reinforced-matrix unit cell were used with the unit-sphere model to predict the biaxial strength response of a unidirectional PMC previously reported from the World-Wide Failure Exercise. Results for nuclear-grade graphite materials under biaxial loading are also shown for comparison. This effort was successful in predicting the multiaxial strength response for the chosen problems. Findings regarding stress-state interactions and failure modes also are provided.

Nemeth, Noel, N.↗
NASA NTRS2007

Comparison of Fluka-2006 Monte Carlo Simulation and Flight Data for the ATIC Detector

We have performed a detailed Monte Carlo (MC) simulation for the Advanced Thin Ionization Calorimeter (ATIC) detector using the MC code FLUKA-2006 which is capable of simulating particles up to 10 PeV. The ATIC detector has completed two successful balloon flights from McMurdo, Antarctica lasting a total of more than 35 days. ATIC is designed as a multiple, long duration balloon flight, investigation of the cosmic ray spectra from below 50 GeV to near 100 TeV total energy; using a fully active Bismuth Germanate(BGO) calorimeter. It is equipped with a large mosaic of.silicon detector pixels capable of charge identification, and, for particle tracking, three projective layers of x-y scintillator hodoscopes, located above, in the middle and below a 0.75 nuclear interaction length graphite target. Our simulations are part of an analysis package of both nuclear (A) and energy dependences for different nuclei interacting in the ATIC detector. The MC simulates the response of different components of the detector such as the Si-matrix, the scintillator hodoscopes and the BGO calorimeter to various nuclei. We present comparisons of the FLUKA-2006 MC calculations with GEANT calculations and with the ATIC CERN data and ATIC flight data.

Gunasingha, R.M.↗
NASA NTRS1991

Advanced materials for space nuclear power systems

The overall philosophy of the research was to develop and characterize new high temperature power conversion and radiator materials and to provide spacecraft designers with material selection options and design information. Research on three candidate materials (carbide strengthened niobium alloy PWC-11 for fuel cladding, graphite fiber reinforced copper matrix composites for heat rejection fins, and tungsten fiber reinforced niobium matrix composites for fuel containment and structural supports considered for space power system applications is discussed. Each of these types of materials offers unique advantages for space power applications.

Titran, Robert H.↗
NASA NTRS1991

Advanced materials for space nuclear power systems

The overall philosophy of the research was to develop and characterize new high temperature power conversion and radiator materials and to provide spacecraft designers with material selection options and design information. Research on three candidate materials (carbide strengthened niobium alloy PWC-11 for fuel cladding, graphite fiber reinforced copper matrix composites for heat rejection fins, and tungsten fiber reinforced niobium matrix composites for fuel containment and structural supports) considered for space power system applications is discussed. Each of these types of materials offers unique advantages for space power applications.

Titran, Robert H.↗
NASA NTRS2015

Revised Point of Departure Design Options for Nuclear Thermal Propulsion

In an effort to further refine potential point of departure nuclear thermal rocket engine designs, four proposed engine designs representing two thrust classes and utilizing two different fuel matrix types are designed and analyzed from both a neutronics and thermodynamic cycle perspective. Two of these nuclear rocket engine designs employ a tungsten and uranium dioxide cermet (ceramic-metal) fuel with a prismatic geometry based on the ANL-200 and the GE-710, while the other two designs utilize uranium-zirconium-carbide in a graphite composite fuel and a prismatic fuel element geometry developed during the Rover/NERVA Programs. Two engines are analyzed for each fuel type, a small criticality limited design and a 111 kN (25 klbf) thrust class engine design, which has been the focus of numerous manned mission studies, including NASA's Design Reference Architecture 5.0. slightly higher T/W ratios, but they required substantially more 235U.

Nuclear Thermal Rockets↗
NASA NTRS1980

Characterization of PMR-15 polyimide composition in thermo-oxidatively exposed graphite fiber composites

The contributions of individual resin components to total resin weight loss in 600 F air aged Celion 6000/PMR-15 polyimide composites were determined from the overall resin weight loss in the composite by chemically separating the PMR-15 matrix resin into its monomeric components. The individual resin components were also analyzed by spectroscopic techniques in order to elucidate curing and degradation mechanisms of the PMR-15 matrix resin. The isothermal weight loss of the individual resin components during prolonged 600 F thermo-oxidative aging of the composite was correlated to the changes observed in the Fourier Transform infrared spectra and Fourier Transform nuclear magnetic resonance spectra of the individual resin components. The correlation was used to identify the molecular site of the thermo-oxidative changes in PMR-15 polyimide matrix resin during 600 F curing the prolonged 600 F thermo-oxidative aging.

Alston, W. B.↗