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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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43 records · Page 3
All sourcesNASA NTRSDOE OSTI
NASA NTRS1989

Small blaze angle gratings with various surface treatments for use in the extreme ultraviolet

Three replica gratings made from a single master, including two epoxy-overcoated gratings with different thicknesses of gold and one grating replicated directly in gold, were studied by measuring and comparing the performances of the gratings in EUV light at grazing incidence angles. The results show that, in situations which require a very low grazing angle, high-quality, mechanically ruled straight groove gratings provide good efficiency and uniform blaze independent of coating thickness up to 1000 A. However, the technique of direct replication in the surface material seems to result in substantially more light scattering and does not seem to have other obvious advantages.

Miller, Anne↗
DOE OSTI2020

A White Paper: Disposition Options for a High-Temperature Gas-Cooled Reactor

The high-temperature gas-cooled reactor (HTGR) is a uranium-fueled, graphite-moderated, gas-cooled nuclear reactor design concept capable of producing very high core outlet temperatures. Both types of HTGR have the tristructural isotropic (TRISO) fuel kernel at the heart of the fuel design. For the prismatic block-type HTGR, the TRISO particles are overcoated with a resinated graphitic matrix and pressed into fuel compacts, which are then heat treated and placed in the fuel channels of the prismatic-block-shaped fuel assemblies. For the pebble-bed-type HTGR, the TRISO particles are dispersed in a graphitic-matrix sphere, which is the basic unit for the reactor core. Despite having very different fuel designs, both types of HTGR are graphite-moderated, gas-cooled, thermal reactors using many of the same materials. As a result, both prismatic-block-type and pebble-bed-type HTGRs have similar radioactive waste streams, all of which require safe and secure storage and eventual disposition. Modern HTGR designs are based on a long and rich operating history of several different graphite-moderated, gas-cooled, thermal reactors. Several of these reactors have been shut down, the fuel has been placed in safe storage, and they have undergone some degree of decommissioning. As such, there is significant experience in the management of the spent nuclear fuel (SNF) and radioactive wastes associated with operating these reactors. This white paper will, (1) identify the definitions and regulations that apply to the safe and secure management, storage, and disposal of radioactive waste; and (2) identify the key radioactive waste streams from HTGRs and their characteristics. Idaho National Laboratory (INL) has significant experience in the management of SNF from HTGR predecessors. This experience should form the basis for the management and disposition efforts of the radioactive waste from any new HTGR-type small modular reactor, or microreactor intended for deployment at the INL site.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1988

A test for interfacial effects and stress transfer in ceramic matrix composites

A test specimen was devised for measuring stress transfer between a high modulus fiber and a ceramic matrix. Single filaments of SiC were embedded in chemically vapor deposited SiC on a thin plate of molybdenum. The CVD overcoating which encapsulated the fiber was continuous with a coating of SiC on the molybdenum. When placed in a microtensile test device and loaded in the fiber direction, the fiber fracture characteristics provide information on the fiber/matrix adhesion and stress transfer. Problems were encountered due to the formation of a weak boundary between the SiC and the molybdenum which obviated any meaningful tensile tests. Also, the high CVD temperature used in fabricating these specimens restrict the fiber, matrix (and substrate) to materials having similar thermal coefficients of expansion in order to minimize thermal stresses.

Source record↗
NASA NTRS2000

Polymeric Coatings for Electrodynamic Tethers

Two polymeric coatings have been developed for the Propulsive Small Expendable Deployer System (ProSEDS) mission. ProSEDS is designed to provide an on-orbit demonstration of the electrodynamic propulsion capabilities of tethers in space. The ProSEDS experiment will be a secondary payload on a Delta II unmanned expendable booster scheduled for launch in August 2000. A 5-km conductive tether is attached to the Delta 11 second stage and collects current from the low Earth orbit (LEO) plasma to facilitate de-orbit of the spent stage. The conductive tether is attached to a 10-km non-conductive tether, the other end of which is attached to an endmass containing several scientific instruments. A bare metal tether would have the best conductivity but thermal concerns preclude this design. A conductive polymer developed by Triton Systems has been optimized for conductivity and thermo-optical properties. The current design for the ProSEDS conductive tether is seven strands of 28 AWG aluminum wire individually coated with 8.7 micrometers (0.35 mil) of an atomic oxygen-resistant conductive polymer composed of a mixture of 87% Clear Oxygen-Resistant polymer (COR) and 13% polyanaline (PANi), wrapped around a braided Kevlar (TM) 49 core. Extensive testing has been performed at the Marshall Space Flight Center (MSFC) to qualify this material for flight on ProSEDS. Atomic oxygen exposure was performed, with solar absorptance and infrared emittance measured before and after exposure. Conductivity was measured before and after atomic oxygen exposure. High voltage tests, up to 1500 V, of the current collecting ability of the COR/PANi have been completed. Approximately 160 meters of the conductive tether closest to the Delta 11 second stage is insulated to prevent any electron reconnection to the tether from the plasma contactor. The insulation is composed of polyimide overcoated with TOR-BP, another polymeric coating developed by Triton for this mission. TOR-BP acts as both insulator and a protective coating against atomic oxygen erosion. The insulation has been tested to withstand 5000 V.

Vaughn, Jason A.↗
NASA NTRS1991

Reflective overcoats for radiation control surfaces

Theoretical models are developed to predict the surface properties of a coating layer composed of particles of a known size distribution, applied to an opaque substrate, such as a metal or reaction cured glass (RCG). The surface temperature attained at radiative equilibrium by an overcoated surface subject to a given heat flux is calculated. The incident radiation was assumed to exhibit the spectral distribution characteristic of a black body at different temperatures or equivalently, having different peak wavelengths, with the energy level scaled to give a range of desired surface radiative heat fluxes. This approach allows a straightforward comparison of the thernal response of a surface to incident radiation having the energy predominantly in a characteristic wavelength band and a well-defined spectral distribution. The ratio of the radiative heat flux to the total heat flux was varied, and the different geometric and material parameters of such overcoat layers were explored. The model was applied to representative surface heating rates to the Aeroasssisted Flight Experiment (AFE) and to Aeroassisted Space Transfer Vehicles (ASTVs). The predicted radiative energy flux to the surface of the AFE vehicle gives a single-point comparison of the surface temperatures attained with and without a selective-reflector overcoat on the vehicle surface. The specific objective of this work is to identify the most desirable radiative properties of an overcoat/substrate system for this environment.

White, Susan M.↗
NASA NTRS2000

Reliable, Low-Cost, Low-Weight, Non-Hermetic Coating for MCM Applications

Through an Air Force Research Laboratory sponsored STM program, reliable, low-cost, low-weight, non-hermetic coatings for multi-chip-module(MCK applications were developed. Using the combination of Sandia Laboratory ATC-01 test chips, AvanTeco's moisture sensor chips(MSC's), and silicon slices, we have shown that organic and organic/inorganic overcoatings are reliable and practical non-hermetic moisture and oxidation barriers. The use of the MSC and unpassivated ATC-01 test chips provided rapid test results and comparison of moisture barrier quality of the overcoatings. The organic coatings studied were Parylene and Cyclotene. The inorganic coatings were Al2O3 and SiO2. The choice of coating(s) is dependent on the environment that the device(s) will be exposed to. We have defined four(4) classes of environments: Class I(moderate temperature/moderate humidity). Class H(high temperature/moderate humidity). Class III(moderate temperature/high humidity). Class IV(high temperature/high humidity). By subjecting the components to adhesion, FTIR, temperature-humidity(TH), pressure cooker(PCT), and electrical tests, we have determined that it is possible to reduce failures 50-70% for organic/inorganic coated components compared to organic coated components. All materials and equipment used are readily available commercially or are standard in most semiconductor fabrication lines. It is estimated that production cost for the developed technology would range from $1-10/module, compared to $20-200 for hermetically sealed packages.

Jones, Eric W.↗
DOE OSTI2020

Destructive PIE and Safety Testing of Six AGR-2 UO 2 Capsule 3 Compacts

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2) was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from June 2010 to October 2013 (Collin 2014). The fuel compacts in this experiment held either tristructural isotropic (TRISO)-coated spherical kernels of uranium oxide (UO2) or TRISO-coated kernels containing both uranium carbide and uranium oxide phases (UCO). There were six separately monitored and controlled capsules in the AGR-2 test train. Capsule 3 held twelve compacts containing UO2-TRISO particles fabricated by BWX Technologies Nuclear Operations Group. The AGR-2 TRISO particles were fabricated in a pilot-scale fluidized-bed chemical vapor deposition (FB-CVD) furnace with a coating chamber inner diameter of 150 mm (Phillips, Barnes, and Hunn 2010), which was a change from the first irradiation experiment (AGR-1) particles that had been coated in a lab-scale FB-CVD coating system with a chamber inner diameter of 50 mm (Lowden 2006). The TRISO particles were overcoated with resinated graphite flake at Oak Ridge National Laboratory (ORNL), and the overcoated particles were pressed into one-inch-long, half-inch-diameter cylinders (Hunn, Montgomery, and Pappano 2010). Each cylindrical compact held an average of 1,543 TRISO particles with 9.6% enriched UO2 kernels that had a nominal diameter of 500 μm (Hunn, Savage, and Silva 2012). Capsule 3 compacts were irradiated to average calculated burnups of 9.01–10.69% fissions per initial metal atom (FIMA), and the average calculated fluences of fast neutrons with energies E n > 0.18 MeV were 3.05–3.53×10 25 n/m 2 (Sterbentz 2014). The calculated time-average, volume-average Capsule 3 compact temperatures were 996–1,062°C. However, Capsule 3 compact temperatures varied several hundred degrees across each compact, and the timeaverage minimum (TA min ) and time-average maximum (TA max ) temperatures were between 889–999°C and 1,072–1,105°C, respectively (Hawkes 2014). After irradiation, the AGR-2 test train was transferred from ATR to the INL Materials and Fuels Complex for inspection and disassembly (Ploger, Demkowicz, and Harp 2015). The initial inspection included dimensional metrology of the compacts and graphite fuel holders. Like all the AGR-2 compacts, the compacts in Capsule 3 shrank slightly during irradiation, as expected, with an average length reduction of 1.07–1.24% and an average diameter reduction of 0.13–0.41%. Post-irradiation examination (PIE) of the capsule components was completed to measure select fission products ( 90 Sr, 110m Ag, 134 Cs, 137 Cs, 144 Ce, and 154 Eu) outside the compacts (Stempien and Demkowicz 2020). This involved gamma counting of the graphite and graphoil spacers at the top and bottom of each capsule, acid leaching for radiochemical analysis of fission products on the metallic capsule components, and burn-leach analysis of the graphite holders. The total amount of 110mAg measured on the Capsule 3 components was 13% of the calculated capsule inventory. This was significantly lower than the amount of 110m Ag measured on the three UCO capsule components, which ranged from 32–70%. The lower 110m Ag release in Capsule 3 was likely due to lower peak temperatures compared with the UCO fuel capsules (Hawkes 2014). Measured inventories of the other select fission products were also lower in Capsule 3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗