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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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Collagen micromorphology is organized into fibers or films based on collagen concentration and aerogel surface topography.
The flowability of biomass is an indicator of how amenable biomass is to handling. The flowability measurements are often determined with empirical or tertiary experiments. This status quo impedes employing engineering principles to advance the design and operation of biomass handling systems. It is imperative to establish an experimental protocol minimizing empirical aspects of flowability characterization to account for the variability of biomass. This study demonstrates the operational principles of the large chamber cubical triaxial tester and the procedure of triaxial tests to quantify bulk flow behaviors of two milled biomass feedstocks, namely corn stover 2 mm (CS) and Douglas fir 1 mm (DF). The Mohr-Coulomb (MC), Drucker-Prager (DP) and modified Cam Clay (mCC) models are calibrated for both biomasses. Analysis indicates that CS will exhibit a cohesive flow with a larger cohesion coefficient of MC (3.8 ± 3.5 kPa) than DF (0.42 ± 0.9 kPa) and a larger d value of DP (6.9 kPa) than DF (0.0 kPa), respectively. CS exhibits a higher spring-back index of mCC (0.39 ± 0.05) than DF (0.27 ± 0.05) also suggesting handling issues, which agrees with the experiences in the industry. This study demonstrates the capability of a CTT in a quantitative investigation of biomass handling characteristics.
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Here, the aim of this study was to retire the risk of maintaining the integrity of S2-glass reinforced CTD-403 (a cyanate ester resin) that is exposed to radiation and elevated temperature over the life of the measurement of a lepton–lepton electroweak reaction (MOLLER) experiment in experimental Hall A at Jefferson Lab, Newport News, VA. In this article, the shear strength and flexural modulus of irradiated S2-glass reinforced CTD-403 specimens were measured at 65 °C (the magnets are to operate at less than 65 °C) under two scenarios: vacuum and gaseous nitrogen. The testing method is the short-beam shear (SBS) test according to ASTM D2344. The specimens were exposed to neutrons and gamma rays up to 124 MGy. The results show that specimens have excellent resistance against radiation, only 23% degradation of apparent shear strength with 124 MGy at 65 °C under vacuum. At the highest dose areas of the coils, tungsten plates are used to reduce the radiation dose to the resin system. The conclusion is that S2-glass reinforced CTD-403 is well-suited for electrical insulation of MOLLER magnets.
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The effect of yttria concentration (0-33.4 mol%), extraction rates (0.17, 0.33, 0.50, and 0.67 mm s -1 ), and the number of layers (up to four) on the phase content, surface defects, thickness, hardness, adhesion strength, and wear rate of yttria-stabilized zirconia coatings produced by sol-gel/dip-coating were studied for its use on thermolabile substrates. At 700°C, a metastable tetragonal phase ($\mathcal{t''}$) was obtained even with 33.4 mol% yttria when heat treated for 24 hours; however, a fully cubic structure was attained by extending the heat treatment up to 48 hours as confirmed by Raman spectroscopy. Furthermore, it was necessary to use withdrawal speeds of at least 0.67 mm s -1 to produce defect-free coatings. Although the coatings were produced at low temperature, they showed 41% lower wear rate than steel and an adhesion strength of 30 MPa. Our work stresses the importance of the heat treatment history on the stabilization of the cubic phase in sol-gel YSZ coatings.
Superconducting spoke cavities are prone to multipactor - resonant raise of a number of electrons due to secondary emission. Recently proposed and tested by TRIUMF balloon-type spoke cavity showed an outstanding multipactor (MP) suppression property but unfortunately serious Q degradation at high fields. A new fully developed design of a dressed balloon cavity which can be used for any proton linac SSR2 section is developed. The design incorporates additional EP ports for high Q-factor demonstration. Superior properties are demonstrated, such as effective multipactor suppression, 40% lower Lorentz force coefficient, zero sensitivity to external pressure. This paper presents the results of coupled structural Multiphysics analysis, and engineering design of the dressed balloon cavity with EP ports.
Ferritic-martensitic steels G92-2b (an optimized Grade 92 heat), NF616 and T91, and austenitic stainless steel 800H and its Grain Boundary Engineering (GBE)-treated version 800H-TMP (ThermoMechanical Processing) were irradiated in the High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) and the Advanced Test Reactor (ATR) of Idaho National Laboratory (INL). Selected G92-2b samples were irradiated up to 14.66 dpa in the HFIR at two temperature ranges: 400–496.7°C and 683.3– 720°C. NF616 and T91 were irradiated in the ATR up to 8.16 dpa with the irradiation temperatures ranged from 241°C to 447.5°C. Alloy 800H and 800H-TMP samples were irradiated in both the HFIR and the ATR. Selected 800H and 800H-TMP samples had HFIR irradiation to 1.28 dpa at 580°C and ATR irradiation up to 9.12 dpa at 359°C to 431°C. Vickers hardness measurements, fractography, and microstructural characterization were performed on the selected samples in the Low Activation Materials Design and Analysis (LAMDA) laboratory. Radiation-hardening of G92-2b was observed at the lower doses and lower irradiation temperatures (400- 496.7°C), with GB03 (0.52 dpa at 400°C) and GB04 (7.44 dpa at ~490°C) showing ~12% and ~8% hardening, respectively. Softening by ~14% was observed for GB05 (14.66 dpa at 496.7°C). Radiationsoftening of G92-2b was more prevalent at the higher irradiation temperatures (683.3-~720°C), with GB10 (0.46 dpa at 683.3°C), GB11 (7.44 dpa at ~720°C), and GB12 (14.63 dpa at ~720°C) showing ~8%, ~8%, and ~40% softening, respectively. Radiation-hardening of NF616 and T91 was observed with the hardness increased by ~37% to ~65% depending on the irradiation doses and irradiation temperatures. Within the studied irradiation conditions of NF616 and T91, samples with a higher dose had a larger hardness after irradiation. All the tested alloy 800H and 800H-TMP samples in this work showed radiation-hardening by ~96±7% to ~152±10%. Alloy 800H-TMP tended to have slightly smaller radiation-hardening than alloy 800H. The fractography results of G92-2b sample GB03, GB10, and GB11, together with the previously characterized fractography of GB04, GB05, and GB12, indicated that the ductility of G92-2b was maintained up to 14.66 dpa at the lower irradiation temperatures of 400-496.7°C, while some loss of ductility (less necking) was observed for higher doses at the higher irradiation temperatures of 683.3-720°C. This agrees with the previously reported tensile test results of G92-2b, where the elongation of G92-2b was reduced at higher doses at the higher irradiation temperatures. Dimple sizes increased at higher doses, which are more evident at the higher irradiation temperatures of 683.3-~720°C. The fractography of NF616 sample D2 (2.96 dpa at 291.5°C), D4 (5.91 dpa at 359°C), and D6 (8.16 dpa at 431°C) indicated loss of ductility with negligible necking for sample D2, while ductile failure for samples D4 and D6. Fractography of alloy 800H and 800H-TMP samples in various irradiation conditions showed ductile failure with obvious necking. Dimples were observed, with some of them containing large Ti-rich particles, in all the 800H/800H-TMP samples. Electron backscatter diffraction characterization of GB12 indicated the recovery of the lath structure, which was generally replaced by an equiaxed grain structure. Transmission electron microscopy (TEM) characterization showed the presence of frequent M 23 C 6 (M = primarily Cr), MX (M = primarily V), spherical Nb(C,N) precipitates, and occasional Laves phase precipitates in the G92-2b samples. MX precipitates with sizes of 20-30 nm were observed at boundaries of smaller grains, indicating the pinning effect of the V-rich precipitates. The lath structure recovery was more evident at the higher irradiation temperatures (683.3-~720°C), with decreased densities of line dislocations and M23C6 precipitates. The irradiated T91 (TA04) showed the growth of M23C6 precipitates to 101 ± 40 nm from the initial 68 ± 22 nm in the unirradiated condition. Dislocation loops of both {100} and {111} types were present in TA04. TEM characterization was also performed on the irradiated 800H (N4, N5, N6, and AR2) and 800H-TMP (P4, P5, P6, and HG1). Accumulation of M 23 C 6 precipitates at grain boundaries was observed in all the xii 800H/800H-TMP samples, and the presence of Ti(C,N) precipitates at grain boundaries and in the matrix was observed in the irradiated 800H-TMP. Some Ti(C,N) precipitates are embedded in the M 23 C 6 precipitates, maintaining specific orientation relationships between the precipitates and between the precipitate and matrix. In addition, nanoscale Si-rich clusters were observed in the matrix of all the 800H/800H-TMP samples, with EDS Si maps tending to have a lower contrast in 800H-TMP samples. Atom probe tomography was conducted on the same samples, supported by a Rapid Turnaround Examination project under Nuclear Science User Facilities. The results are being analyzed to be integrated with the TEM results for a confident description of the γ’ precipitates. Dislocation loops also formed in all the 800H/800H-TMP samples. The density and the average size of dislocation loops were quantified to be in the order of 10 22 – 10 23 m -3 and 11.7 – 15.9 nm, respectively, in the ATR-irradiated 800H/800H-TMP samples. The loop densities in irradiated 800H were higher than that in irradiated 800H-TMP under the same irradiation conditions. Further systematic data analyses, together with some complementary experiments, will be pursued for these samples to foster peer-reviewed journal article publications.
Accident Tolerant Fuel concepts have been developed and tested in diverse research programs presently around the world. Industry teams have developed proprietary coating compositions and geometries, but significant knowledge gaps exist in terms of the resulting properties of the coating and the cladding and how those components impact each other in reactor conditions. Furthermore, the national laboratory and university teams currently studying the industry concepts are limited in their ability to freely disseminate and discuss results obtained on commercial coatings. The Advanced Fuels Campaign is therefore working to develop an understanding of ‘generic’ coatings applied to commercial nuclear cladding alloys free of these constraints. This milestone report provides insights in the microstructure of a 7-micrometer Cr coating on Zircaloy-2 and a 5- micrometer coating on Zry-4 as reference materials. SEM and TEM methods were utilized to describe and identify the interface, while nanoindentation was used to determine hardness and E-Moduli of coating and cladding material. This milestone summarizes the acquired knowledge to build a necessary verified foundation for future simulation and qualification efforts.
FeCrAl alloys are being developed for accident tolerant fuels for current light water reactors. Specifically as fuel claddings with enhanced safety. These alloys are being developed to have excellent corrosion resistance in light water reactor coolant environments and structural integrity for longer durations in the case of a loss of coolant scenario where time is critical for safety. Tensile, shear punch, and Vickers hardness testing was performed on neutron irradiated FeCrAl alloys with three variations in composition. Testing was done at room temperature for comparison of the three techniques. FeCrAl alloys in this report were irradiated at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR) to 7 dpa and at 320 °C.
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