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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
NASA NTRS1987

Chemical effects in ion mixing of a ternary system (metal-SiO2)

The mixing of Ti, Cr, and Ni thin films with SiO2 by low-temperature (- 196-25 C) irradiation with 290 keV Xe has been investigated. Comparison of the morphology of the intermixed region and the dose dependences of net metal transport into SiO2 reveals that long range motion and phase formation probably occur as separate and sequential processes. Kinetic limitations suppress chemical effects in these systems during the initial transport process. Chemical interactions influence the subsequent phase formation.

Banwell, T.↗
DOE OSTI2022

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 1-3, 4-3, 10-1, and 10-2

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support modeling improvement. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Thermochemically-informed mass transport model for interdiffusion of U and Zr in irradiated U-Pu-Zr fuel with fission products

A new formulation for redistribution of constituents in metallic U-Pu-Zr nuclear fuel is presented that can incorporate the contributions from fission products. The formulation is based on the thermodynamic driving forces derived from the generalized chemical potential that includes effects of composition and temperature. As a result, the redistribution model can readily account for the composition changes due to the generation of fission products while using only a limited set of transport coefficients. The thermodynamic model for the metallic fuel and thermochemistry solver Thermochimica were coupled with the nuclear fuel performance code BISON to implement the redistribution model. The simulations reproduce an experimentally observed Zr-depletion zone in the mid-radius region of an irradiated fuel slug. The generation of fission products during burnup is shown to have a stabilizing effect on the fuel chemistry, slowing the rate of U-Zr interdiffusion and reducing the size of the Zr-depleted zone. Here, the overall objective of the work is to advance predictive capabilities of fission product behaviour in U-Zr metallic fuel in the context of fuel performance and safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Metallic fuel cladding degradation model development and evaluation for BISON

Fuel cladding chemical interaction (FCCI) and coolant cladding chemical interaction (CCCI), both consume stainless-steel cladding in sodium-cooled fast reactors (SFRs), and effectively degrade cladding mechanical performance through thickness reduction. Therefore, FCCI/CCCI and corresponding cladding degradation models are essential for advanced fuel performance codes, such as BISON, to reliably predict cladding behavior in SFRs. Here, we report the development and implementation efforts for BISON's FCCI/CCCI correlations and cladding degradation models based on U.S. legacy metallic fuel data. The models were evaluated using Integral Fast Reactor (IFR) program X447/A experiment data supported by the ongoing integration project enabling coordinated application of BISON and EBR-II fuel irradiation and physics database (FIPD). Furthermore, the implemented models were demonstrated to improve BISON's capabilities of predicting cladding damage and degradation behavior that is more consistent with post-irradiation examination observations. Additionally, some limitations of current BISON modules are identified, which are to be overcome through the BISON-FIPD integration efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Microstructural evolution in doped high entropy alloys NiCoFeCr-3X (X=Pd/Al/Cu) under irradiation

Commonly studied equatomic single-phase FCC high entropy alloys based on 3d transition metals like NiCoFeCr do not provide adequate strength and radiation resistance at high doses for nuclear structural applications. In the current study, the major alloying effects like lattice distortion, ordering and clustering tendencies were investigated by adding low concentration of Pd, Al, or Cu respectively to study the doping effects on the ion irradiation response of NiCoFeCr alloy. The alloys were irradiated with 3 MeV Ni 2+ ions at 500 °C to a fluence of 1 × 10 17 /cm 2 at a beam flux of approximately 2.8 × 10 12 ions/cm 2 /s. The microstructural evolution upon irradiation i.e., formation of dislocation networks, radiation induced segregation and precipitation, and void formation were studied in detail. Further, post-irradiation characterization results showed that a Pd addition leads to a high void nucleation rate but controlled void growth, which may be attributed to increased lattice distortion. In Al added HEA, our microstructural analysis indicates that radiation induced ordered L1 2 precipitates do not affect void swelling significantly. Cu addition led to Cu precipitation that drastically suppressed dislocation density and void swelling of the alloy. Additionally, a model was developed to qualitatively describe the trend in void swelling of typical FCC alloys under ion irradiation. This model was able to qualitatively explain the suppression and reappearance of void swelling in ion irradiated alloys that generally occurs near the region with peak implanted ion concentration.

36 MATERIALS SCIENCE↗
NASA NTRS1993

Vacuum ultraviolet instrumentation for solar irradiance and thermospheric airglow

A NASA sounding rocket experiment was developed to study the solar extreme ultraviolet (EUV) spectral irradiance and its effect on the upper atmosphere. Both the solar flux and the terrestrial molecular nitrogen via the Lyman-Birge-Hopfield bands in the far ultraviolet (FUV) were measured remotely from a sounding rocket on October 27, 1992. The rocket experiment also includes EUV instruments from Boston University (Supriya Chakrabarti), but only the National Center for Atmospheric Research (NCAR)/University of Colorado (CU) four solar instruments and one airglow instrument are discussed here. The primary solar EUV instrument is a 1/4 meter Rowland circle EUV spectrograph which has flown on three rockets since 1988 measuring the solar spectral irradiance from 30 to 110 nm with 0.2 nm resolution. Another solar irradiance instrument is an array of six silicon XUV photodiodes, each having different metallic filters coated directly on the photodiodes. This photodiode system provides a spectral coverage from 0.1 to 80 nm with about 15 nm resolution. The other solar irradiance instrument is a silicon avalanche photodiode coupled with pulse height analyzer electronics. This avalanche photodiode package measures the XUV photon energy providing a solar spectrum from 50 to 12,400 eV (25 to 0.1 nm) with an energy resolution of about 50 eV. The fourth solar instrument is an XUV imager that images the sun at 17.5 nm with a spatial resolution of 20 arc-seconds. The airglow spectrograph measures the terrestrial FUV airglow emissions along the horizon from 125 to 160 nm with 0.2 nm spectral resolution. The photon-counting CODACON detectors are used for three of these instruments and consist of coded arrays of anodes behind microchannel plates. The one-dimensional and two-dimensional CODACON detectors were developed at CU by Dr. George Lawrence. The pre-flight and post-flight photometric calibrations were performed at our calibration laboratory and at the Synchrotron Ultraviolet Radiation Facility (SURF) at the National Institute of Standards and Technology (NIST) in Gaithersburg, Maryland.

Woods, Thomas N.↗
DOE OSTI2020

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compact 1-4 and 10-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a “fuel body”, in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 8-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both to be focused on obtaining fission product transport data to support improvements in modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These goals were combined in the AGR-3/4 irradiation experiment, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body, in which the outer graphite ring included a floor and cap to fully enclose the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to obtain data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Determination of Average Burnup in AGR-3/4 Compacts 1-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite. These two goals were combined in the AGR-3/4 irradiation, which consisted of 12 independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. Two capsule types were used: a standard capsule and a so-called fuel body, in which the outer graphite ring included a floor and cap that fully enclosed the fuel. The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from the matrix and graphite at accident temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2005

Surface characteristics of spacecraft components affect the aggregation of microorganisms and may lead to different survival rates of bacteria on Mars landers

Layers of dormant endospores of Bacillus subtilis HA101 were applied to eight different spacecraft materials and exposed to martian conditions of low pressure (8.5 mbar), low temperature (-10 degrees C), and high CO(2) gas composition and irradiated with a Mars-normal ultraviolet (UV-visible- near-infrared spectrum. Bacterial layers were exposed to either 1 min or 1 h of Mars-normal UV irradiation, which simulated clear-sky conditions on equatorial Mars (0.1 tau). When exposed to 1 min of Mars UV irradiation, the numbers of viable endospores of B. subtilis were reduced three to four orders of magnitude for two brands of aluminum (Al), stainless steel, chemfilm-treated Al, clear-anodized Al, and black-anodized Al coupons. In contrast, bacterial survival was reduced only one to two orders of magnitude for endospores on the non-metal materials astroquartz and graphite composite when bacterial endospores were exposed to 1 min of Mars UV irradiation. When bacterial monolayers were exposed to 1 h of Mars UV irradiation, no viable bacteria were recovered from the six metal coupons listed above. In contrast, bacterial survival was reduced only two to three orders of magnitude for spore layers on astroquartz and graphite composite exposed to 1 h of Mars UV irradiation. Scanning electron microscopy images of the bacterial monolayers on all eight spacecraft materials revealed that endospores of B. subtilis formed large aggregates of multilayered spores on astroquartz and graphite composite, but not on the other six spacecraft materials. It is likely that the formation of multilayered aggregates of endospores on astroquartz and graphite composite is responsible for the enhanced survival of bacterial cells on these materials.

Mars↗
DOE OSTI2021

Initial demonstration of automated fuel performance modeling with 1977 EBR-II metallic fuel pins using BISON code with FIPD and IMIS databases

Using the BISON fuel performance code, simulations were conducted using an automated process to read initial and operating conditions from the Fuels Irradiation and Physics Database (FIPD) and Integral Fast Reactor materials information system (IMIS) database, which contains metallic fuel data from the Experimental Breeder Reactor-II (EBR-II). This work demonstrates use of an integrated framework to access the vast majority of EBR-II experimental fuel pin data to support rapid development of fuel performance models for next-generation metallic fuel systems. With this capability, validation for fuel qualification can be performed rapidly. Between IMIS and FIPD, there is enough information to conduct 1977 unique EBR-II metallic fuel pin histories from 24 different experiments, at varying levels of detail between the two databases. Each of these histories includes a high-resolution power history, flux history, coolant channel flow rates, and coolant channel temperatures. Fission gas release (FGR), cumulative damage fraction (CDF), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON. The results were compared to post-irradiation examination (PIE) results for the initial demonstration of automated BISON modeling. BISON simulations conducted with IMIS and FIPD were in rough agreement with PIE measurements and calculations. Cladding profilometry, FGR, and fuel axial swelling were found to be in rough agreement with PIE measurements, depending on the physics used within the BISON input files. Here, the mechanical contact solver chosen was found to significantly impact axial fuel swelling and cladding strain predictions. CDF values were assessed to see whether pin failure may have been predicted (CDF ≥ 1). This work suggests that continued development of an automated tool for BISON should focus on inclusion of the Fast Flux Test Facility (FFTF) experimental data for a larger database for metallic fuel, improved physical models to better capture fuel performance, such as fuel-cladding interactions, and a more detailed comparison with available PIE data to further the BISON model development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Radiolytic Production of Radical Species in Gibbsite Doped with Iron and Chromium Ions

Gibbsite (aluminum hydroxide, Al(OH) 3 ) nanoparticles, synthesized from aluminum chloride or aluminum nitrate, were doped with metal ions, Cr(III) or Fe(III), and then irradiated with γ rays to determine the effect of the dopants on radiolytic hydrogen (H 2 ) production and radical generation. The addition of Cr(III) and Fe(III) ions at concentrations of 0.5% or 5% decreased the concentration of stable oxygen-centered radicals, with the strongest suppression in Cr(III) doped samples. A decrease in H 2 yields was observed with increasing Cr(III) or Fe(III) concentrations, with the greatest effect observed for the Fe(III)-doped samples. Reduction of the Cr(III) to Cr(II) and Fe(III) to Fe(II) was also observed, probably due to scavenging of radiolytically produced electrons. However, further processes differ for Fe(III)- and Cr(III)-doped systems. Both ions are reduced by the free electrons, leading to a decrease in H 2 production, but they react differently with the oxygen radicals. Cr(III) can be oxidized by oxygen radicals, whereas Fe(III) cannot. Fe(II) can interact with peroxides, possible products of intermediate oxygen oxidation, converting back to Fe(III) and leaving oxygen radicals behind. These oxidation reactions lead to a difference in the observed relative effects on H 2 yields and oxygen radical production between Cr(III)- and Fe(III)-doped gibbsite. In conclusion, the connection between electron scavenging and H 2 production indicates that radiolytically produced electrons are precursors to H 2 .

Anions↗
NASA NTRS1988

Mapping Redistribution Of Metal In Welds

Radioactive-tracer technique applied to map redistribution of metal caused by welding process. Surfaces of parts welded irradiated by particle-beam generators to make them slightly radioactive. Used to verify predictions of computer codes for dynamics of fluids in weld pools.

Barkhoudarian, Sarkis↗
NASA NTRS1987

BE Ursae Majoris: A detached binary with a unique reprocessing spectrum

New infrared photometry, optical and UV spectrophotometry, and a photographic ephemeris are presented for the detached binary BE UMa. Results show the primary to be a DO white dwarf with an effective temperature of 80,000 + or - 15,000 K and a mass of 0.6 + or - 0.1 solar masses. No evidence is found for variability of the primary. The main sequence secondary star is shown to be of early M spectral type, with a formal range of M1 to M5 being possible. A reflection effect in reprocessed line and continuum radiation is produced by EUV radiation from the primary incident on the secondary atmosphere. It is suggested that the temperature of the reprocessed component of the secondary's atmosphere is in the 5000 to 8500 K range, and that emission lines of decreasing ionization form deeper in the irradiated envelope. Relatively narrow He II and high excitation metal lines are formed from recombination and continuum fluorescence processes.

Ferguson, Donald H.↗
NASA NTRS1995

Fission foil detector calibrations with high energy protons

Fission foil detectors (FFD's) are passive devices composed of heavy metal foils in contact with muscovite mica films. The heavy metal nuclei have significant cross sections for fission when irradiated with neutrons and protons. Each isotope is characterized by threshold energies for the fission reactions and particular energy-dependent cross sections. In the FFD's, fission fragments produced by the reactions are emitted from the foils and create latent particle tracks in the adjacent mica films. When the films are processed surface tracks are formed which can be optically counted. The track densities are indications of the fluences and spectra of neutrons and/or protons. In the past, detection efficiencies have been calculated using the low energy neutron calibrated dosimeters and published fission cross sections for neutrons and protons. The problem is that the addition of a large kinetic energy to the (n,nucleus) or (p,nucleus) reaction could increase the energies and ranges of emitted fission fragments and increase the detector sensitivity as compared with lower energy neutron calibrations. High energy calibrations are the only method of resolving the uncertainties in detector efficiencies. At high energies, either proton or neutron calibrations are sufficient since the cross section data show that the proton and neutron fission cross sections are approximately equal. High energy proton beams have been utilized (1.8 and 4.9 GeV, 80 and 140 MeV) for measuring the tracks of fission fragments emitted backward and forward.

Benton, E. V.↗
NASA NTRS2020

NASA GRC Hosts Lattice Confinement Fusion Virtual Workshop

For several years (since 2014) the Advanced Energy Conversion (AEC) Project has examined novel nuclear reactions in materials that absorb large quantities of deuterium fuel tightly held in a lattice. These experiments culminated in a bremsstrahlung irradiation campaign that repeatedly induced nuclear reactions in deuterated metals.

Theresa L Benyo↗
NASA NTRS1974

Influence of ultrasounds on electrolytic polishing

On the basis of study of the electrolytic polishing of metal pieces, the following conclusions can be drawn: (1) Utilization of ultrasonically irradiated electrolytes in the electrolytic polishing of copper and steel pieces leads to enhanced polishing quality. (2) Polishing time for the reference pieces was twice as long as for the test pieces. (3) The chemical attack exerted on the surface of the test pieces in order to get the structure of the respective metal lasted less than half as long and yielded superior qualitative results in the copper as well as in the OT 60-2 steel test pieces polished with ultrasonically irradiated electrolytes. (4) The utilization of higher polishing current densities can be explained by the fact that the cavitational effect that arises in the electrolyte owing to ultrasounds contributes to an increase in the solution's electrical conductivity and elimination of gases from the electrolyte.

Gavrila, I.↗
DOE OSTI2022

A Review of Irradiation Damage and Effects in α-Uranium

Understanding irradiation damage and effects in α-uranium (α-U) is critical to modeling the behavior of U-based metallic fuels. The aim of this review is to address the renewed interest in U-based metallic fuels by examining the state-of-the-art knowledge associated with the effect of irradiation on the microstructure, dimensional changes, and properties of α-U. We critically review the research progress on irradiation-induced growth and swelling, the enhancement of plastic flow and superplasticity by irradiation, and the effect of irradiation on thermal and electrical properties of α-U. Finally, we outline the research directions that require advancements, specifically the need to carry out fundamental research on several of the less understood mechanisms of irradiation damage and effects in α-U.

36 MATERIALS SCIENCE↗