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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Lattice damage and compositional changes in Xe ion irradiated In{sub x}Ga{sub 1-x}N (x = 0.32−1.0) single crystals

Lattice disorder and compositional changes in In{sub x}Ga{sub 1-x}N (x = 0.32, 0.47, 0.7, 0.8, and 1.0) films on GaN/Al{sub 2}O{sub 3} substrates, induced by room-temperature irradiation of 5 MeV Xe ions, have been investigated using both Rutherford backscattering spectrometry under ion-channeling conditions and time-of-flight secondary ion mass spectrometry. The results show that for a fluence of 3 × 10{sup 13 }cm{sup −2}, the relative level of lattice disorder in In{sub x}Ga{sub 1-x}N increases monotonically from 59% to 90% with increasing indium concentration x from 0.32 to 0.7; a further increase in x up to 1.0 leads to little increase in the disorder level. In contrast to Ga-rich In{sub x}Ga{sub 1-x}N (x = 0.32 and 0.47), significant volume swelling of up to ∼25% accompanied with oxidation in In-rich In{sub x}Ga{sub 1-x}N (x = 0.7, 0.8, and 1.0) is observed. In addition, irradiation-induced atomic mixing occurs at the interface of In-rich In{sub x}Ga{sub 1-x}N and GaN. The results from this study indicate an extreme susceptibility of the high In-content In{sub x}Ga{sub 1-x}N to heavy-ion irradiation, and suggest that cautions must be exercised in applying ion-implantation techniques to these materials at room temperature. Further studies of the irradiation behavior at elevated temperatures are warranted.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2016

Irradiation effects in tungsten-copper laminate composite

Tungsten-copper laminate composite has shown promise as a structural plasma-facing component as compared to tungsten rod or plate. The present study evaluated the tungsten-copper composite after irradiation in the High Flux Isotope Reactor (HFIR) at temperatures of 410-780°C and fast neutron fluences of 0.02-9.0×1025 n/m2, E>0.1 MeV, 0.0039-1.76 displacements per atom (dpa) in tungsten. Tensile tests were performed on the composites, and the fracture surfaces were analyzed with scanning electron microscopy. Before irradiation, the tungsten layers had brittle cleavage failure, but the overall composite had 15.5% elongation at 22°C. After only 0.0039 dpa this was reduced to 7.7% elongation, and no ductility was observed after 0.2 dpa at all irradiation temperatures when tensile tested at 22°C. For elevated temperature tensile tests after irradiation, the composite only had ductile failure at temperatures where the tungsten was delaminating or ductile.

Garrison, L. M.↗
DOE OSTI2015

The dynamics of Al/Pt reactive multilayer ignition via pulsed-laser irradiation

Reactive multilayers consisting of alternating layers of Al and Pt were irradiated by single laser pulses ranging from 100 μs to 100 ms in duration, resulting in the initiation of rapid, self-propagating reactions. The threshold intensities for ignition vary with the focused laser beam diameter, bilayer thickness, and pulse length and are affected by solid state reactions and conduction of heat away from the irradiated regions. High-speed photography was used to observe ignition dynamics during irradiation and elucidate the effects of heat transfer into a multilayer foil. For an increasing laser pulse length, the ignition process transitioned from a more uniform to a less uniform temperature profile within the laser-heated zone. A more uniform temperature profile is attributed to rapid heating rates and heat localization for shorter laser pulses, and a less uniform temperature profile is due to slower heating of reactants and conduction during irradiation by longer laser pulses. Finite element simulations of laser heating using measured threshold intensities indicate that micron-scale ignition of Al/Pt occurs at low temperatures, below the melting point of both reactants.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2012

Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2012

Irradiation effects on microstructure change in nanocrystalline ceria – Phase, lattice stress, grain size and boundaries

With a wide variety of applications in numerous industries, ranging from biomedical to nuclear, ceramics such as ceria are key engineering materials. It is possible to significantly alter the materials functionality and therefore its applications by reducing the grain size to the nanometer size regime, at which point the unique varieties of grain boundaries and associated interfaces begin to dominate the material properties. Nanocrystalline films of cubic ceria deposited onto Si substrates have been irradiated with 3 MeV Au+ ions at temperatures of 300 and 400 K to evaluate their response to irradiation. It was observed that the films remained phase stable. Following a slight stress relief stage at low damage levels, the overall lattice is extremely stable up to high irradiation dose of ~34 displacements per atom. The grains were also observed to undergo a temperature-dependent grain growth process upon ion irradiation. This is attributed to a defect-driven mechanism in which the diffusion of defects from the collision cascade is critical. Formation of dislocations that terminate and stabilize at symmetric grain boundaries may be the limiting factor in the grain growth and overall energy reduction of the system. Utilizing ion modification, possible improvement of the adhesion of thin films and reduction of the probability of detrimental effects of stress-induced problems are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Bubble Formation and Lattice Parameter Changes Resulting from He Irradiation of Defect-Fluorite Gd2Zr2O7

Pyrochlores have long been considered as potential candidates for advanced ceramic waste-forms for the immobilization of radioactive waste nuclides. This work provides evidence that Gd2Zr2O7, often considered the most radiation tolerant pyrochlore, could be susceptible to radiation damage in the form of bubble nucleation at the highest He doses expected over geological time. Ion irradiations were utilized to experimentally simulate the radiation damage and He accumulation produced by ..alpha..-decay. Samples were pre-damaged using 7 MeV Au3+ to induce the pyrochlore to defect-fluorite phase transformation, which would occur due to ..alpha..-recoil damage within several hundred years of storage in a Gd2Zr2O7 waste-form. These samples were then implanted to various He concentrations in order to study the long-term effects of He accumulation. Helium bubbles 1-3 nm in diameter were observed in TEM at a concentration of 4.6 at.% He. Some bubbles remained isolated, while others formed chains 10-30 nm in length parallel to the surface. GIXRD measurements showed lattice swelling after irradiating pristine Gd2Zr2O7 with 7 MeV Au3+ to a fluence of 2.2 x 1015 Au/cm2. An increase in lattice swelling was also measured after 2.2 x 1015 Au/cm2 + 2 x 1015 He/cm2 and 2.2 x 1015 Au/cm2 + 2 x 1016 He/cm2. A decrease in lattice swelling was measured after irradiation with 2.2 x 1015 Au/cm2 + 2 x 1017 He/cm2, the fluence where bubbles and bubble chains were observed in TEM. Bubble chains are thought to form in order to reduce lattice strain normal to the surface, which is produced by the Au and He irradiation damage.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Response of nanostructured ferritic alloys to high-dose heavy ion irradiation

A latest-generation aberration-corrected scanning/transmission electron microscope (STEM) is used to study heavy-ion-irradiated nanostructured ferritic alloys (NFAs). Results are presented for STEM X-ray mapping of NFA 14YWT irradiated with 10 MeV Pt to 16 or 160 dpa at -100°C and 750°C, as well as pre-irradiation reference material. Irradiation at -100°C results in ballistic destruction of the beneficial microstructural features present in the pre-irradiated reference material, such as Ti-Y-O nanoclusters (NCs) and grain boundary (GB) segregation. Irradiation at 750°C retains these beneficial features, but indicates some coarsening of the NCs, diffusion of Al to the NCs, and a reduction of the Cr-W GB segregation (or solute excess) content. Ion irradiation combined with the latest-generation STEM hardware allows for rapid screening of fusion candidate materials and improved understanding of irradiation-induced microstructural changes in NFAs.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2017

Initial results from safety testing of US AGR-2 irradiation test fuel

Two cylindrical compacts containing tristructural isotropic (TRISO)-coated particles with kernels that contained a mixture of uranium carbide and uranium oxide (UCO) and two compacts with UO 2 -kernel TRISO particles have undergone 1600°C safety testing. These compacts were irradiated in the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test (AGR-2). The time-dependent releases of several radioisotopes ( 110m Ag, 134 Cs, 137 Cs, 154 Eu, 155 Eu, 90 Sr, and 85 Kr) were monitored while heating the fuel specimens to 1600°C in flowing helium for 300 h. The UCO compacts behaved similarly to previously reported 1600°C-safety-tested UCO compacts from the AGR-1 irradiation. No failed TRISO or failed SiC were detected (based on krypton and cesium release), and cesium release through intact SiC was very low. Release behavior of silver, europium, and strontium appeared to be dominated by inventory originally released through intact coating layers during irradiation but retained in the compact matrix until it was released during safety testing. Both UO 2 compacts exhibited cesium release from multiple particles whose SiC failed during the safety test. Europium and strontium release from these two UO 2 compacts appeared to be dominated by release from the particles with failed SiC. Silver release was characteristically like the release from the UCO compacts in that an initial release of the majority of silver trapped in the matrix occurred during ramping to 1600°C. However, additional silver release was observed later in the safety testing due to the UO 2 TRISO with failed SiC. Failure of the SiC layer in the UO 2 fuel appears to have been dominated by CO corrosion, as opposed to the palladium degradation observed in AGR-1 UCO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evaluation of scenario-specific modeling approaches to predict plane of array solar irradiation

The two main objectives of this study are to evaluate the extent to which the prediction of plane of array irradiance is improved when diffuse radiation is predicted using location-specific regression models developed from onsite measured data as against using generalized models and to estimate the expected uncertainties associated with plane of array irradiance predictions under different data collection scenarios likely to be encountered in practical situations.

Moslehi, Salim↗
DOE OSTI2020

ORNL Post-irradiation Examination (LAMDA) Final Report on Characterization of Irradiated Grade 92 Ferritic-Martensitic Steels

Irradiation effects on ferritic-martensitic steel Grade 92 were investigated in this project. Samples from two heats of Grade 92, i.e., G92-2b and NF616, were selected, which were irradiated in the High Flux Isotope Reactor (HFIR) and the Advanced Test Reactor (ATR), respectively. Samples from two heats of ferriticmartensitic steel T91, irradiated in the ATR, were also selected as references for the Grade 92 samples. The selected G92-2b samples were irradiated to 0.5–14.7 displacements per atom (dpa) at two temperature regimes: 400–497 °C and 683–720°C. The selection of the high irradiation temperatures up to ~720°C is to test the performance at such upper bound temperatures, which would provide insights for the upper bound applicable temperature for Grade 92. The selected NF616 and T91 samples were irradiated to 3–8.2 dpa at 292–448°C. Except for T91 heat2 samples in 3-mm diameter discs, all the other selected samples are of type SS-J2 miniature tensile specimens. Two Nuclear Science User Facilities (NSUF), i.e., the Irradiated Materials Examination and Testing (IMET) hot cell facility and the Low Activation Materials Design and Analysis (LAMDA) facility, were used in the Post-Irradiation Examination (PIE) activities for the selected samples, which include: Tensile tests at room temperature (IMET); Fractography of the tensile fracture surfaces (LAMDA: scanning electron microscopy and energy; dispersive spectroscopy); Vickers hardness measurements (LAMDA); and Microstructural characterization (LAMDA: electron backscatter diffraction, focused-ion beam, and transmission electron microscopy).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

STEM imaging of irradiation induced defects in Eurofer97 steel variants irradiated in the EUROfusion collaboration

Ten exploratory variants of Eurofer97 reduced activation ferritic martensitic (RAFM) steel were irradiated in HFIR to ~2.94 – 32.4 dpa, 300±30 °C as a part of the EUROfusion collaboration. The irradiations were performed in rabbit capsules. ES21-22 capsule were used for irradiating SS-J3 tensile samples while M4CVN bend bar samples were irradiated in ES31-35 capsules. The mechanical properties and multi-length scale characterization of the microstructures of these ten steels, designated as H, I, P, J, K, L, M, N, O, P and reference E, in the nonirradiated form was reported in "Mechanical properties and microstructure characterization of unirradiated Eurofer-97 steel variants for the EUROfusion project, ORNL/SPR-2018/882," while their irradiated properties and some microstructure analysis was reported in "Post-irradiation examination of Eurofer97 steel variants irradiated to 2.5 dpa, ~300 °C in HFIR for the EUROfusion program, ORNL/SPR-2020/1440." Here, additional microstructures using TEM/STEM and STEM-EDX of the irradiated steels is presented. Characterization was performed at the Low Activation Materials Development and Analysis (LAMDA) laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Irradiation And Post-Irradiation Examination Plan For A709

The Advanced Materials and Manufacturing Technologies program is proposing a neutron irradiation campaign for alloy 709 (A709) plate product developed by the Advanced Reactor Technologies program. This irradiation campaign will support the deployment of A709 (included welded components) by providing data of engineering importance, such as data that can be used for establishing end-of-life design parameters and knockdown factors for irradiated materials properties versus properties for non-irradiated material. It will also advance the scientific understanding of A709 irradiation behaviors and material property evolution, which is vital for deploying A709 components into reactors with different neutron spectra, neutron fluxes, and irradiation temperatures as well as accelerating the deployment of new product forms and weld procedures. This document presents the plan for the neutron irradiation and post-irradiation examination for A709 and the scientific and technical basis for the plan.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1990

An Estimate of Changes in the Sun's Total Irradiance Caused by UV Irradiance Variations from 1874 to 1988

Enhanced emission from bright solar faculae is a source of significant variation in the sun's total irradiance. Relative to the emission from the quiet sun, facular emission is known to be considerably greater at UV wavelengths than at visible wavelengths. Determining the spectral dependence of facular emission is of interest for the physical insight this may provide to the origin of the sun's irradiance variations. It is also of interest because solar radiation at lambda less than 300 nm is almost totally absorbed in the Earth's atmosphere. Depending on the magnitude of the UV irradiance variations, changes in the sun's irradiance that penetrates to the Earth's surface may not be equivalent to total irradiance variations measured above the Earth's atmosphere. Using an empirical model of total irradiance variations which accounts separately for changes caused by bright faculae from those associated with dark sunspots, the contribution of UV irradiance variations to changes in the sun's total irradiance is estimated during solar cycles 12 to 21.

Lean, J.↗
DOE OSTI2020

A quantitative study of retention and release of deuterium and tritium during irradiation of y-LiAlO2 pellets

This study reports on the quantification of deuterium in ion-irradiated gamma-LiAlO2 pellets as a function of dose and temperature. The gamma-LiAlO2 pellets were sequentially irradiated with He+ and D2+ ions to the same fluences of 5E16, 1E17 and 2E17 He++D+/cm2 at 188 K. Additional irradiation was performed to 1E17, 2E17 and 4E17 He++D+/cm2 at 573 K. A set of the pellets irradiated at 188 K was shipped and stored at low temperatures from 80 to 132 K and characterized using time-of-flight secondary ion mass spectrometry at ~173 K. The deuterium depth profiles show a Gaussian-like distribution in the low-temperature pellets. The total deuterium retention is found to be directly proportional to the ion fluence. About 27 at.% of the implanted deuterium atoms were released from the pellet irradiated to 2E17 He++D+/cm2 at 188 K during storage at room temperature for ~1 month. Retention of the trapped or bound deuterium during ion irradiation at 573 K increases initially with ion fluence and tends to saturate at a high fluence. The amount of the released deuterium is observed to be quantitatively consistent with that of the released tritium from similar standard pellets during neutron irradiation at 573 K.

Deuterium retention and release, ion irradiation, ↗