Search NASA⌕ Search

SEARCH · Search NASA

Results for “Fluence”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Mitigating Heavy Ion Irradiation‐Induced Degradation in p‐type SnO Thin‐Film Transistors at Room Temperature

The study investigates the mitigation of radiation damage on p‐type SnO thin‐film transistors (TFTs) with a fast, room‐temperature annealing process. Atomic layer deposition is utilized to fabricate bottom‐gate TFTs of high‐quality p‐type SnO layers. After 2.8 MeV Au 4+ irradiation at a fluence level of 5.2 × 10 12 ions cm −2 , the output drain current and on/off current ratio ( I on / I off ) decrease by more than one order of magnitude, field‐effect mobility ( μ FE ) reduces more than four times, and subthreshold swing (SS) increases more than four times along with a negative shift in threshold voltage. The observed degradation is attributed to increased surface roughness and defect density, as confirmed by scanning electron microscopy (SEM), high‐resolution micro‐Raman, and transmission electron microscopy (TEM) with geometric phase analysis (GPA). A technique is demonstrated to recover the device performance at room temperature and in less than a minute, using the electron wind force (EWF) obtained from low‐duty‐cycle high‐density pulsed current. At a pulsed current density of 4.0 × 10 5 A cm −2 , approximately four times increase in I on / I off is observed, 41% increase in μ FE , and 20% decrease in the SS of the irradiated TFTs, suggesting effectiveness of the new annealing technique.

Al-Mamun, Nahid Sultan↗
DOE OSTI2020

Effect of Ion Irradiation on Amorphous and Crystalline Ge–Se and Their Application as Phase Change Temperature Sensor

Research on phase change materials is predominantly focused on their application as memory devices or for temperature control which requires low phase change temperature. The Ge–Se binary chalcogenide glass system with its wide glass-forming region is a potential candidate for high-temperature and high-radiation phase change applications. In this study, the concept of employing Ge x Se 100– x glasses to monitor high temperature (450–528 °C) using the phase change effect, is reported. Materials selection, device structure, and performance of prototype sensors are analyzed. In addition, the effect of heavy ion irradiation by Xe ions with energies of 200, 600, and 1000 keV (fluence ≈10 14 cm –2 ) on the Ge x Se 100– x ( x = 30, 33, 40) thin films and phase change devices is studied. The irradiation effect on the amorphous and crystalline structure of the thin films is evaluated by Raman spectroscopy and X-ray diffraction (XRD). Although the changes in the structural units of amorphous films are negligible, in crystalline films orthorhombic-GeSe 2 crystals are found to be most affected by irradiation and a new phase, orthorhombic GeSe is found in the thin films after irradiation. The performance of a sensor with an active film of Ge 40 Se 60 is also shown as an example.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Characterizing hole trap production due to proton irradiation in germanium cross-strip detectors

We present an investigation into the effects of high-energy proton damage on charge trapping in germanium cross-strip detectors with the goal of accomplishing three important measurements. First, we calibrated and characterized the spectral resolution of a spare COSI-balloon detector in order to determine the effects of intrinsic trapping, finding that electron trapping due to impurities dominates over hole trapping in the undamaged detector. Second, we performed two rounds of proton irradiation of the detector in order to quantify, for the first time, the rate at which charge traps are produced by proton irradiation. We find that the product of the hole trap density and cross-sectional area, [nσ] h , follows a linear relationship with the proton fluence, F p , with a slope of (5.4 ± 0.4) x 10 -11 cm/p + . Third, by utilizing our measurements of physical trapping parameters, we performed calibrations which corrected for the effects of trapping and mitigated degradation to the spectral resolution of the detector.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2025

Neutron capture signatures determined from gadolinium isotope compositions of uranium ore concentrates using MC-ICPMS

We present Gd isotope compositions on 25 well-characterized uranium ore concentrates (UOCs). About half the UOCs have isotope depletions in 157 Gd coupled with anticorrelated excesses in 158 Gd due to neutron capture effects. UOCs from older ore bodies and with higher U contents show larger neutron capture effects than UOCs from younger ore bodies with lower U contents. These Gd data are correlated with their previously measured Sm isotope compositions, and we use the Sm data to estimate the neutron fluence of these samples. In conclusion, this work demonstrates how Gd isotope signatures can be employed as a new tool for nuclear forensics.

Earth Sciences↗
DOE OSTI2022

Impact of Low-Temperature Neutron Irradiation on Tensile Behavior of Base Metal and Electron-Beam Welded 316L Stainless Steel

Here, a study of 316L type stainless steel in its base metal and electron-beam (e-beam) welded conditions was performed to observe the effects of low-temperature (60 °C to 100 °C) neutron irradiation on the tensile behavior of the samples. Fractography was used in understanding the tensile-tested fracture surfaces of the 316L samples in these different forms with the characterization of the both base metal and welded samples using electron microscopy. Irradiation of the tensile specimens made free of defects of cutting and mechanical polishing showed a reduction in their tensile ductility with increased radiation-induced hardening up to 1.40 × 10 19 n/cm 2 (E > 0.1 MeV) fluence that corresponds to 1.1x 10 -2 dpa, even at the low irradiation temperatures. These low-temperature neutron irradiated base metal and e-beam welded 316L specimens also consisted of closely similar fracture surfaces characteristic of ductile rupture.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Light–Material Interactions Using Laser and Flash Sources for Energy Conversion and Storage Applications

Abstract This review provides a comprehensive overview of the progress in light–material interactions (LMIs), focusing on lasers and flash lights for energy conversion and storage applications. We discuss intricate LMI parameters such as light sources, interaction time, and fluence to elucidate their importance in material processing. In addition, this study covers various light-induced photothermal and photochemical processes ranging from melting, crystallization, and ablation to doping and synthesis, which are essential for developing energy materials and devices. Finally, we present extensive energy conversion and storage applications demonstrated by LMI technologies, including energy harvesters, sensors, capacitors, and batteries. Despite the several challenges associated with LMIs, such as complex mechanisms, and high-degrees of freedom, we believe that substantial contributions and potential for the commercialization of future energy systems can be achieved by advancing optical technologies through comprehensive academic research and multidisciplinary collaborations.

Materials Science↗
DOE OSTI2020

Structural damage and phase stability of Al 0.3 CoCrFeNi high entropy alloy under high temperature ion irradiation

In this study, an initially single phase high entropy alloy (HEA) Al 0.3 CoCrFeNi was irradiated by 3 MeV Au ions to a fluence of 6 × 10 15 cm –2 (~31 dpa at damage peak) at four different temperatures ranging from 250 °C to 650 °C. Transmission electron microscopy (TEM) and Atom probe tomography (APT) were employed to study the evolution of structural damage and phase stability with irradiation temperature. Al 0.3 CoCrFeNi exhibited a similar evolution of irradiation-induced defects with temperature as compared with conventional FCC alloys. At 250 °C and 350 °C, most of the visible irradiation-induced defects were faulted 1/3$\langle{111}\rangle$ dislocation loops. As the irradiation temperature increased to 500 °C, perfect 1/2$\langle{110}\rangle$ dislocation loops were observed along with the faulted loops. At the highest irradiation temperature 650 °C, only dislocation lines and networks could be observed. Regarding phase stability, the 3 MeV Au irradiation was observed to suppress the precipitation of (Ni, Al)-enriched nano clusters and the L12 ordered structure at irradiation temperatures 250 °C to 500 °C whereas precipitation of the B2 ordered structure was accelerated at 650 °C. This resulted in qualitatively opposite precipitation behavior between the ion irradiated damage region and unirradiated region at 500 °C and 650 °C. The opposite phase stability of the ion-irradiated damage region and unirradiated region at different temperatures is attributed to the competing effects of ballistic dissolution vs irradiation enhanced diffusion on precipitation.

36 MATERIALS SCIENCE↗
DOE OSTI2020

In-situ re-crystallization of heavily-irradiated Gd 2 Ti 2 O 7

Gadolinium titanate (Gd 2 Ti 2 O 7 , or GTO) and other lanthanide pyrochlores are interesting candidates for actinide waste disposal and fast ion conduction because the relevant material properties are intimately dependent on local cation structure. Therefore, a deep understanding of the kinetics associated with cation ordering and disordering is required if such material properties are to be tuned for specific device designs. To this end, single crystals of GTO were irradiated with 190 keV helium ions to a total fluence of 1 × 10 17 ions/cm 2 , amorphizing the sample surface to a depth of ~1 μm and resulting in significant He bubble accumulation. FIB lamellae lifted out from the irradiated sections were examined during heat treatment in the (scanning) transmission electron microscope. Two distinct stages of the re-crystallization of the amorphized material were observed. The material near the end of the ions’ range transformed first and with the same orientation as the pristine material. This was due to the close proximity of the pristine material and the presence of small defect fluorite seeds, but the propagation of this growth front was frustrated by the large pores (formerly bubbles) in the He accumulation layer. This was followed by heterogeneous nucleation of new crystallites at random orientations at the top of the He accumulation layer, which is attributed to the high surface area associated with the many small He bubbles in that region. Finally, it is inferred that the kinetics of grain growth in this material are significantly faster than the kinetics of grain nucleation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Helium effects on the surface and subsurface evolutions in single-crystalline tungsten

Tungsten (W) has been perceived as one of the most promising plasma facing materials (PFMs) for future fusion reactors. In the past decade, its behavior under irradiation and helium (He) plasma interaction has been extensively studied. However, some key knowledge gaps still exist, such as the influence of crystallographic orientation on the surface and subsurface evolutions. In this work, we focus on the He ion-beam irradiation damage effects in mirror-polished single-crystalline W samples with three different surface planes of {100}, {110} and {111}. Irradiation was performed at room temperature using 40 keV He+ to a fluence of 1 × 10 16 /cm 2 , followed by thermal desorption spectroscopy (TDS) up to ~1920 K. The microstructures of He-irradiated W before and after TDS heat treatment were characterized by scanning and transmission electron microscopy. Subsurface He bubbles were imaged in all irradiated samples, but newly formed <111>-oriented surface grains and surface blisters were only observed in W {100} and {110} starting orientations. These results reveal that radiation damage, He thermal desorption, and surface/subsurface evolution are all strongly dependent on crystallographic orientation. Underlying physical mechanisms are discussed based on ion channeling effects, He-vacancy interactions, and surface diffusion. In conclusion, these findings provide new insights into He effects in W.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Effect of sink strength on coherency loss of precipitates in dilute Cu-base alloys during in situ ion irradiation

In situ irradiations with 1 MeV Kr ions at 50~613 K up to a fluence of 6.25 × 10 14 ions/cm 2 (~1.25 displacements per atom, dpa) have been performed on pre-aged dilute Cu-0.9%Co, Cu-0.9%Fe and Cu-0.8%Cr alloys containing uniform matrix dispersions of coherent precipitates in order to study the effects of initial precipitate sink strength, damage dose and irradiation temperature on radiation-induced coherency loss of precipitates. Coherent precipitates with different point defect sink strengths (2πNd, where N and d are the precipitate density and diameter) were used in this work to examine potential differences in atomic relaxation during absorption of point defects. In all cases, irradiation to low doses (<~1 dpa) was very effective at inducing loss of precipitate coherency. At low sink strengths (~10 13 m -2 ), loss of precipitate coherency could be induced for doses ~0.01 dpa. This suggests there might be an efficient preferential medium-range strain-induced bias for absorption of interstitial defects due to tensile strains emanating from the undersized precipitates, which induces relatively rapid loss of coherency at low precipitate sink strengths. High precipitate sink strength (~10 14 m -2 ) conditions were relatively resistant to the radiation-induced loss of coherency (requiring higher doses approaching ~1 dpa) and this might be due to nearly equal numbers of interstitial and vacancy defects arriving at the precipitate interface for such high sink strength conditions. The precipitate coherency loss was observed to have a weak dependence on irradiation temperature. Molecular dynamics simulations confirm a strong effect of precipitate sink strength on the probability of interstitial absorption at precipitates.

36 MATERIALS SCIENCE↗
DOE OSTI2022

In-situ irradiation-induced studies of grain growth kinetics of nanocrystalline UO 2

The thermal conductivity of UO 2 fuel needs to be high enough to dissipate the heat generated from the fission reaction. Since grain size affects thermal conductivity and grain size can evolve with irradiation, it is critical to understand in-reactor UO 2 grain growth. Most studies of grain growth in UO 2 are based on thermally driven processes at elevated temperatures. However, studies have shown that grain growth can occur even at cryogenic temperatures by ballistic processes. Such irradiation-induced grain growth in UO 2 is yet to be studied. Advanced in-situ Kr ion irradiation and transmission electron microscopy were systematically performed on nanocrystalline UO 2 thin films at temperatures ranging from 50 K to 1073 K; grain growth was observed at all temperatures. A combination of manual and machine learning techniques was used to measure and plot grain size evolution against irradiation fluence at various irradiation temperatures. The machine learning method has significantly improved the analysis efficiency and reduced human labors. The grain diameter data were fitted using classical grain growth and thermal spike models to describe grain growth kinetics with and without irradiation effect. Grain growth during low temperature irradiation (≤ 475 K) can be well described by the thermal spike model. Above 475 K, there were additional thermally assisted processes that further accelerate the grain growth. At the highest irradiation temperature about 1075 K, both irradiation-induced dislocation loops and cavities/bubbles were observed to form in the UO 2 . In this report, the effects of irradiation-induced defects on grain growth kinetics are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Enhanced resistance to helium irradiations through unusual interaction between high-entropy-alloy and helium

Finding high performance plasma-facing materials (PFMs) is one of the most important and challenging tasks for realizing the commercial application of fusion reactors. Herein, we found the CrMoTaWV high entropy alloy (HEA) is highly resistant to low-energy and high-flux He plasma exposure. The nanochannel HEA film has 20 times higher initial fluence for the formation of fuzz and a remarkable 8.9 times slower fuzz growth rate than those of W. Combining the in-situ TEM observation and the Molecular dynamics (MD) simulation of the He bubble growth process, a new mechanism for the enhanced radiation resistance in HEA with the unusual interaction between HEA and He is found, where, differing from traditional metal, bubble growth in HEA leads to non-directional emission of interstitial atoms while HEA greatly suppress the growth of He bubbles. Additionally, the special nanochannel structure further rise the radiation resistance through releasing He out of the HEA film and reducing the He concentration. This new nanochannel refractory HEA material presents a promising choice as the PFMs with excellent performance and a much longer serving lifetime for future commercial fusion reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Grain boundary softening from stress assisted helium cavity coalescence in ultrafine-grained tungsten

The formation of helium cavities in coarse-grained materials produces hardening proportional to the number density and size of the cavities and due to the interaction of dislocations with intragranular helium defects. In nanostructured metals containing a high density of interfacial sinks, preferential cavity formation in the grain boundaries instead produces softening that is often attributed to enhanced interfacial plasticity. Here, employing two grades of ultrafine-grained tungsten, we explore this effect using targeted implantation studies to map cavity evolution as a function of the irradiation conditions and quantify its impact on the mechanical response through nanoindentation. Softening is reported at implantation temperatures above the threshold for preferential grain boundary cavity formation but at a sufficiently low fluence prior to the growth of intragranular cavities. Collective changes in the mean cavity size, density, and morphology beneath a residual impression on an implanted surface indicate that cavity coalescence accompanied the reduction in hardness. Complementary atomistic simulations demonstrate that, in tungsten grain structures exhibiting softening, grain boundary bubble coalescence is driven by stress concentrations that further act to localize strain in the grain boundaries through cooperative deformation processes involving local atomic shuffling and sliding, dislocation emission, and even the nucleation of unstable twinning events.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Response of 11 B enriched ZrB 2 ultra-high temperature ceramic to neutron irradiation at elevated temperatures

ZrB 2 , an ultra-high temperature ceramic (UHTC) is being considered for use in fusion reactor first-wall structures, yet its response to irradiation remains poorly understood. This study employed scanning/transmission electron microscopy (S/TEM), synchrotron X-ray diffraction (XRD), finite element calculations, and thermal property measurements to thoroughly investigate the neutron-irradiation effects on 11 B-enriched ZrB 2 . Neutron irradiations were conducted at 220 °C and 620 °C, with a neutron fluence of 2.2 × 10 25 neutron/m 2 (energy > 0.1 MeV), resulting in 3.9 dpa and 4200 appm He. The study revealed the unusual prevalence of prism loops and a > c anisotropic lattice swelling, likely linked to the low c/a ratio of ZrB 2 , leading to grain boundary microcracking. Reducing the grain sizes was effective in reducing intergranular cracking and macroscopic swelling. The observation of cavities in ZrB 2 irradiated at 620 °C, as opposed to 220 °C, prompts questions about the temperature at which vacancies in ZrB 2 become mobile, and the role of neutron absorption by 10 B in elevating irradiation temperatures. Isotopic enrichment in 11 B proves to be a viable strategy for mitigating helium production in transition-metal diborides, which is a critical consideration for nuclear applications. Irradiation-induced defects reduce the thermal diffusivity and conductivity of ZrB 2 by a factor of 4–9, which has important implications for its role as a plasma-facing material in fusion reactors that drive high heat fluxes through first-wall materials. Here, this comprehensive study lays the foundation for understanding ZrB 2 behavior under neutron irradiation and highlights important phenomena to consider for various material applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Fast neutron irradiation capability in existing thermal test reactors

In today’s nuclear industry, momentum towards the design, licensing, and construction of advanced nuclear demonstration plants, including fast reactors, is at a remarkably high level. However, there are currently no dedicated fast spectrum irradiation test facilities in the United States to support the development of fast spectrum technologies. As a result, a unique situation is developing where most of these plants will likely be designed by leveraging historic nuclear material technologies, but where the further optimization and advancement is impeded by the lack of fast neutron irradiation test facilities. While these circumstances present a challenge, there are some near-term opportunities that, if seized, can still help develop advanced fast reactor materials to a meaningful level of readiness to support future commercial fast reactors. Here, in this paper, we assess the feasibility of using thermal neutron filtering materials in existing experiment positions in the Advanced Test Reactor (ATR) at Idaho National Laboratory and the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory to simulate fast reactor test environments for nonfuel test specimens. Items investigated include the incident neutron flux (both fast and thermal), the total neutron fluence and cumulative atom displacements, helium production rate due to thermal neutron capture in nickel, and the potential impact that the thermal neutron filter material has on the cycle length of a given reactor. It is concluded that while HFIR provides the highest fast flux of all the options investigated, it is limited in the amount of thermal neutron filtering material that can be introduced into an experiment position without significantly affecting the operation of the reactor. Irradiation in Outboard-A positions in the ATR was found to be the most realistic near-term experiment avenue due to having ample space for several capsules in a moderately fast flux.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Physics analysis and design of heavy water reflected thermal test reactor

Here, this work investigates the option of modifying the Advanced Test Reactor by replacing the current beryllium reflector with heavy water. Such a change may provide some potential benefits for not only increasing the thermal irradiation capabilities but also resolving other problems such as reflector integrity issues due to fast fluence damage, which is always a limiting factor in the lifetime of the current beryllium reflector. This paper presents the analysis and estimation of the ATR core physics parameters by replacing the current beryllium reflector with heavy water (D 2 O). The paper first describes the details of two selected conceptual designs, which are partially reflected with either beryllium or graphite, and how they are derived from the baseline beryllium reflector concept. Then, reactor physics performance parameters for the two new concepts are assessed by comparing with those of the baseline concept. The performance parameters considered in this paper include in-pile tube neutron and gamma fluxes and heating rates, maximum loop voiding reactivity, core power behavior with different power splits, predicted cycle length with a given fuel loading, and thermal hydraulic analysis with a higher lobe power split. It is important to note that this study focuses on the reactor physics aspects and does not delve into the engineering challenges associated with such a design modification.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Uranium integral fission product yields for a spectrally-shaped 14.1 MeV neutron source at the National Ignition Facility

This paper describes the experimental results for an energy tuning assembly created to modify the National Ignition Facility deuterium–tritium fusion neutron source into a notional thermonuclear and prompt fission neutron spectrum, which has applications in integral measurements, nuclear data benchmarks, and radiation effects on microelectronics. The Monte Carlo neutron transport utilized MCNP5 to estimate the ETA-modified fluence using the ENDF-B/VIII.0 and IRDFF-II continuous energy nuclear data libraries, and SCALE Sampler was used to estimate the systematic nuclear data covariance using ENDF-B/VII.1 and IRDFF-II in a 252-group structure. The experiment fielded eight activation foils and a highly enriched uranium sample. This provided fifteen reaction channels that were used in a forward-fit comparison to the modeled results and to unfold the neutron spectrum using STAYSL. Gamma-ray spectrometry was performed on the activation and highly enriched uranium foils, and the reduced Χ 2 between the modeled and experimental values was 1.21. The results from the STAYSL unfold, reduced Χ 2 = 1.62, indicated that the modeled neutron spectrum was achieved and the systematic nuclear data uncertainty associated with the neutron transport and activation product cross sections was representative of the experiment. Integral cumulative fission product yield data were collected for 37 mass chains with a combination of gamma-ray spectrometry and radiochemical analysis. Fission product analysis was generally in agreement with two models using a semi-empirical fit and the General Observables of Fission code, with the exception of mass chains 88, 109, 111, 112, 113, 129, 139, 142, 144, 151, and 156.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Surface and subsurface characterization of laser-interference structured Ti6Al4V

Ti6Al4V specimens were exposed to high-energy laser pulses of 1.24 J/cm 2 fluences per pulse using a two-beam optical configuration to attain laser-interference. In addition to the formation of periodic surface textures by laser-interference, scanning electron microscope (SEM) images demonstrated several microstructural changes on the surface processed. Moreover, the surface and subsurface of Ti6Al4V specimens were characterized for the as-received and laser-interference processed conditions using x-ray photoelectron spectroscopy (XPS) and electron energy dispersive spectroscopy (EDS). The XPS data showed that this laser technique is effective at removal of surface contaminants for Ti6Al4V. The XPS analysis also revealed the formation of an oxy-nitride layer on the laser treated material. Finally, the microstructure analysis indicates that grain refinement, dendrite and twin structures were found near the surface due to laser processing.

36 MATERIALS SCIENCE↗