SEARCH · Engineering Papers

Results for “material irradiation”

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.

SourceResearch connectionFrom yearThrough year

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

At least 523 records · Page 29

All sources NASA NTRS DOE OSTI

NASA NTRS2001

Energetic Ion and Electron Irradiation of the Icy Galilean Satellites

Galileo Orbiter measurements of energetic ions (20 keV to 100 MeV) and electrons (20-700 keV) in Jupiter's magnetosphere are used, in conjunction with the JPL electron model (less than 40 MeV), to compute irradiation effects in the surface layers of Europa, Ganymede, and Callisto. Significant elemental modifications are produced on unshielded surfaces to approximately centimeter depths in times of less than or equal to 10(exp 6) years, whereas micrometer depths on Europa are fully processed in approximately 10 years. Most observations of surface composition are limited to optical depths of approximately 1 mm, which are indirect contact with the space environment. Incident flux modeling includes Stormer deflection by the Ganymede dipole magnetic field, likely variable over that satellite's irradiation history. Delivered energy flux of approximately 8 x 10(exp 10) keV/square cm-s at Europa is comparable to total internal heat flux in the same units from tidal and radiogenic sources, while exceeding that for solar UV energies (greater than 6 eV) relevant to ice chemistry. Particle energy fluxes to Ganymede's equator and Callisto are similar at approximately 2-3 x 10(exp 8) keV/square cm-s with 5 x 10(exp 9) at Ganymede's polar cap, the latter being comparable to radiogenic energy input. Rates of change in optical reflectance and molecular composition on Europa, and on Ganymede's polar cap, are strongly driven by energy from irradiation, even in relatively young regions. Irradiation of nonice materials can produce SO2 and CO2, detected on Callisto and Europa, and simple to complex hydrocarbons. Iogenic neutral atoms and meteoroids deliver negligible energy approximately 10(exp 4-5) keV/square cm-s but impacts of the latter are important for burial or removal of irradiation products. Downward transport of radiation produced oxidants and hydrocarbons could deliver significant chemical energy into the satellite interiors for astrobiological evolution in putative sub-surface oceans.

Cooper, John F.↗

DOE OSTI2024

Formation of uranium oxy-carbide and uranium carbide via conversion of polymer covered uranium dioxide by laser-based thermal processing

Conventional formation of carbonaceous uranium compounds requires bulk processing using furnace-based approaches. Here, a methodology employing polymer covered uranium dioxide and laser-based heating is explored to enable rapid, localized formation of carbonaceous uranium compounds. Specifically, heating of poly(methyl methacrylate) covered uranium dioxide powder to high temperatures using laser irradiation in argon and methane gaseous environments was investigated. Decomposition of material and reactions induced by laser irradiation were probed in situ by residual gas analysis using a benchtop mass spectrometer. In this study, to determine the effect on the resultant material phase, three different process parameters were varied: gaseous atmosphere, laser power, and laser irradiation time. Material processed under varying conditions was analyzed using powder X-ray diffraction and scanning electron microscopy. This work realized the conversion of uranium dioxide into uranium oxy-carbide(s) and uranium carbide(s) phases, at over 60 wt.%, via the polymer surface application and laser-based thermal decomposition methodology.

36 MATERIALS SCIENCE↗

NASA NTRS1995

Laser Material Processing for Microengineering Applications

The processing of materials via laser irradiation is presented in a brief survey. Various techniques currently used in laser processing are outlined and the significance to the development of space qualified microinstrumentation are identified. In general the laser processing technique permits the transferring of patterns (i.e. lithography), machining (i.e. with nanometer precision), material deposition (e.g., metals, dielectrics), the removal of contaminants/debris/passivation layers and the ability to provide process control through spectroscopy.

Helvajian, H.↗

NASA NTRS2021

Reflection Modeling of the Black Hole Binary 4U 1630–47: The Disk Density and Returning Radiation

We present the analysis of X-ray observations of the black hole binary 4U 1630−47 using relativistic reflection spectroscopy. We use archival data from the Rossi X-ray Timing Explorer, Neil Gehrels Swift Observatory, and Nuclear Spectroscopic Telescope Array observatories, taken during different outbursts of the source between 1998 and 2015. Our modeling includes two relatively new advances in modern reflection codes: high-density disks, and returning thermal disk radiation. Accretion disks around stellar-mass black holes are expected to have densities well above the standard value assumed in traditional reflection models (i.e., n(e) ∼ 10^(15) per cu. cm). New high-density reflection models have important implications in the determination of disk truncation (i.e., the disk inner radius). This is because one must retain self-consistency in the irradiating flux and corresponding disk ionization state, which is a function of disk density and system geometry. We find that the disk density is n(e) ≥ 10^(20) per cu. cm across all spectral states. This density, combined with our constraints on the ionization state of the material, implies an irradiating flux impinging on the disk that is consistent with the expected theoretical estimates. Returning thermal disk radiation—the fraction of disk photons that bend back to the disk producing additional reflection components—is expected predominantly in the soft state. We show that returning radiation models indeed provide a better fit to the soft-state data, reinforcing previous results that show that in the soft state, the irradiating continuum may be blackbody emission from the disk itself.

Riley M. T. Connors↗

DOE OSTI2025

The state of the art for neutron irradiation experiments from the perspective of the High Flux Isotope Reactor (HFIR)

Irradiation experiment campaigns are critical to advancing nuclear energy technologies by providing data on material performance under relevant radiation conditions. Successful irradiation experiments require integrated design efforts that balance technical goals with facility constraints. Here, this paper presents an expert-informed overview of irradiation experiment design at the High Flux Isotope Reactor. It addresses the nuclear materials research and irradiation experiment communities to guide them toward developing technically sound, facility-compatible campaigns. The High Flux Isotope Reactor is a multipurpose reactor supporting isotope production, neutron scattering, and materials testing. Its high, steady-state neutron flux is ideal for irradiation experiments, but successful execution demands coordinated thermal, structural, and reactor physics analyses. The paper outlines the complete development workflow from concept definition and design optimization to safety qualification and post-irradiation examination. Standardized capsule platforms are also discussed in terms of flexibility, specimen capacity, and thermal performance. Common failure modes such as unanticipated geometric variations, can impact temperature-dose profiles and compromise data reliability. Therefore, detailed thermal modeling and accurate as-built characterization are essential for meaningful post-irradiation data interpretation. Key recommendations include early engagement all stakeholders, clearly defined design expectations, and alignment of specimen geometries with post-irradiation examination capabilities. This approach reduces design iterations, enhances data quality, and supports more efficient use of irradiation resources. Strategic and well-planned irradiation testing not only improves individual campaign success but also accelerates the deployment of advanced nuclear technologies. By closing critical data gaps and reducing development risks, the nuclear materials community can more effectively contribute to the future of clean, resilient energy systems.

Experiments↗

DOE OSTI2024

IASCC of 304 SS in BWR environments: Effects of post-irradiation annealing and surface condition

To investigate the impact of low-temperature post-irradiation annealing (PIA) on the stress corrosion cracking of neutron-irradiated 304 stainless steel, constant extension rate tests were conducted in simulated boiling water reactor normal water chemistry (BWR-NWC) and hydrogenated water chemistry (HWC) environments. Ten tensile samples, comprising five as-irradiated and five PIA specimens, were prepared by electropolishing the gauge section of electric discharge machined (EDM) samples. Here, the annealing treatment reduced the yield strength from approximately 550 MPa to around 425 MPa and significantly restored the ductility and the strain hardening capability of the alloy. Consequently, the susceptibility of this material to irradiation-assisted stress corrosion cracking (IASCC) was effectively mitigated, which is more prominent in HWC, as evident from fractography, which indicated a decreased propensity for intergranular (IG) fracture. Furthermore, it was observed that the polished surface facilitated crack initiation more readily than the EDM surface, suggesting that the EDM process suppressed crack initiation to some extent.

Crack initiation↗

DOE OSTI2021

Matrix Graphite Material Models In Pebbles and Compacts For Bison

The cores and reflectors in high-temperature gas-cooled reactors (HTGRs) are made of graphite materials, with the graphite acting as a moderator, a fuel host matrix, or the foundation for various structural components. This study aims to survey the models in the literature for graphite materials being used as host matrices in pebble/fuel compacts and to implement those surveyed models into Bison to conduct an early assessment of graphite's thermo-mechanical response under various reactor conditions. In this study, thermal (e.g., thermal conductivity, and specific heat capacity) and mechanical (e.g., elastic properties, thermal expansion, irradiation-induced dimensional changes, and irradiation-induced creep) material models for various graphite grades (e.g., H-451, IG-110, G-348, 2020, A3-3, and A3-27) are incorporated into Bison. Two benchmark problems are then exercised utilizing these new graphite-related capabilities: (1) modeling an Advanced Gas Reactor (AGR)-2 fuel compact, and (2) modeling the debonding of a particle-matrix interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Preliminary prediction of long-term aging and creep behavior of AM 316 SS

This report describes the development of initial mechanism models for the long term behavior of additively manufactured (AM), laser powder-bed fusion 316H stainless steel under the conditions expected in future advanced nuclear reactors. These models focus on key features of the material microstructure and response that differ from the conventionally-manufactured wrought material. Specifically, the report describes the development of models to capture the unique response of the AM material focusing on irradiation creep and swelling, the effect of internal stress, for example caused by dislocation structure, on precipitation, and the effect of the AM grain and dislocation structure on the macroscale creep and thermal aging behavior. This single mechanism models represent progress towards a complete, physics-based model for the long-term material behavior as well as elucidate key differences in the AM material behavior, when compared to the better-understood, conventionally-manufactured 316H.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Review of Recent Progress in Plasma-Facing Material Joints and Composites in the FRONTIER U.S.-Japan Collaboration

The plasma-facing components (PFCs) of future fusion reactors will have intricate structures and require multiple materials because no one material can simultaneously satisfy all the requirements of the component. The dissimilar material joints in PFCs must withstand extreme thermal and stress gradients under neutron irradiation. The Fusion Research Oriented to Neutron Irradiation and Tritium Behavior at Material Interfaces (FRONTIER) U.S.-Japan collaboration seeks to explore and explain the behavior of internal solid interfaces in PFCs under neutron irradiation. The first step of the collaboration was to identify the leading PFCs that should be studied further and prepare them for the next step, which will include neutron irradiation. Different strategies for material development are being pursued worldwide to produce robust PFCs. Here, in this work, an overview is presented of some of the most promising materials in the areas of copper alloys, tungsten-copper composites, tungsten-steel composites, additively manufactured tungsten, particle-reinforced tungsten, and tungsten and SiC fiber composites. Each material’s fabrication and benefits are described, and some discussion of remaining questions is given.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

In situ TEM annealing of neutron-irradiated Ti reveals a two-stage mechanism for elevated temperature radiation damage recovery

Understanding how irradiation-induced defects evolve at elevated temperatures is of critical importance to predicting materials' behavior under steady-state and accident scenarios. However, such mechanistic insight into microstructural evolution is limited by the nature of ex situ annealing and subsequent imaging. Here we show direct observation and quantification of defect recovery in neutron-irradiated Ti using in situ transmission electron microscopy (TEM) annealing experiments. In agreement with our prior work, and at temperatures below the irradiation temperature (T irr = 300 °C), dislocation loops are observed to glide. At elevated temperatures (>500 °C), dislocation lines become mobile and promote significant recovery of the microstructure. These mechanisms challenge the established electron irradiation-based model for radiation damage recovery, which originally suggests dissolution of static defect clusters, and demonstrates the importance of in situ characterization in understanding defect evolution in irradiated materials.

Hirst, Charles A. [Univ. of Michigan, Ann Arbor, M↗

NASA NTRS2023

Investigating the Role of Incident Ion Flux in Solar Wind Space Weathering of Carbon-Rich Asteroidal Regolith Via H+ and He+ Irradiation of the Murchison Meteorite

We present results from a set of low and high flux 1 keV/amu H + and He + irradiation experiments performed on slabs of the Murchison CM2 carbonaceous chondrite. The low flux conditions for H + and He + irradiation were ~1-1.5 orders of magnitude lower than the high flux conditions, and each experiment was irradiated to a total fluence between ~3 × 10 16 to ~6 × 10 16 ions/cm 2 . Irradiation-induced changes in the surface chemistry and optical properties of the Murchison samples were evaluated using in situ X-ray photoelectron spectroscopy (XPS) and visible and near-infrared spectroscopy (VNIR). We characterized the microstructure and composition of ion damaged rims in focused ion beam (FIB) cross-sections extracted from olivine and matrix material in each irradiated Murchison slab using transmission electron microscopy (TEM). XPS results suggest that both low flux and high flux H + and He + irradiation cause minor sputtering of surface carbon as well as a reduction in the valence state of iron, from Fe 3+ to Fe 2+ . Slope bluing is observed in VNIR spectra of the irradiated samples which may reflect carbonization and dehydrogenation of organic species and contrasts with reddening trends associated with npFe 0 formation. Although we do not observe a strong flux-dependence on the crystallinity of ion-damaged olivine, TEM analyses reveal a variety of microstructures in all olivine FIB-sections, suggesting that crystallographic orientation affects amorphization efficiency. Analyses of matrix FIB-sections indicate that phyllosilicate alteration is mainly driven by He+ irradiation, where the higher incident flux leads to greater amorphization and the formation of more distinct ion-damaged layers, similar to smooth layers in returned Ryugu particles. TEM results also provide some evidence that higher ion flux leads to greater vesiculation, with He + irradiation being more efficient at vesiculation than H + irradiation, and that higher ion flux may promote the segregation of Mg and Si into laterally extensive lenses and layers in olivine samples. We discuss the implications of these findings for constraining the role that ion flux plays in the development of space weathering characteristics in silicate phases present in carbonaceous asteroidal regoliths. These results will be important for understanding the complexity of this process and how it operates on carbon-rich airless bodies like asteroids Bennu and Ryugu.

meteorites↗

DOE OSTI2026

Analysis of attenuation data from the decommissioned ZIon unit 1 reactor pressure vessel beltline weld

In order to examine the attenuation of radiation damage through the thickness of an irradiated reactor pressure vessel (RPV), four segments were acquired from the Zion Unit 1 power plant RPV after the plant was decommissioned. The Zion Unit 1 RPV Beltline Weld Segment 1 was cut into seven blocks, consisting of five base metal and two beltline welds from the high fluence region of the segment. Through-wall test specimens were machined and tested. Specimens included those used for Charpy impact, Master Curve fracture toughness testing, and chemical analysis. The observed through-thickness ductile-to-brittle transition temperatures in the beltline weld deviated significantly from the expected behavior based on the attenuation of fast fluence as a function of depth into the RPV. Beginning at the inside surface, the 41-J Charpy transition temperature was either flat or slightly increasing until the ¾ -T location. The results of a simple, model-based analysis of the Zion beltline weld material that included the irradiation conditions and material chemistry were generally consistent with industry trend curves and the standard attenuation model, rather than the observed data. Although there was no archive material from the RPV available to permit measurement of the unirradiated properties, fracture toughness specimens fabricated from archive surveillance weld were used to obtain an estimate of the initial through-thickness values of the Charpy transition temperature. The Charpy shifts obtained using this approach were similarly in disagreement with the predictions of the US NRC Regulatory Guide 1.99, Rev. 2. However, testing of irradiated Charpy specimens taken from the RPV following post-irradiation annealing (10 hr. at 500 °C) provided a quite different estimate of the unirradiated properties which improved the agreement between the inferred through-thickness Charpy shifts and exponential attenuation model included in Regulatory Guide 1.99/2. In conclusion, the analysis of the Zion data and data obtained in previous post-mortem examinations of decommissioned RPVs indicates that more work is needed to understand the through-thickness properties of RPV materials in order to properly assess through-wall damage attenuation.

Charpy impact↗

DOE OSTI2020

A macro-scale ruck and tuck mechanism for deformation in ion-irradiated polycrystalline graphite

A vein structure, which becomes more pronounced with increasing ion dose, was found on the surface of polycrystalline HOPG (highly oriented pyrolytic graphite) implanted by ex situ C + (up to 1.8 × 10 17 ions/cm 2 ), and in situ Ar + in a transmission electron microscope (TEM). These veins are found to be independent of the crystallographic orientations and are associated with the formation of pores. Underneath the veins, a triangular-shaped core was formed with the graphite platelet inside the core displaced up towards the surface. A macro-scale ‘ruck&tuck’ geometry was thus generated at these triangle structure boundaries. Progressive movement of dislocations along basal planes during irradiation was observed, and a mechanistic model was proposed on this basis to explain the vein formation. A small increase of c-spacing was observed with irradiation but it is believed that macro-scale vein formation plays a more vital role in the dimensional and property changes in polycrystalline graphite, especially when a stress gradient is present. Here, the model proposed also explains the change of thermal expansion in HOPG with irradiation. Together with Heggie’s ‘ruck&tuck’ and Barsoum’s ‘ripplocations’ models, the present model is considered to have provided an additional experimentally proven mechanism responsible for irradiation behaviour in graphite materials.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Effect of Helium Flux on Near-surface Helium Accumulation in Plasma-Exposed Tungsten

We report results of object kinetic Monte Carlo (OKMC) simulations aimed at understanding the effect of helium flux on the near-surface helium accumulation in plasma-facing tungsten, which is initially defect-free and has a W(100) surface orientation. These OKMC simulations are performed at 933 K for fluxes ranging from 10 22 to 4 × 10 25 He/m 2 s, with 100 eV helium atoms impinging on a W(100) surface up to a maximum fluence of 4×10 19 He/m 2 . In the near-surface region, helium clusters interact elastically with the free surface. The interaction is attractive and results in the drift of mobile helium clusters towards the surface as well as increased trap mutation rates. The associated kinetics and energetics of the above-mentioned processes obtained from molecular dynamics simulations are also considered. The OKMC simulations indicate that as the flux decreases, the retention of implanted helium decreases, and its depth distribution shifts to deeper below the surface in initially pristine tungsten. Furthermore, the fraction of retained helium diffusing into the bulk increases as well, so much so that for 10 22 He/m 2 s, almost all of the retained helium diffused into the bulk with minimal/negligible near-surface helium accumulation. At a given flux, with increasing fluence, the fraction of retained helium initially decreases and then starts to increase after reaching a minimum. The occurrence of the retention minimum shifts to higher fluences as the flux decreases. Although the near-surface helium accumulation spreads deeper into the material with decreasing flux and increasing fluence, the spread appears to saturate at depths between 80 and 100 nm. Finally, we present a detailed analysis of the influence of helium flux on the size and depth distribution of total helium and helium bubbles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2020

Helium effects and bubbles formation in irradiated Ti 3 SiC 2

Ti 3 SiC 2 is a potential structural material for nuclear reactor applications. However, He irradiation effects in this material are not well understood, especially at high temperatures. Here, we compare the effects of He irradiation in Ti 3 SiC 2 at room temperature (RT) and at 750 °C. Irradiation at 750 °C was found to lead to extremely elongated He bubbles that are concentrated in the nano-laminate layers of Ti 3 SiC 2 , whereas the overall crystal structure of the material remained intact. In contrast, at RT, the layered structure was significantly damaged and highly disordered after irradiation. Our study reveals that at elevated temperatures, the unique structure of Ti 3 SiC 2 can accommodate large amounts of He atoms in the nano-laminate layer, without compromising the structural stability of the material. The structure and the mechanical tests results show that the irradiation induced swelling and hardening at 750 °C are much smaller than those at RT. Overall, these results indicate that Ti 3 SiC 2 has an excellent resistance to accumulation of radiation-induced He impurities and that it has a considerable tolerance to irradiation-induced degradation of mechanical properties at high temperatures.

36 MATERIALS SCIENCE↗

DOE OSTI2026

On the uncertainties in helium generation predictions for fission and fusion alloys

With ongoing advances in fusion and advanced fission reactors, quantifying irradiation effects in materials is critical. Transmutation-induced helium in cladding and structural materials can drive swelling and embrittlement, thereby reducing these components’ lifespans. Yet most studies ignore the considerable uncertainties in predicting helium generation rates. In this work, we created a code wrapper, F-SCATTER, that automatically performs simulations in FISPACT-II. We used this tool to investigate potential variance in helium generation rate, or He/dpa, calculations based on deviations in alloy composition, irradiating neutron flux spectrum, computational methodology, and nuclear data sources. We used 12 wt% Cr HT9 steel as the reference case and observed a 6.5%–98.3% He/dpa spread based on compositional variation within a single chemical specification, a 1.8%–11.5% He/dpa variation upon the incorporation of a 15% artificial uncertainty in flux at each energy, and a He/dpa difference as high as 231% when using ENDF/B-VIII.0 versus TENDL-2021 data libraries. Similar results were found for other prominent iron-based alloys, including Grade 91, castable nano-structured alloy, and 316H—where additional variations exist based on reactor type (e.g. thermal, fast, or fusion) and alloying elements such as carbon, nitrogen, and nickel. Based on the simulated results, we conclude that a significant part of the heat-to-heat variability in swelling responses of Fe-based alloys can be driven by impurity content in alloy compositions, and, therefore, chemical control should be a key element in supply chain design for advanced nuclear energy systems. Furthermore, we provide critical recommendations on best practices for evaluating and reporting helium production and lattice damage rates when computing predictions with multiphysics programs such as FISPACT-II.

FISPACT-II↗

NASA NTRS2003

Material for surface-enhanced Raman spectroscopy, and SER sensors and method for preparing same

Metal-doped sol-gel materials, suitable for use as sensors for surface-enhanced Raman spectroscopic analysis for trace chemical detection, are produced by effecting gelation and solvent removal of a doped sol-gel under mild temperature conditions. At least in certain instances reaction and drying will desirably be effected in an oxygen-starved environment. The metal of the sol-gel material functions, when irradiated, to produce a plasmon field for interaction with molecules of an analyte in contact therewith, increasing by orders of magnitude Raman photons that are generate by excitation radiation, and the method allows matching of the metal and metal particle size to a wavelength of light (or incident radiation, e.g., laser radiation) to generate surface plasmons. The porosity of the sol-gel material dramatically increases the surface area, and thereby the amount of metal exposed for analyte interaction. The sensors provided may be in the form of glass vials, fiber optics, multi-well micro-sample plates, etc., having surface coatings of the doped sol-gel material, to provide sampling systems for use in a Raman instrument.

Farquharson, Stuart↗

DOE OSTI2022

Fascinating physics at the edge of magnetic fusion devices

Abstract The physics of the processes at the edge of magnetic fusion devices is multifaceted and exhibits complex, nonlinear synergistic effects. Even though this region occupies only a small portion of the whole device, it plays a crucial role in overall plasma confinement, heat exhaust, and plasma–wall interactions. The latter affects not only the performance but also the lifetime of plasma-facing components and therefore remains an outstanding challenge for future fusion reactors. At the edge of fusion devices, researchers are dealing with phenomena including classical and anomalous plasma transport, atomic physics effects, and physics of plasma-facing material under strong irradiation by particle and energy fluxes. The diversity of edge physics makes it particularly attractive for young scientists. Working in this field, they can find endless possibilities to demonstrate their talents and creativity. This short review describes just some of the basic scrape-off layer and divertor plasma phenomena including divertor plasma detachment, intermittent bursts of anomalous cross-field plasma transport, plasma–material interactions, and dust in fusion plasmas, which are of particular interest in relation to fusion reactors.

Physics↗