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DOE OSTI2020

Evaluation of Commercially Available Precursors for Obscuring Wet Industrial Irradiator Facility Pools to Delay Adversarial Source Removal (Part 2)

The behavior of commercially available potential obscurants for cobalt-60 ( 60 Co) wet-source storage industrial irradiator facilities (IRFs) were further evaluated for corrosive behavior of Nordion C-188 pencil stubs and obscurant properties under radiation exposure ( 60 Co). The potential obscurants studied included: titania aqueous dispersions (TAD - water soluble white paint), Chlorazol Black (CBOD - Chlorazol Black organic dye), powdered milk (COW - calcium obscurant in water), diatomaceous earth (DEA - diatomaceous earth additive), and rhodamine 6G (R6G). For corrosion efforts, stubs from an inert C-188 pencil-source rod were soaked in obscurant solutions and visually inspected. For radiation stability, obscurant samples were exposed to 60 Co radiation sources at 207 rad/s. The results from these studies reveal: the obscurants had no impact on the surrogate samples and may assist in terms of corrosion resistance; materials that did not rely on organic compounds to provide obscurance performed the best, as the organic compounds decomposed too rapidly in the high radiation environment, whereas the materials survived.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2022

Ultra-rapid, physics-based development pathway for reactor-relevant RF antenna materials

This paper presents a rapid, atomistically-informed, experimental development pathway for fusion reactor-relevant radio frequency (RF) antenna materials in the Cu-Cr-(Nb,Al,Zr) composition system, with the goal of improving upon GRCop-84. RF antennas in a tokamak fusion reactor will face a unique set of challenges as both structural and functional materials. The desired material must simultaneously achieve and maintain high electrical conductivity, high strength, high thermal conductivity, resist high temperatures, possess low nuclear activation, and incur low damage due to neutron bombardment. The GRCop-84 alloy serves as a starting point for iterative improvement, with the desire to reduce or eliminate Nb from the material to minimize nuclear activation. The rapid development pathway makes use of a multi-target combinatorial thick film sputtering process to produce full ternary phase diagrams on a Si wafer substrate. Transient grating spectroscopy (TGS), a laser-ultrasonic method, will determine spatially-varying thermo-elastic properties, while four terminal electrical conductivity measurements will map out the best per- forming regions of the sample for in-depth study at larger length scales. High energy proton and self-ion irradiation emulates the effects of neutron damage on the thermal/electric properties. With rapid turnaround time (∼days) in terms of mapping radiation damage-induced material property changes in the full ternary system, these techniques allow rapid iteration towards an optimal material, testing hundreds of nearby compositions in the time it took to test one. Focused testing of larger, single composition samples (produced in an arc furnace or by laser sintering) provides data on structural and high power RF properties, and validates our thick-film based workflow.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Non-destructive depth-resolved characterization of residual strain fields in high electron mobility transistors using differential aperture x-ray microscopy

Localized residual stress and elastic strain concentrations in microelectronic devices often affect the electronic performance, resistance to thermomechanical damage, and, likely, radiation tolerance. A primary challenge for the characterization of these concentrations is that they exist over sub-μm length-scales, precluding their characterization by more traditional residual stress measurement techniques. Here, in this study, we demonstrate the use of synchrotron x-ray-based differential aperture x-ray microscopy (DAXM) as a viable, non-destructive means to characterize these stress and strain concentrations in a depth-resolved manner. DAXM is used to map two-dimensional strain fields between the source and the drain in a gallium nitride (GaN) layer within high electron mobility transistors (HEMTs) with sub-μm spatial resolution. Strain fields at various positions in both pristine and irradiated HEMT specimens are presented in addition to a preliminary stress analysis to estimate the distribution of various stress components within the GaN layer. γ-irradiation is found to significantly reduce the lattice plane spacing in the GaN along the sample normal direction, which is attributed to radiation damage in transistor components bonded to the GaN during irradiation.

42 ENGINEERING↗

NASA NTRS1997

A Novel, Poly-Etalon, Fabry-Perot for Planetary Research

In an effort to develop a mechanically robust, high throughput and solid state spectrometer several liquid crystal Fabry-Perot etalons were constructed. The etalons were tested for spectral response, radiation resistance and optical transmission. The first year of this project was spent developing and understanding the properties of the liquid crystal etalons; in the second year an intensified all-sky imaging system was developed around a pair of LC etalons. The imaging system, developed jointly with SRI International represents a unique brassboard to demonstrate the use of LC etalons as tunable filters. The first set of etalons constructed in year one of this project were tested for spectral response and throughput while etalon surrogates were exposed to proton radiation simulating the exposure of an object in Low Earth Orbit (LEO). The 2" diameter etalons had a measure finesse of approximately 10 and were tunable over five orders. Liquid crystals exposed to proton irradiation showed no signs of damage. In year two two larger diameter (3") etalons were constructed with gaps of 3 and 5 microns. This pair of etalons is for use in a high resolution, all-sky spectral imager. The WATUMI imager system follows the heritage of all sky, narrow band, intensified imagers however it includes two LC Fabry-Perot etalons to provide tunability and the ability to switch wavelengths rapidly, an import consideration in auroral airglow imaging. This work also resulted in two publications and one poster presentation. The instrument will be uniquely capable, with superior throughput and speed, to measure optical airglow of multiple emission lines in harsh conditions.

Kerr, Robert B.↗

DOE OSTI2025

Implications of point defect accumulation on UO 2 thermal conductivity and fission gas release under accelerated fuel irradiation

Evaluation of thermal properties is a crucial factor for nuclear fuel performance. During reactor operation, the accumulation of fission products and irradiation-induced lattice defects are responsible for degradation in thermal conductivity. Consequently, it affects fuel temperature and fission gas release (FGR) among other Multiphysics processes important for economics and safety analysis. We analyze the implications of point defects (PD) accumulation described using a rate theory (RT) Model on lattice thermal conductivity of UO 2 . Here, we demonstrate that fission rate-dependent point defect concentrations have the largest impact on in-pile thermal conductivity in the periphery of light water reactor fuels below a temperature threshold governed by the migration barrier of defects. Our analysis provides a mechanistic description of this phenomena which current fuel performance codes treat empirically. The reduction of thermal conductivity in the low -temperature rim region acts as additional thermal resistance and leads to a temperature notably larger than suggested by Lucuta thermal conductivity correlation. These effects are anticipated to have notable impacts when fuels are exposed to accelerated radiation. The impact of such point defect-informed treatment of thermal conductivity on fuel performance is evaluated by a detailed analysis of fission gas behavior and its release. We consider several models capturing different stages of fission gas bubble evolution and fission gas release (FGR). Finally, a new fission rate-dependent correction to the Lucuta correlation is proposed. The results show a significant reduction in thermal conductivity at the fuels’ periphery and an increase in fuel centerline temperature specifically at low burnups. Ultimately a modified LC shows a higher FGR compared to the original LC, while the acceleration process results in a reduction in overall FGR.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

High temperature gamma radiation-induced chromium redox chemistry via in situ spectroscopic measurements

Chromium ions can make their way into the primary coolant of nuclear power reactors from the corrosion of stainless-steel reactor components, decreasing the material's corrosion resistance and resulting in increased transport of further corrosion products. Despite these potential effects, the radiation-induced redox speciation of chromium ions in aqueous solution is not well understood, especially at the elevated temperatures experienced by reactor coolants. Here, in the present work, we report new experimental results demonstrating that in aerated aqueous solution, the radiolytic oxidation of Cr(III) to Cr(VI) occurs at pH 4, while the reduction of Cr(VI) to Cr(III) occurs at pH 2. The oxidation of Cr(III) is primarily attributed to the reaction of the hydroxyl radical (˙OH) with the Cr(OH) 2+ species, while the reduction of Cr(VI) is attributed to reactions involving the hydrated electron (e aq - ) and hydrogen atom (H˙). Additionally, the steady-state equilibrium yield of Cr(VI) from the gamma irradiation of pH 4 Cr(III) solutions decreased with increasing temperature (over a range of 37–195 °C). This observation indicates that the activation energy of the Cr(VI) reduction reactions is higher than that for the Cr(III) oxidation reactions, such that it becomes relatively more favorable at higher temperatures. Overall, these data are important for the development of complementary multiscale models for the prediction of metal ion speciation in high temperature radiation environments.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2022

Overview of Base Model in Parametric Studies Specific to Performance of U-Mo Plates

This paper provides an overview of the base model specifically developed to perform parametric sensitivity studies on the U-10Mo monolithic fuel system. U-Mo monolithic fuels are being considered for the conversion of test reactors into high-performance research reactors that operate using proliferation-resistant, low-enriched uranium (LEU) fuels. These plate-type fuels contain a high-density, low-enrichment fuel sandwiched between zirconium diffusion barriers and encapsulated in aluminum claddings. All U.S. high-performance research reactors have released the designs of their LEU monolithic fuel reactor cores. These designs include nearly 50 distinct fuel plate geometries with different operational parameters. Consequently, a single generic plate geometry representing all the extreme points in this design matrix is unrealistic. To evaluate the performance for various parameters, a set of sensitivity studies was performed. These studies considered various input parameters (i.e., geometric, operational, and material property-related). The results revealed valuable information about plate performance and the sensitivity of this performance to various modeling inputs. To establish a reference state for comparing these result, base model featuring representative irradiation conditions was developed. To capture in-reactor behavior accurately, incorporation of representative constitutive models capable of evolving properties with respect to temperature, irradiation time, and burnup was needed. The behavioral models considered burnup-dependent properties, swelling, creep, and degradation. This paper introduces the base model created for the parametric sensitivity studies. The detailed description of the procedure includes the model geometry, model discretization, thermo-mechanical coupling, material properties and behavioral models. This paper also provides selected results and assesses the performance of the base model.

42 ENGINEERING↗

DOE OSTI2021

Thermoluminescence and optically stimulated luminescence response of Al 2 O 3 coatings deposited by mist-chemical vapor deposition

Aluminum oxide (Al 2 O 3 ) is a widely used ceramic material which can be applied as a protective coating on metallic structures to improve corrosion resistance and mechanical properties. Al 2 O 3 also has great potential as a functional thin film device in optoelectronics. When it is doped with carbon (Al 2 O 3 :C), this material system can serve as a reliable radiation dosimeter. The combination of optoelectronic, dosimetric, and protective properties demonstrated by Al 2 O 3 uniquely positions it for dual use applications in nuclear forensics, nonproliferation and safeguards, emergency response, retrospective dosimetry, and nondestructive inspection. Here, in this study, an innovative ultrasonic mist-chemical vapor deposition (Mist-CVD) system was used to deposit thin (nm level) Al 2 O 3 coatings on 304 stainless steel substrates. The single aliquot regeneration (SAR) and additive dose reconstruction techniques were then used to investigate the thermoluminescence (TL) and optically stimulated luminescence (OSL) response of the coated substrates to increasing doses of β irradiation. This study demonstrates that nm-thick Al 2 O 3 coatings hold promise as novel nano-dosimeters which could be leveraged for a series of applications such as nonproliferation and safeguards, retrospective dosimetry, and nondestructive inspection.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2020

Effect of Confinement of SIA Cluster Diffusion by Impurities on Radiation Defect Accumulation due to 14MeV neutrons in Tungsten

Impurities acting as traps, reduce the self-interstitial atom (SIA) cluster mobility, thereby increasing recombination and in turn enhancing radiation-resistance of polycrystalline materials. However, at elevated temperatures, depending on SIA-trap (impurity) binding energy, SIA clusters may trap and detrap (detach) numerous times. In tungsten, SIA clusters of all sizes glide one-dimensionally (1D) along their Burgers vector direction. Hence, detrapped SIAs would retrace their path, confining their 1D-glide between traps. However, small SIA clusters can change the direction of their 1D-glide by overcoming a rotation barrier. In addition, impurities or solutes are known to reduce this barrier. A lower rotation barrier can result in detrapped SIA clusters that diffuse in a random 1D direction, effectively leading to 3D (net-3D) diffusion. Here we investigate the effect of two cases of SIA diffusion, namely confined-1D diffusion and net-3D diffusion, on the damage accumulation in polycrystalline tungsten under 14 MeV neutron irradiation using the object kinetic Monte Carlo method. Simulations were performed using cascade debris obtained from molecular dynamics simulations with primary knock-on atom (PKA) energies corresponding to 14 MeV neutrons. In the simulations with net-3D diffusion, SIA clusters of all sizes diffuse in a random 1D direction after the clusters detrap. While for the confined 1D diffusion case, SIA clusters larger than size 5 retain their original direction of 1D-glide. In both types of simulations, trapped vacancies are assumed to be permanently immobilized. Unexpectedly, the damage accumulation was lower with confined 1D diffusion than that with net-3D. We present a systematic comparison of the influence of the two cases of SIA diffusion on the radiation damage accumulation as a function of dose rate, detrapping barrier, and impurity concentration at 1025 K.

Accumulation and recovery of lattice damage, tungs↗

DOE OSTI2020

Thermomechanical response of thickly tamped targets and diamond anvil cells under pulsed hard x-ray irradiation

In the laboratory study of extreme conditions of temperature and density, the exposure of matter to high intensity radiation sources has been of central importance. Here, we interrogate the performance of multi-layered targets in experiments involving high intensity, hard x-ray irradiation, motivated by the advent of extremely high brightness hard x-ray sources, such as free electron lasers and 4th-generation synchrotron facilities. Intense hard x-ray beams can deliver significant energy in targets having thick x-ray transparent layers (tampers) around samples of interest for the study of novel states of matter and materials’ dynamics. Heated-state lifetimes in such targets can approach the microsecond level, regardless of radiation pulse duration, enabling the exploration of conditions of local thermal and thermodynamic equilibrium at extreme temperature in solid density matter. The thermal and mechanical responses of such thick layered targets following x-ray heating, including hydrodynamic relaxation and heat flow on picosecond to millisecond timescales, are modeled using radiation hydrocode simulation, finite element analysis, and thermodynamic calculations. Assessing the potential for target survival over one or more exposures and resistance to damage arising from heating and resulting mechanical stresses, this study doubles as an investigation into the performance of diamond anvil high pressure cells under high x-ray fluences. Long used in conjunction with synchrotron x-ray radiation and high power optical lasers, the strong confinement afforded by such cells suggests novel applications at emerging high intensity x-ray facilities and new routes to studying thermodynamic equilibrium states of warm, very dense matter.

Meza-Galvez, J. (ORCID:000000032699849X)↗

NASA NTRS1993

Radiation effects in Ga(0.47)In(0.53)As solar cells

The effects of irradiating Ga(0.47)In(0.53)As p-i-n junctions with 1 MeV electrons were measured using Deep Level Transient Spectroscopy (DLTS) and both dark and illuminated (1 sun, air mass zero (AM0)) current-voltage (I-V) measurements. The I-V measurements were made over the range 100K less than T less than 350K. Temperature coefficients of the Ga(0.47)In(0.53)As photovoltaic parameters are presented which follow the same general behavior as other solar cell materials (e.g. Si and GaAs). Fits of the dark I-V data to the two term diode equation before irradiation were satisfactory, yielding an estimated band-gap energy of 0.79 eV. The recombination component of the dark current was found to increase linearly with fluence. DLTS detected two radiation-induced defect levels one shallow (E(sub c)-0.10 eV) and one near mid-gap (E(sub c)-0.29 eV), and it is the near mid-gap level which is expected to be the cause of the dark current increase. The radiation-induced degradation of the open circuit voltage is shown to be accurately predicted from the dark I-V measurements. The degradation of the open circuit voltage dominated the radiation response of the photovoltaic parameters while the short circuit current was only moderately degraded. This behavior is qualitatively explained in terms of the base thickness and dopant level. Appropriate changes in the device structure are suggested which should increase the radiation resistance. Isochronal thermal annealing induced recovery in the photovoltaic parameters at approximately 400K, coinciding with an annealing stage of the near mid-gap defect level.

Walters, R. J.↗

DOE OSTI2024

Effect of Thermal Aging on Microstructure and Stress Corrosion Cracking Behavior of an Alloy 152 1st Layer Butter Weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40- 80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is critical for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for different durations up to 75,000h (equivalent to 60 years of reactor service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged at 370°C and 450°C to an equivalent of 60 years of service. However, the first weld butter layer has high levels of deformation and is highly susceptible to SCC even in its non-aged condition. Nanohardness testing confirmed the extreme hardening with aging (ΔHV ≅ 100) at two locations within this weldment. In absence of LRO, hardening is suspected to be due to thermally-induced Cr carbide precipitation and coarsening. Testing in a primary water environment of the 1st layer of Alloy 152 weld butter aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. Similarly, the SCC CGR response of the aged weld butter does not appear to show a deterioration in performance, however, the difficult-to-test weldment geometry may affect the test results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Effect of thermal aging on microstructure and stress corrosion cracking behavior of Alloy 152 weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged to an equivalent of 60 years of service. Testing in a primary water environment of two heats of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. However, the SCC CGR response of the aged samples appears to show a deterioration in performance, confirming our previous observation.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Effect of thermal aging on microstructure and crack growth response of Alloy 152 weld

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray did not find evidence of LRO in any of the three heats aged to an equivalent of 60 years of service. Testing in a primary water environment of a heat of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. However, the SCC CGR response of the aged sample appears to show a deterioration in performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

From suppressed void growth to significant void swelling in NiCoFeCr complex concentrated solid-solution alloy

Void swelling can result in dimensional instability and undermine the safe operation of nuclear reactors. Current strategies to inhibit void swelling mainly focus on enhancing defect absorption and recombination by introducing high-density defect sinks. Complex concentrated solid-solution alloys (CSAs), including high-entropy alloys, can withstand severe radiation damage due to their inherent chemical complexity without interfaces. However, the underlying mechanisms for void suppression in CSAs are far from clear. In this research, we studied the void evolution with respect to irradiation depths, doses, and temperatures in equiatomic NiCoFeCr under 3 MeV Ni ion irradiations. At relatively low doses (16 and 54 displacements per atom, dpa), voids form mainly outside of the ion-damaged region, and void formation in the peak damage region is suppressed, leading to negligible swelling. However, with further increase of dose (86 up to 250 dpa), significant void growth occurs in the peak damage region and extended dislocation lines dominate instead of short dislocation lines and loops formed at lower doses. From 500 to 700 °C, the dislocation density decreases while dislocations grow. Although the overall void swelling increases dramatically from 500 to 580 °C at 54 dpa, void growth in the peak damage region is still suppressed. The transition from suppressed void growth to significant void swelling is attributed to dislocation evolution and local chemical inhomogeneity (enrichment of Fe/Cr in the matrix) at higher doses. Our study shows that controlling element diffusion and defect evolution through tuning chemical complexity can further enhance the swelling resistance of CSAs.

36 MATERIALS SCIENCE↗

NASA NTRS2004

NASA Tech Briefs, September 2004

Topics covered include: Brazing SiC/SiC Composites to Metals; Composite-Material Tanks with Chemically Resistant Liners; Thermally Conductive Metal-Tube/Carbon-Composite Joints; Improved BN Coatings on SiC Fibers in SiC Matrices; Iterative Demodulation and Decoding of Non-Square QAM; Measuring Radiation Patterns of Reconfigurable Patch Antennas on Wafers; Low-Cutoff, High-Pass Digital Filtering of Neural Signals; Further Improvement in 3DGRAPE; Ground Support Software for Spaceborne Instrumentation; MER SPICE Interface; Simulating Operation of a Planetary Rover; Analyzing Contents of a Computer Cache; Discrepancy Reporting Management System; Silicone-Rubber Microvalves Actuated by Paraffin; Hydraulic Apparatus for Mechanical Testing of Nuts; Heat Control via Torque Control in Friction Stir Welding; Manufacturing High-Quality Carbon Nanotubes at Lower Cost; Setup for Visual Observation of Carbon-Nanotube Arc Process; Solution Preserves Nucleic Acids in Body-Fluid Specimens; Oligodeoxynucleotide Probes for Detecting Intact Cells; Microwave-Spectral Signatures Would Reveal Concealed Objects; Digital Averaging Phasemeter for Heterodyne Interferometry; Optoelectronic Instrument Monitors pH in a Culture Medium; Imaging of gamma-Irradiated Regions of a Crystal; Photodiode-Based, Passive Ultraviolet Dosimeters; Discrete Wavelength-Locked External Cavity Laser; Flexible Shields for Protecting Spacecraft Against Debris; Part 2 of a Computational Study of a Drop-Laden Mixing Layer; Controllable Curved Mirrors Made from Single-Layer EAP Films; and Demonstration of a Pyrotechnic Bolt-Retractor System.

Source record↗

DOE OSTI2026

Fabrication of Catalytic Distillation Membranes with Atomic Layer Deposition

The integration of catalysts onto the surface of membranes enables simultaneous physical separation and catalytic transformation of constituents in a feed stream, facilitating improved contaminant removal and fouling mitigation. Distillation membranes are a particularly attractive platform for catalytic membranes because they reject nonvolatile species and exhibit exceptional resistance to oxidative and radical-driven degradation. However, imparting catalytic functionality onto hydrophobic, porous distillation membranes has proven challenging since the membranes used are chemically inert and difficult to modify. Furthermore, catalysts on the membrane surface can decrease hydrophobicity and increase the membrane’s susceptibility to pore wetting and failure. In this work, we create a catalytic distillation membrane by coating a polytetrafluoroethylene membrane surface with titanium dioxide (TiO 2 ) via plasma-assisted atomic layer deposition (ALD). By precisely tuning the ALD parameters, we demonstrate localized growth of TiO 2 near (within approximately 1 μm) the surface of polytetrafluoroethylene membranes, forming a catalytically active interface while preserving the underlying hydrophobic pore structure. Localized growth of TiO 2 is confirmed by electron microscopy and spectroscopy techniques, and membranes coated with 500 cycles of ALD show pressure tolerance up to 12.8 bar and higher than 95% salt rejection in pressure-driven distillation. Photocatalytic activity is demonstrated via the degradation of methylene blue dye under UV irradiation, where increasing TiO2 loading leads to an enhancement in dye degradation. These results establish a general strategy for integrating catalytic functionality into chemically inert, hydrophobic membranes without compromising distillation performance, providing a pathway toward multifunctional membranes that couple advanced oxidation with membrane separation for water treatment.

atomic layer deposition↗

NASA NTRS2010

Multiphysics Modeling for Dimensional Analysis of a Self-Heated Molten Regolith Electrolysis Reactor for Oxygen and Metals Production on the Moon and Mars

The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a self-heating reactor in which the electrolytic currents generate enough Joule heat to create a molten bath. In a first phase, a thermal analysis model was built to study the formation of a melt of lunar basaltic regolith irradiated by a focused solar beam This mode of heating was selected because it relies on radiative heat transfer, which is the dominant mode of transfer of energy in melts at 1600 C. Knowing and setting the Gaussian-type heat flux from the concentrated solar beam and the phase and temperature dependent thermal properties, the model predicts the dimensions and temperature profile of the melt. A validation of the model is presented in this paper through the experimental formation of a spherical cap melt realized by others. The Orbitec/PSI experimental setup uses an 3.6-cm diameter concentrated solar beam to create a hemispheric melt in a bed of lunar regolith simulant contained in a large pot. Upon cooling, the dimensions of the vitrified melt are measured to validate the thermal model. In a second phase, the model is augmented by multiphysics components to compute the passage of electrical currents between electrodes inserted in the molten regolith. The current through the melt generates Joule heating due to the high resistivity of the medium and this energy is transferred into the melt by conduction, convection and primarily by radiation. The model faces challenges in two major areas, the change of phase as temperature increases, and the dominance of radiative heat flux as heat transfer mechanism within the melt the change of phase concerns the regolith itself which is present in states ranging from a fine grain regolith with low thermal conductivity and low density to a vitrified melt with much higher thermal conductivity, and higher density. As the regolith is heated, it starts to soften around 1300 C the melt iS very viscous and evolving gas bubbles out in thick, lava-like fashion. By 1600 C the regolith is completely melted and the viscosity is low The second challenge resides in the proper modeling of the radiative heat flux requiring the addition of the computing-demanding radiative-heat-transfer function to the general heat transfer equation. The model Includes temperature-dependent properties (density, thermal conductivity, heat capacity, and viscosity, and absorption coefficients) and solves the radiative heat flux equation assuming gray (fine grains) and semi-transparent (melt) media and using an absorption coefficient spectral found in the literature for terrestrial minerals similar in composition to those of lunar regolith simulant

Dominguez, Jesus A.↗