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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

An Investigation of Liquefaction in Irradiated TRIGA Fuel Exposed to Relatively High Temperatures

This report presents the findings of an investigation into high temperature fuel cladding chemical interactions (FCCI) in Training, Research, Isotopes, General Atomics (TRIGA) fuel rods. A TRIGA fuel-rod core or meat is principally composed of uranium (U) particles dispersed in a zirconium hydride (Zr H) matrix. The fuel is clad in sealed 304SS or Incoloy 800 tubes. At high temperatures, the fuel will interact with the cladding, resulting in FCCI. To investigate the FCCI (in this case, liquefaction), irradiated TRIGA fuel rods were exposed to relatively high temperatures in furnace tests. The tests were performed at different temperatures using segments from an irradiated TRIGA fuel rod at the Hot Fuel Examination Facility located at Idaho National Laboratory (INL). The data from this study is important for developing a better understanding of the performance of U-ZrH fuel type during transient events (e.g., a Loss of Coolant Accident (LOCA) event) that may result in high temperatures while operating a TRIGA reactor. Furnace tests were run for 6 hours at 730, 800, 900, 950, and 1,000°C. One test was run for 12 hours at 950°C. Post-test microstructural characterization was performed on the heat-treated samples using optical metallography to look for evidence of liquefaction. Additionally, microstructural characterization was performed using scanning electron microscopy to investigate the microstructure of the as-irradiated fuel before heat treatment and the microstructural changes that occurred after heat treating for the 730 and 950°C heat treatments for 6 hours. Results of the analysis were compared to those produced during diffusion studies performed using unirradiated TRIGA fuel meat and type 304 stainless steel. A gap is typically present in as-irradiated fuel that limits any possibility for chemical interaction between the fuel and cladding. Optical metallography (OM) was used to identify areas in the specimens heat treated at 950 and 1000°C where fuel and cladding contact was satisfactory for the FCCI to transpire. The bulk of the interaction zones that form due to FCCI developed in the fuel meat and not as much in the cladding. Even in the sample tested at 1,000°C, no evidence was found that gross fuel meat or cladding melting had occurred. Only limited porosity was found in the fuel meat that possibly could have been due to melting of a particular precipitate phase or phases that had formed during FCCI. These results agreed with those reported for the unirradiated diffusion couple studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1971

Investigation of high temperature fracture of T-111 and ASTAR-811C

The high temperature deformation and fracture behavior of T-111 and ASTAR-811C were studied over the temperature range 982 to 2205 C (1800 to 4000 F). As-cast and wrought-recrystallized material as well as GTA welds in sheet and plate were evaluated using conventional tensile and creep tests. Post test examinations were performed using optical metallography, scanning electron microscopy and transmission electron microscopy. A high temperature region of reduced ductility, in terms of tensile elongation, was identified for both alloys. The reduction in tensile elongation became more severe with increase in grain size, being near catastrophic for the as-cast specimens. Optical and electron metallography indicated that even for failures at very low total strain, considerable deformation of a very localized nature had occurred prior to fracture.

Gold, R. E.↗
NASA NTRS2000

Laser Brazing of High Temperature Braze Alloy

The Space Shuttle Main Engine (SSME) consists of 1080 conical tubes, which are furnace brazed themselves, manifolds, and surrounding structural jacket making almost four miles of braze joints. Subsequent furnace braze cycles are performed due to localized braze voids between the coolant tubes. SSME nozzle experiences extremely high heat flux (180 mW/sq m) during hot fire. Braze voids between coolant tubes may result in hot combustion gas escape causing jacket bulges. The nozzle can be disqualified for flight or result in mission failure if the braze voids exceed the limits. Localized braze processes were considered to eliminate braze voids, however, damage to the parent materials often prohibited use of such process. Being the only manned flight reusable rocket engine, it has stringent requirement on the braze process. Poor braze quality or damage to the parent materials limits the nozzle service life. The objective of this study was to develop a laser brazing process to provide quality, localized braze joints without adverse affect on the parent materials. Gold (Au-Cu-Ni-Pd-Mn) based high temperature braze alloys were used in both powder and wire form. Thin section iron base superalloy A286 tube was used as substrate materials. Different Laser Systems including CO2 (10.6 micrometers, 1kW), ND:YAG (1.06 micrometers, 4kW). and direct diode laser (808nm. 150W) were investigated for brazing process. The laser process variables including wavelength. laser power, travel speed and angle of inclination were optimized according to bead geometry and braze alloy wetting at minimum heat input level, The properties of laser brazing were compared to that of furnace brazing. Microhardness profiles were used for braze joint property comparison between laser and furnace brazing. The cooling rate of laser brazing was compared to furnace brazing based on secondary dendritic arm spacing, Both optical and Scanning Electron Microscope (SEM) were used to evaluate the microstructures of the braze materials and tube substrate. Metallography of the laser braze joint was compared to the furnace braze. SEM Energy Disperse X-Ray Spectra (EDX) and back scattered imaging were used to analyze braze alloy segregation. Although all of the laser systems, CO2, ND:YAG, and direct diode laser produced good braze joint, the direct diode laser was selected for its system simplicity, compactness and portability. Excellent laser and braze alloy coupling is observed with powder alloy compared to braze alloy wire. Good wetting is found with different gold based braze alloys. The laser brazing process can be optimized so that the adverse affect on the parent materials can be eliminated. Metallography of the laser braze joint has shown that quality braze joint was produced with laser brazing process. Penetration of the laser braze to the substrate is at neglectable level. Zero penetration is observed. Microstructure examinations shown that no observable changes of the microstructure (grain structure and precipitation) in the HAZ area between laser braze and furnace braze. Wide gaps can be laser brazed with single pass for up to 0.024 inches. Finer dendritic structure is observed in laser brazing compared with equiaxial and coarser grain of the furnace brazing microstructure. Greater segregation is also found in the furnace braze. Higher hardness of the laser braze joint comparing to furnace braze is observed due to the fast cooling rate and Finer microstructure in the laser brazing. Laser braze joint properties meet or exceed the furnace joint properties. Direct diode laser for thin section tube brazing with high temperature braze alloys have been successfully demonstrated. The laser's high energy density and precise control has shown significant advantages in reducing process heat input to the substrates and provide high quality braze joints comparing to other localized braze process such as torch, TIG, and MPTA processes. Significant cost savings can be realized particularly with localized braze comparing to a full furnace braze cycle.

Gao, Y. P.↗
NASA NTRS2010

Three-Dimensional Computed Tomography as a Method for Finding Die Attach Voids in Diodes

NASA analyzes electrical, electronic, and electromechanical (EEE) parts used in space vehicles to understand failure modes of these components. The diode is an EEE part critical to NASA missions that can fail due to excessive voiding in the die attach. Metallography, one established method for studying the die attach, is a time-intensive, destructive, and equivocal process whereby mechanical grinding of the diodes is performed to reveal voiding in the die attach. Problems such as die attach pull-out tend to complicate results and can lead to erroneous conclusions. The objective of this study is to determine if three-dimensional computed tomography (3DCT), a nondestructive technique, is a viable alternative to metallography for detecting die attach voiding. The die attach voiding in two- dimensional planes created from 3DCT scans was compared to several physical cross sections of the same diode to determine if the 3DCT scan accurately recreates die attach volumetric variability

Brahm, E. N.↗
NASA NTRS2014

Computational Process Modeling for Additive Manufacturing

Computational Process and Material Modeling of Powder Bed additive manufacturing of IN 718. Optimize material build parameters with reduced time and cost through modeling. Increase understanding of build properties. Increase reliability of builds. Decrease time to adoption of process for critical hardware. Potential to decrease post-build heat treatments. Conduct single-track and coupon builds at various build parameters. Record build parameter information and QM Meltpool data. Refine Applied Optimization powder bed AM process model using data. Report thermal modeling results. Conduct metallography of build samples. Calibrate STK models using metallography findings. Run STK models using AO thermal profiles and report STK modeling results. Validate modeling with additional build. Photodiode Intensity measurements highly linear with power input. Melt Pool Intensity highly correlated to Melt Pool Size. Melt Pool size and intensity increase with power. Applied Optimization will use data to develop powder bed additive manufacturing process model.

Bagg, Stacey↗
DOE OSTI2023

Spall strength of additively repaired 304L stainless steel

Additive manufacturing has the potential to repair damaged parts, but the performance of additive materials under high strain rate loading is still uncertain—especially with the added complexity of an interface with an existing wrought material. In this work, 304L stainless steel samples were intentionally damaged and then repaired with wire-fed laser additive manufacturing. The samples were subjected to shock loading to generate incipient spall. Velocimetry and post-mortem metallography results show that when the additive repair process parameters are optimized to reduce porosity and match the equation of state of the original material, the influence of the repair region on the shock propagation is negligible. In conclusion, the free-surface velocity profile and internal damage morphology of the repaired sample are shown to be practically identical to the pristine material.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Small-scale production of bespoke accelerated aging plutonium alloy

Here, this study demonstrates the 100 g scale manufacture of a plutonium alloy that ages at an accelerated rate. The resulting alloy ages six times faster than typical weapons-grade plutonium due to the addition of 238 Pu. As a major innovation, the process involved using a partial direct oxide reduction technique. This method was achieved by developing a new, complex geometry stirrer using additive manufacturing to reduce the 238Pu oxide and efficiently incorporate it into weapons-grade plutonium metal. The material was then purified by molten salt extraction and electrorefining before being alloyed with gallium. The alloy was then cold-rolled and annealed in a homogenization heat treatment. The resulting disk was characterized by metallography and differential scanning calorimetry, and the impurity content was determined using analytical chemistry techniques. The results show that a homogeneous delta phase plutonium alloy was achieved with expected microstructure and minimal impurities. This study was also successful in changing the plutonium isotopic composition by incorporating additional 238 Pu to accelerate the effects of radiation damage. This enables researchers to study long-term aging phenomena in a reduced time frame, thus avoiding the need for large-scale material production and circumventing the limitations of using naturally aged, archived plutonium.

Aging theory↗
DOE OSTI2020

235-F GoldSim Fate and Transport Model: Uncertainty Quantification

Building 235-F was configured with two missions in mind: Actinide Billet Line (ABL) and the fabrication of Pu-238 oxide for space program applications. ABL produced Np-237 billets for use in SRS reactors, whereas the design process, fabrication, and examination of Pu-238 oxide powder occurred within the following areas, respectively: Plutonium Experimental Facility (PEF), Plutonium Fuel Form (PuFF), and Old Metallography Lab (OML). By 1990 production ceased and by 2006 de-inventory occurred; however, assays have shown significant holdup remains within ABL and PuFF. As a result, 235-F is a Category 2 nuclear facility, with plans to undergo deactivation and decommission (D and D) via In-Situ Disposal (ISD). The purpose of this project is to ensure United States Environmental Protection Agency (USEPA) groundwater radiation maximum contaminant level (MCL) and dosage standards are met during the D and D of 235-F by quantifying uncertainty through probabilistic modeling and evaluation of various ISD alternatives. GoldSim is a dynamic modeling software package with a graphical, object-oriented interface capable of capturing the influence of complex system input variability on probabilistic system outcomes. A GoldSim stochastic fate and transport model for 235-F was developed and matched with a PORFLOW deterministic model to simulate probabilistic release and flow of radionuclides from ABL and PuFF into the vadose zone, the Upper Three Runs (UTR) Aquifer, and UTR Creek. The GoldSim model was used to probabilistically evaluate four ISD scenarios against USEPA groundwater radiation MCLs and dosage standards. The deterministic 235-F GoldSim fate and transport model continues to be refined to match the results of the PORFLOW deterministic model to ensure the model accurately represents radionuclide movement through the groundwater system. The stochastic variables that are utilized within the GoldSim model are founded on the most current data; a conservative perspective is taken where needed. Alignment with the PORFLOW deterministic model, coupled with input stochastic variability, allows the probabilistic 235-F GoldSim model to capture the conservative breadth of possible outcomes for radionuclide fate within this particular system. This ensures that the USEPA MCLs and dosage limits hold even in the worst case scenarios.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Image-Based Fracture Surface Defect Characterization Methods for Additively Manufactured Ti-6Al-4V Tested in Fatigue

Abstract Fatigue initiation in additively manufactured samples/parts often occurs at processed-induced defects such as lack-of-fusion (LoF), keyhole, or other morphological/microstructural defects that have unique characteristics and measurable qualities. Attempts at identifying and minimizing such defects have utilized optimized processing conditions along with in situ and ex situ characterization that includes metallography and/or X-ray computed tomography (XCT). This paper highlights the benefits of using fracture surface analyses to detect and quantify defects that may not be detected by metallography/XCT due to sectioning and resolution limits. In addition to using manual quantification of fatigue initiating LoF and keyhole defects on fracture surfaces, image-based machine learning using convolutional neural networks such as U-Net were also used to automate the process. Statistical analyses were used to identify the extreme cases of defects that initiated and accelerated fatigue and to model the distribution of defect size and shape characteristics to distinguish the type of defect. Initial results show agreement between trained machine learning models and ground truth data in defect segmentation, and the distributions of defect characteristics are distinguishable to particular process-induced defect types.

Materials Science↗
DOE OSTI2020

Detecting keyhole porosity and monitoring process signatures in additive manufacturing: an in situ pyrometry and ex situ X-ray radiography correlation

Creation of pores and defects during laser powder bed fusion (LPBF) can lead to poor mechanical properties and thus must be minimized. Post-build inspection is required to ensure the printed parts contain acceptably low defect concentrations. These inspections are time consuming and costly, especially for large or complex parts. As a potential solution, in situ process monitoring can be used to detect the creation of defects, characterize local material behavior and predict expected component properties. However, the precise relationship between pore creation and in situ process monitoring still needs to be understood. In this work, high-speed infrared diode-based pyrometry and high-speed optical imaging signals were used to monitor LPBF printing of 446 stainless steel 316 L single tracks with varying laser power and velocity. Results indicate an increase in pyrometer signal and melt pool dimensions with increasing laser power and decreasing velocity in agreement with previous work. In addition, careful analysis of pyrometer signal reveals a distinct signature of the conduction-to-keyhole mode transition which was confirmed by metallography. Critically, pore defect initiation as characterized by ex situ X-ray radiography was correlated with in situ thermal monitoring signals to derive the probability of defect creation. Our results show that, in principle, a probabilistic prediction of pore formation can be achieved based on in situ high-speed pyrometry monitoring of the LPBF melt pool.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The use of U3Si2/Al dispersion fuel for high power research reactors

In order to investigate the irradiation performance of U3Si2/Al dispersion fuel at relatively aggressive reactor conditions, compared to how this fuel is typically used in a research or test reactor, post irradiation examination was performed on two fuel plates that were irradiated at ~270 W/cm2 average surface heat flux and to a maximum local burnup greater than 8 × 1021 fissions/cm3 in the Advanced Test Reactor as part of the RERTR-8 experiment. As part of the non-destructive examinations that were performed on both fuel plates, detailed thickness measurements using a high-fidelity measurement bench were performed on one fuel plate and a segment of the second fuel plate. This paper provides the first reported results of such high-resolution thickness measurements for U3Si2/Al dispersion fuel plates, containing primarily the U3Si2 fuel phase, irradiated at such power levels to high burnup. For destructive examination using optical metallography, samples were generated from both irradiated plates. Localized regions of high swelling (characteristic of internal blistering) were observed on the surface of both fuel plates, suggesting that the performance limits had been reached for each plate. Destructive examinations showed that the fuel particles that are primarily U3Si2, which become amorphous during irradiation, in both plates contained relatively large fission gas bubbles with evidence of fission gas bubble interconnection.

Wachs, Daniel M.↗
DOE OSTI2020

Irradiation performance of nonfertile (Pu-MA-Zr) fast reactor metal fuels

This work was part of a program begun in 2001 to develop advanced nuclear fuels, originally as carriers for plutonium and minor actinides (neptunium, curium, and americium) taken from spent commercial light-water reactors (LWR) so that the plutonium and minor actinides could be ‘burned’ or transmuted in an accelerator or a fast nuclear reactor. A central part of these experiment programs has been the development of advanced fast reactor fuels, because a fast reactor was considered the most efficient vehicle to transmute the actinide waste products, and metallic fuels is a central focus of these tests. An experiment design was developed in which a thermal test reactor, the Advanced Test Reactor (ATR), was used to test small fuel pin prototypes, by creating areas in the core shielded by cadmium filters to produce a largely epithermal and fast neutron spectrum environment in which the pins could be irradiated. The results of non-fertile metallic fuel (no uranium) tests are presented here. Pu-Am-Np-Zr fuels were irradiated to fission densities up to 33 × 10 20 fission/cm 3 and Pu-239 depletions of up to 39%. The depletions were created by roughly 2/3 by fission and 1/3 by transmutation neutron capture. Up to five fuel ‘rodlets’ were irradiated in three sealed capsules stacked axially in the core, and the peak cladding temperatures ranged from 300°C to 500°C, depending on axial location as those near the core centerline are operating hotter and to higher fission densities. Several post-irradiation examinations (precision gamma scanning and fission gas release) were similar to other historical metal fuel experiments in fast reactors. However, optical metallography indicated that two of the rodlets had breached. The exact reasons are unclear. Due to the design of this irradiation experiment a rodlet breach could have increased the temperature in others in the same capsule by contaminating the thermal gap helium with heavier and less conductive fission product gases. Finally, some of those rodlets showed high amounts of fuel/cladding chemical interaction (FCCI).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Influence of neutron irradiation on Al-6061 alloy produced via ultrasonic additive manufacturing

Samples of aluminum alloy 6061 produced via ultrasonic additive manufacturing (UAM) were irradiated in the High Flux Isotope Reactor (HFIR) up to 17.3 dpa at ~70°C while in contact with water using perforated rabbit capsules. The irradiation campaign included as-received (AR) material, specimens subjected to various post-weld heat treatments (PWHTs, including hot isostatic pressing [HIP]), and reference (wrought) alloy samples. Mechanical tensile tests, accompanied by digital image correlation (DIC) analysis, fractography, and metallography, were performed as a part of the post-irradiation evaluation. The X- and Y-specimens (i.e., oriented in the sonotrode moving and vibration directions, respectively) showed pronounced radiation hardening and ductility decrease. Specific serration flow behavior and propagation of deformation bands were observed under various material conditions up to 3.5 dpa but disappeared at 17.3 dpa. In all cases, the fracture mechanism of X- and Y-specimens was ductile; ductile dimples dominated the fracture surface. Irradiated X- and Y-specimens showed good performance, regardless of material conditions (AR or PWHT). The performance of Z-specimens oriented in the build direction was strongly dependent on the PWHT. The AR and aged specimens showed fracture stress decrease with dose, and they experienced fracture under irradiation after 3.5 dpa; specimen cross section analysis revealed specific interface degradation that was likely related to corrosion. Recrystallization significantly improved in-reactor performance. Finally, HIP suppressed interface degradation due to recrystallization and pore removal, which led to good in-reactor performance for Z-specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Post-irradiation examination of low burnup U 3 Si 5 and UN-U 3 Si 5 composite fuels

Here this work presents post-irradiation examination data on UN-U 3 Si 5 and U 3 Si 5 fuels at low burnup (i.e., <10–15 GWd/tHM) with Kanthal AF® cladding. The results suggest good irradiation performance for both the silicide and nitride-silicide composite pellets. Optical microscopy revealed that the pellet-cladding gap is still open, and limited axial cracking was observed only in UN-U 3 Si 5 pellets. Microcracking was isolated to the U 3 Si 5 phase in all cases and was observed in pre-irradiation and depleted pellets, indicating that it was not irradiation induced. The fission gas release was minimal for the calculated fission density achieved (2.6 – 3.15 × 10 20 fiss/cm 3 ). No fission gas bubbles were observed in the optical metallography. These results suggest acceptable swelling and fission gas behavior for both the single phase and composite compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗