SEARCH · Engineering Papers

Results for “SCANNING ELECTRON MICROSCOPY”

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 1,279 records · Page 71

All sources NASA NTRS DOE OSTI

DOE OSTI2020

Radiation-Induced Modifications in Copper Oxide Growth

Radiation-induced effects and their influence on oxidation processes were evaluated for their use as a forensic tool for special nuclear material (SNM). A beam of 10 MeV Au³⁺ ions was used to mimic the accumulation of microstructural damage from self-irradiation through the decay of radionuclides. Several copper samples were irradiated as suitable surrogate materials at 200 °C with a flux of 1x10¹² ions/cm·s to damage levels of 5, 10, and 15 displacements per atom (dpa). This corresponds to about 50, 100, and 150 years, respectively, of accumulated α-decay damage in a PuGa alloy assuming a damage rate of 0.1 dpa/year. After irradiation, all samples were exposed to an accelerated aging process induced by thermal treatment at 350 °C for 1 hour in air. This resulted in the growth of a mixed oxide layer (Cu₂O and CuO) which was characterized in detail using several complementary analytical techniques: Scanning Electron Microscopy, Raman spectroscopy, Synchrotron X-ray diffraction (transmission mode), and Grazing Incidence X-ray diffraction. The oxide layer growth of irradiated Cu at 350°C is distinctly modified as a result of the ion irradiation. Most notably, the growth of the CuO phase is suppressed with increasing radiation damage on the Cu substrate, and structural changes occurred in the Cu₂O phase. These results indicate that damage from self-irradiation over time can cause quantifiable modifications in the oxidation process of metals that could be harnessed for their use as a novel forensic tool.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Aluminum Spent Fuel Performance in Dry Storage Task 4 Aluminum Oxide Sampling of ATR Dry Stored Fuel

Milestone report on sampling of long-term dry-stored ATR spent fuel elements. It describes tooling, sample acquisition methods and analysis of the aluminum oxide samples by scanning electron microscopy, thermogravimetric analysis, X-ray diffraction and transmission electron microscopy. The observations provide input to projections on controls necessary for maintaining the integrity of the fuel during extended interim storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Mechanical Properties and Deformation Behavior of Additively Manufactured 316L Stainless Steel (FY2020)

The Transformational Challenge Reactor (TCR) program plans to build most of the TCR core components through additive manufacturing (AM) processes. These processes include laser powder bed fusion (LPBF) for the metallic (316L) components and the newly developed combined process of binderjet printing and chemical vapor infiltration (CVI) for the SiC fuel matrix. Mechanical testing and characterization tasks have been carried out since the beginning of the TCR program to (1) build a property database for the AM materials that will be used in TCR core and (2) to assess the materials’ performance in TCR-relevant conditions. This document reports the outcome of the testing and characterization efforts for the fiscal year with a focus on the mechanical performance data of AM 316L stainless steel (SS). Baseline tensile testing over a wide temperature range of room temperature–600 °C was completed for the AM 316L alloy in as-built, stress-relieved, and solution-annealed conditions. The as-built 316L showed the highest strength, and the alloy after the post-build treatments showed reduced strengths in the low-strain range. However, the strength differences among the AM materials became insignificant in the later part of deformation. Furthermore, regardless of post-build processing, the AM 316L SS showed higher strength and comparable ductility when compared with wrought 316L SS. Thermal creep testing and microstructural evolution during creep deformation were also performed under selected conditions. It was found that the AM 316L steel showed the best creep resistance in the stress-relieved condition. In-situ tensile tests were performed using scanning electron microscopy and 1-ID beamline at the Advanced Photon Source to elucidate the deformation and fracture behavior of AM 316L and the evolution of crystalline stress, dislocations, and pore distribution. Using these in-situ testing data, an in-depth analysis of the roles of microstructural features in deformation and fracture processes is presented herein. The final section of the document introduces ongoing and future activities for materials testing and characterization, including irradiation effect studies and ball punch testing on AM 316L and AM IN718 alloys.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Impact Response of Cold Spray Deposited Materials

Solid-state cold spraying (CS) of metals and respective blends is becoming increasingly attractive compared to conventional high temperature processes due to the unique properties such as increased yield strength, low ductility, and differences in tensile and compressive strengths that result from microstructural features due to the CS process. Here we report the results of plate impact experiments applied to CS deposits of tantalum (Ta), niobium (Nb), and a tantalum- niobium blend (TaNb). These methods allowed for definition of the Hugoniot for each material type and allowed for assessment of the Hugoniot Elastic Limit (HEL). Scanning electron microscopy was used on recovered samples to characterize the fracture mechanism during spallation.

36 MATERIALS SCIENCE↗

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↗

DOE OSTI2020

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

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

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Examination of Intergranular Crack on SC0234

The microstructure of the SC0234 weld was evaluated per DA-SPEC-01067, ''General Purpose Heat Source (GPHS) Plutonium-238 Dioxide Fueled Clad Product Specification'' and it was determined that the weld microstructure conformed to the quality requirements. During the examination of weld cross section, the metallographic analysis revealed the presence of a surface crack. Prior visual examination per RPS-TA55-303 had not revealed the crack, nor did the surface photography performed per RPS-TA55-302. The presence of the crack does not violate any of the product specifications called out on DA-SPEC-01067. Although there is a well-documented history of weld hot crack susceptibility in the early development of the DOP-26 iridium alloy, a surface crack such as this had not been documented during welding at Los Alamos. In the interest of better understanding the cause; the product engineer/program recommended further examination of the crack. This report discusses the examination of the crack found in the weld cross section of simulant clad SC0234. Metallography, Scanning Electron Microscopy (SEM) and Energy Dispersive Spectroscopy techniques (EDS) where used in the analysis of the crack and the surrounding weld microstructure.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Blast furnace slag reactions in various solutions (Interim Report)

Blast furnace slag (BPS) is a critical component of grout formulations for the remediation of low activity nuclear waste. This interim report describes the initial results of a study designed to investigate the fundamental mechanisms of BPS dissolution. The goal of the work is to delineate the physicochemical aspects of the slag that impart beneficial properties to grout formulations to enable the consistent production of a slag-replacement material (“designer slag”) that performs better than what is commercially available. This report documents the dissolution experiments that were conducted using nine different initial solution compositions. Solution measurements for each solution, including pH and solution composition, are given for 1, 7, 14, 28, and 56 day intervals samples. Scanning electron microscopy images are presented for the 56 day samples from each of the solutions. Inferences about reaction mechanisms and 56-day mineralogy are made based on the information collected to date. Some conclusions are made based on the work performed for this interim report, and a path forward including the strategy for making stronger conclusions based on X-ray diffraction (XRD) analyses is presented.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Wear in particle based CSP systems from particle abrasion and attrition at high temperature

High temperature particle-based receivers offer distinct advantages over conventional molten salt receivers due to their ability to achieve temperatures above 700 ºC, direct absorption of solar energy as they fall through a beam of concentrated sunlight, and the relative ease of storage (and retrieval using a secondary working fluid) in insulated storage tanks. The use of particles, however, also raises concerns with material degradation from the flow of hot or cold particles through discharge hoppers or along the inner receiver surfaces and other system components (e.g. tubes, valves etc.), depending on operating mode of the receiver. The flow of particles over surfaces may result in loss of material from abrasive wear, impact erosion from impingement under gravitational fall and particle attrition as particles fall and move on top of each other. In the present study, the performance of candidate materials and particles were evaluated through a series of abrasion erosion and particle attrition experiments at room temperature as well as at 800 °C. Candidate materials were subject to abrasive wear from particles at low particle to material velocities inside a resistance heated kiln, and analyzed for changes in mass and surface morphology using cross-sectional scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDS) tests. The wear rate for different specimens was noted to be largely driven by the strength of chromia scales built on the specimens from exposure to high temperature. Particle attrition measurements explored the susceptibility of particles to breakdown from particle to particle interaction and the generation of fines from this process. At the low velocities expected in particle based CSP plants, the particles tested exhibited near negligible breakdown. However, changes in the particle hardness at 800 ºC resulted in a significantly higher particle breakdown to sizes <40 microns raising potential environmental concerns. In addition to particle breakdown, it was also noted that the sample had oxides from the stainless steel test setup mixed in with the particles. Similar oxides can be expected to turn up in the utility scale particle based CSP plants as well. The presence of oxide was also noted to affect the solar absorptivity of the mixture compared to a clean initial specimen, potentially resulting in a change in the overall efficiency of the CSP plant.

14 SOLAR ENERGY↗

DOE OSTI2021

p-i-n High-Speed Photodiodes for X-RAY and Infrared Imagers Fabricated by In-Situ -Doped APCVD Germanium Homoepitaxy

Design, simulation, fabrication, and measurement have been performed using a new method for obtaining high-speed p-i-n germanium photodiodes for use in backside imaging arrays. These devices are proof-of-principal in situ-doped vertical mesa structures fabricated using single-wafer, atmospheric-pressure chemical vapor deposition (APCVD) homoepitaxy on Czochralski-grown Ge substrates. Excellent impurity profiles resulting from in situ doping, as measured by spreading-resistance profilometry, were 5 × 10 16 , 2 × 10 15 , and 2 × 10 18 (cm -3 ) for the p-type anode, i-region absorber, and n+ cathode, respectively. Scanning electron microscopy analysis revealed very-low defect densities in the epitaxial layers. The diodes, with diameters ranging 100-500 μm, exhibited excellent ideality (n = 1.15) and less than 1-μA dark current, corresponding to up to 80 kV-cm -1 electric field in the smallest devices. Avalanche breakdown bias was consistently at a magnitude of 50 V or lower. External quantum efficiencies in the near-infrared spectrum peaked at 70%. Pulsed illumination experiments, both in the infrared and X-ray regions, showed pulse rise times in the 100-140 ps range, indicating excellent applicability of this method to high-speed X-ray or infrared applications.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Analysis of Dust Samples Collected from an Inland ISFSI Site (''Site B'').

In October of 2020, dust samples were collected from the surface of spent nuclear fuel (SNF) dry storage canisters during an inspection at an inland Independent Spent Fuel Storage Installation, the second inland site at which surface deposits have been sampled. The purpose of the sampling was to assess the composition and abundance of the soluble salts present on the canister surface, information which provides a metric for potential corrosion risks. The samples were delivered to Sandia National laboratories for analysis. At Sandia, the soluble salts were leached from the dust and quantified by ion chromatography. In addition, subsamples of the dust were taken for scanning electron microscopy to determine the texture and mineralogy of the dust and salts. The results of those analyses are presented in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Performance of Pulsed Thermal Tomography Imaging with Machine Learning-Based Classification of Defects in Additively Manufactured Structures

Additive manufacturing (AM) is an emerging method for cost-efficient fabrication of complex topology nuclear reactor parts from high-strength corrosion resistance alloys, such as stainless steel and Inconel. AM of metallic structures for nuclear energy applications is currently based on laser powder bed fusion (LPBF) process. Some of the challenges with using LPBF method for nuclear manufacturing include the possibility of introducing pores into metallic structures. Integrity of AM structures needs to be evaluated nondestructively because material flaws could lead to premature failures in high temperature nuclear reactor environment. Currently, there exist limited capabilities to evaluate actual AM structures non-destructively. Pulsed Thermal Tomography Imaging (PTT) provides a capability for non-destructive evaluation (NDE) of subsurface defects in arbitrary size structures. The PTT method is based on recording material surface temperature transients with infrared (IR) camera following thermal pulse delivered on material surface with flash light. The PTT method has advantages for NDE of actual AM structures because the method involves one-sided non-contact measurements and fast processing of large sample areas captured in one image. Following initial qualification of an AM component for deployment in a nuclear reactor, a PTT system can also be used for in-service nondestructive evaluation (NDE) applications. In this report, we describe recent progress in enhancing PTT capabilities in detecting and visualizing microscopic defects in metallic specimens. The thermal tomography (TT) algorithm obtains depth reconstructions of spatial effusivity from the data cube of sequentially recorded surface temperatures. However, interpretation of TT images is non-trivial because of blurring of images with increasing depth. To address this challenge, we have developed a deep learning convolutional neural network (CNN) to classify size and orientation subsurface defects in simulated TT images. CNN is trained on a database of TT images created for a set of simulated metallic structures with elliptical subsurface voids. Test of CNN performance demonstrate the ability to classify radii and angular orientation of subsurface defects in TT images. In addition, we have shown that CNN trained on elliptical defects is capable of classifying irregular-shaped defects obtained from scanning electron microscopy (SEM) of stainless steel sections printed with LPBF.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

SR19021 Tritium Aging of Regenerated LANA.75 (Final Project Report)

The Savannah River Tritium Enterprise (SRTE) has used the metal hydride LaNi 4.25 Al 0.75 (LANA.75) in the Tritium Facilities for over two decades. LANA.75 beds store significant quantities of tritium but have a limited service life due to the radiolytic decay of tritium to He-3 within the metal matrix. It has been shown that heating tritium-aged LANA.75 under vacuum can reverse tritium aging effects, eliminating the heel of trapped hydrogen, and restoring the reversible capacity. Additional investigation is needed to ensure there are no unexpected changes to the hydride before this restoration technique is employed in full scale beds in the Tritium Facilities. This project was to be comprised of three distinct scopes: obtain tritium aging data on the sample regenerated in 2018, regenerate a tritium-aged sample at 600 °C, and perform thermal stability testing on a non-tritiated sample. Isotherms were collected on the previously regenerated sample after approximately two years of tritium aging. Isotherms were collected at 80, 100, and 120 °C. As expected, there was a decrease in the plateau pressure, an increase in plateau slope, and a portion of the “heel” of tritium trapped in the metal had been reestablished. Unexpectedly, it appeared that the plateau had shortened at the higher tritium to metal ratios as well. This is typically seen in older samples. The second scope, to regenerate a second tritium-aged LANA.75 sample, was not completed. A Task Technical and Quality Assurance Plan was written and approved, a high temperature test cell was fabricated, several pre-job briefs were held, and the hydride sample was passivated with air. Despite these successes, the hydride sample was not recovered from the legacy test cell. The third scope was to perform thermal stability testing on a “cold” LANA.75 sample. A non-tritiated sample of LANA.75 was held at 750 °C under vacuum for 200 hours to simulate exposure to multiple regeneration evolutions. Hydride isotherm performance, chemical composition, crystallinity, particle size, and morphology are compared between the pre- and post-regeneration samples. No significant changes were observed in composition, crystallinity, or particle size. Comparison of before and after isotherms showed that performance improved rather than deteriorated during the evolution. Scanning Electron Microscopy (SEM) analysis showed small growths on the particle surface after exposure to regeneration conditions. Additional testing will be required to determine the cause of these growths.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Production and Performance of High Sulfur Concentration Waste Glasses (FY21)

In FY21 Savannah River National Laboratory (SRNL) was tasked to investigate the practical aspects of low activity waste (LAW) glass production with high sulfur loading for the Office of River Protection (ORP). Relationships between sulfur solubility, retention, and melter resonance need to be evaluated to refine waste loading limits and the super saturated melting (triple melt, grind, and wash) methodology. The impact of high sulfur concentration on durability and phase stability (crystallization and liquid-liquid immiscibility) during slow cooling also needs to be further evaluated. To accomplish this, SRNL proposed scaled melter testing using simulants with matched physicochemical properties to melter feed, as opposed to laboratory reagents. This type of testing is expected to validate applicability to design parameters and functional criteria expected during unit operations. This work was expected to commence in FY21, with the first melter tests being performed near the end of the FY. However, reduced access to Savannah River Site and personnel availability during the COVID-19 pandemic significantly reduced the amount of time and resources available to perform the task. This report documents the experimental plan and progress towards this task. The following is a summarized list of scheduled testing that will be performed on feed, condensate, cold cap and glass: 1) Melter runs; 2) Product consistency test (PCT); 3) Inductively coupled plasma optical emission spectroscopy (ICP-OES); and 4) Optical (OM) and scanning electron microscopy (SEM).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Fiscal Year 2021 Filtration of Hanford Tank 241-AP-107 Supernatant Samples Obtained at Prototypic Tank Level and Filtered at 16 °C

Bench-scale filtration testing of 8.5 liters of supernatant from Hanford waste tank 241-AP-107, chilled to 16 °C, was conducted using a backpulse dead-end filter (BDEF) filtration system equipped with a feed vessel and a Mott inline filter Model 6610 (Media Grade 5) in the hot cells of the Radiochemical Processing Laboratory at Pacific Northwest National Laboratory. This was done to assess the impact of a lower sampling location within the tank as well the lowered filtration temperature on supernatant stability and fouling. The as-received samples were transferred to 1.5-liter poly bottles and held at 16 °C for approximately 1 week prior to filtration. The feed was filtered through the BDEF system at a targeted flux of 0.065 gpm/ft 2 to match the prototypic operation of the TSCR system. During the initial period of filtration, the differential pressure required to effect filtration at 0.065 gpm/ft 2 increased until it reached 2 psid [the Tank Side Cesium Removal (TSCR) action limit] at 26 hours. After this, the filter was backpulsed to dislodge accumulated solids and reduce this pressure differential. An additional eight backpulses were conducted during the initial filtration period; each time, the target pressure was reached sooner than during the previous interval. Volume filtered decreased from 0.9 to 0.03 m 3 /m 2 over the course of 14 processing hours. After the ninth backpulse, the backpulse frequency had become unsustainable, and it was decided to perform an extensive filter cleaning. The filter was cleaned by draining the chilled AP-107 feed into chilled holding containers, introducing 0.1 M NaOH into the feed vessel, and recirculating the NaOH through the system for 20 minutes before allowing the system to soak for 2 hours without temperature control. A measurable decrease in filter resistance during this recirculation indicated that the 0.1 M NaOH was likely dissolving some of the solids that had deposited on the filter. Post cleaning, the filter resistance was effectively restored to initial conditions as the initial transmembrane pressure was restored to original levels. However, resumed processing of the AP-107 feed at 16 °C continued to result in an increased rate of filter resistance. An additional five backpulses were conducted (four during feed processing, one during subsequent cleaning) before the conclusion of the test. Solids concentrated from the backpulse solutions displayed sodium oxalate-type phases, Al-oxides and sodium Al-oxides/carbonates, iron oxides, and Ca-bearing phases (calcite). The scanning electron microscopy analyses also revealed a large distribution of particles, with some particles, notably sodium phosphate dodecahydrate, having diameters close to 0.5 mm.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Development of an Open-source Alloy Selection and Lifetime Assessment Tool for Structural Components in CSP

Lack of sufficient data on high temperature mechanical and corrosion behavior of structural materials is a huge barrier in the technological maturity of current and future Concentrating Solar Power (CSP) technologies. Rapid development and selection of materials cannot be achieved by expensive and time-consuming acquisition of experimental data. The goal of the proposed work is development of an open-source alloy selection and lifetime prediction tool that will integrate validated physics-based models to describe influence of temperature, alloy composition, environment and component geometry (thickness) on mechanical and corrosion behavior of Ni and Fe-based alloys employed in molten salts/sCO 2 heat exchangers. This one-year project leveraged the extensive dataset on the creep\corrosion behavior of candidate materials generated at ORNL through past projects and input from current collaborations with industrial partners. Based on previous experience and the feedback provided by industry (Brayton Energy and Echogen), three candidate materials of interest, Ni-based alloys 740H, 282 and 625 and application-specific operating conditions (max. temperature of 730 °C and stress of 150 MPa) were identified for the heat exchanger. An extensive corrosion and creep dataset was assimilated for the relevant operating conditions and was supported by detailed characterization of about 100 metallographic cross-sections. The corrosion dataset consisted of scanning electron microscopy images (secondary electron and backscatter electron), measured concentration profiles of alloying elements using energy dispersive X-ray spectroscopy (EDS), widths of denuded zones (dissolution of strengthening phases) and depths of attack in molten KCl-MgCl 2 mixtures using image analyses. The creep dataset comprised of creep rupture data and creep strain curves (for 740H and 282). Coupled thermodynamic-kinetic microstructure-based models were employed to predict the stress-corrosion induced compositional and phase evolutions in the alloy during operation under the identified operating conditions. Reduced order models were developed from advanced physics-based models and were integrated in a user-friendly alloy selection tool. The corrosion model was able to predict the time to a critical Cr concentration at the oxide/alloy interface (chemical lifetime) within ±10% (1 standard deviation) of typical statistical variation in corrosion tests and EDS measurement errors (±0.5 wt%). The initial scope of the project was limited to predict creep rupture times (Larson-Miller parameter). Based on the input provided by industry, the mechanical lifetime of the heat exchanger is governed by accumulated creep strains (2%) rather than creep rupture. To be able to predict the times to specific creep strains, a more extensive creep model development was undertaken largely beyond the initial scope of the project. The continuum damage mechanics creep model was able to predict times to 2% creep strain, t 2% with an accuracy of ±500h. Ultimately, a screening protocol for SiC was generated to demonstrate the pathway for integration of one of the currently immature materials from a commercial adoption standpoint in the current material evaluation tool. The modeling tool developed here is accessible to the science community and stakeholders and lays the foundation for methods that will enable a rapid evaluation of optimum materials for CSP applications and reliable prediction of material degradation thereby considerably reducing operational costs, improving reliability and increasing overhaul intervals. However, the complete potential of such a tool to include a wider range of materials and test conditions can only be realized with a more concentrated combined experimental-characterization-computation effort.

14 SOLAR ENERGY↗

DOE OSTI2022

Comprehensive Study of the Potential of Extracting and Processing Critical Minerals from Coal-Based Resources - Phase I

The Phase I report prepared for the Department of Energy addresses U.S. Executive Order 13817 titled A Federal Strategy to Ensure Secure and Reliable Supplies of Critical Minerals, issued on December 20, 2017, that lists 35 critical minerals that are vulnerable to supply disruption. A comprehensive review of each of the minerals was conducted to determine the criticality based primarily on extractability from coal-based resources. Several other factors were also considered such as gaps in supply and demand, use in current technology, and the existence of viable substitutes. It was determined that the critical minerals that show highest potential for extraction from coal-based resources are lithium, rare earth elements (REEs), cobalt, and manganese. All four of the critical minerals listed serve an important role in the technology industry, have few substitutes, and have a heavy import reliance. Most notably are lithium, which is widely used in the electric vehicle industry, and the REEs which can be found in virtually all electronic devices. The research of critical mineral extraction from coal-based resources was completed using a combination of literature review from public sources as well as cooperation from coal mines and power plants across the United States. Samples collected from six different geographical locations across the U.S. were subjected to sample preparation (i.e., pH measurement, moisture content, particle size analysis) and characterization studies using Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Scanning Electron Microscopy, Energy Dispersive X-Ray Spectroscopy (SEM-EDX) instruments. 27 samples of coal waste materials such as refuse, sludge, and fly ash were tested to characterize the rare earth element concentration by total rare earth elements (TREEs), heavy rare earth elements (HREEs), and light rare earth elements (LREEs). Of the 27 samples tested, 22 contained a TREE concentration higher than the threshold of 300 ppm, which is considered a viable feedstock material. 3 samples contained less than 300 ppm of TREEs; however, they were within 20 ppm of the threshold, and could potentially be considered viable sources in the future pending the advancement of more efficient extraction technologies. 2 of the 27 samples had significantly low TREE concentrations, which does not imply any potential for being a source for REEs. For the minerals identified as most critical in the literature review, a conceptual process flow diagram (PFD) was developed for their extraction from different coal-based feedstocks. The process targets selective recovery of one commodity (i.e., rare earths, lithium, cobalt, and manganese) via several hydrometallurgical separation methods. By identifying potentially extractable coal-based critical mineral resources, a study of the current and future market environments for each critical mineral, and a review of current processing methodologies for critical mineral extraction from coal-based resources, the foundation has been laid to further characterize and explore new resources and extraction techniques. As reliance on technologies in industries such as the production of electronic devices, batteries, and alloys containing critical minerals utilizing critical minerals continues to increase, a sound understanding of our nation’s dependence on and even the global criticality of certain critical minerals, will serve as a catalyst for innovation in virtually all fields of science.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2022

Impact Response of Control Atmosphere Plasma Spray Deposited Materials

Thermal spray processing of metals and respective blends is becoming increasingly attractive due to the unique properties such as increased yield strength, low ductility, and differences in tensile and compressive strengths that result from microstructural features due to the spray process compared to other additive manufacturing methods. Here we report the results of plate impact experiments applied to Controlled Atmosphere Plasma Spray deposits of tantalum (Ta), niobium (Nb), and a tantalum-niobium blend (TaNb). These methods allowed for definition of the Hugoniot for each material type and the assessment of the Hugoniot Elastic Limit (HEL). Spallation experiments were conducted, and soft recovery of each material type allowed for scanning electron microscopy to characterize the fracture mechanism during tensile loading.

36 MATERIALS SCIENCE↗