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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 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Surface grain refinement of casting A380 aluminum alloy by ultrasonic-assisted surface mechanical grinding treatment

A novel surface processing technique, ultrasonic-assisted surface mechnical grinding treatment (U-SMGT), was proposed and applied to modify the surface micriostructure of the casting aluminum alloy (A380). Comprehensive post-process chracterization techniques including 3D profilometry, optical microscope, scanning electric microscopy (SEM), and electron backscatter diffraction (EBSD) were carried out to investigate the surface morphology and microstructure of the processed region (PR). It was found that a surface layer of grain refinement with a thickness of about 50 µm was formed. In the as-received base material (BM), large clusters of dendritic eutectic phase with a size of approximately 20 µm were isolated by α-Al matrix. Nano-indentation tests showed that hardness of the eutectic and α-Al in the as-received BM was 2.15 GPa and 1.2 GPa respectively. After U-SMGT, the large dendritic eutectic phase was broken down into much fine particles with an average size of 0.57 µm. These fine particles were uniformly distributed into the α-Al matrix. Nano-indentation measurement exhibited a hardness of 2.27 GPa within the processed region. The overall high hardness in the processed region is a result of the pinning effect of uniformly distributed fine dendrites and grain refinement of Al-matrix.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Friction stir processing on a strontium modified, thin-wall, vacuum-assisted high-pressure die-cast Aural-5 alloy to improve tensile and fatigue performance

Here, this study explores the application of friction stir processing (FSP) to enhance the material properties of Sr-modified Aural-5 alloy, with a focus on improved tensile and fatigue properties. Aural-5 is a well-known vacuum-assisted high-pressure die-cast (HPDC) Al-Si7-Mg alloy used in the automotive industry to reduce vehicle weight, enhance fuel efficiency, and lower carbon emissions. This alloy modifies its material chemistry with Sr for fine fibrous networks of eutectic silicon and manganese (Mn) to reduce die soldering. It has significantly less iron (Fe) content resulting in the elimination of detrimental needle-shaped Fe-bearing ß-phase intermetallic and improving ductility. The initial microstructure of as-received HPDC Aural-5 exhibits shrinkage porosity in the middle section, a dendritic microstructure with fibrous Al-Si eutectic colonies, a shear-band structure beneath the die-wall, large dendritic externally solidified crystals (ESCs), needle-shaped Mg 2 Si phase and significant second-phase particulates. Some of those microstructural features, such as porosity, ESCs, needle-shaped Mg 2 Si phase, and large second-phase particles, serve as initiation sites for cracks under mechanical loading, resulting in adverse effects on tensile properties, particularly ductility. FSP effectively transforms the microstructure into a wrought configuration with uniform particle distribution by eliminating porosity and disintegrating dendrites, eutectic colonies, ESCs, second-phase particles, and shear-band structures. FSP-driven microstructure modification enhances yield strength and tensile ductility by ~30% and ~35%, respectively. The fatigue life of the material in a bending mode configuration (stress ratio R = 0.1) after FSP exhibits enhancements ranging from 2.0 to 3.9 times that of the original HPDC Aural-5 alloy, depending on the applied stress level.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Volatilized Molten Salts: An Alternative Avenue for Synthesizing Single-Phase Perovskites

Single phase La 0.8 Sr 0.2 MnO 3 (LSM) and core-shell La 0.6 Sr 0.4 Co 0.2 Fe 0.8 O 3 (LSCF)-LSM have been synthesized using a volatilized mol-ten salt synthesis (vMSS) method for the first time with a LiCl-KCl eutectic. While deleterious to LSM formation when the MSS takes place in the liquid phase, LiCl-KCl eutectic successfully facilitates LSM formation when volatilized. Specifically, KCl evaporates before LiCl and promotes formation of LSM via the gaseous phase. As time progresses, LiCl volatilizes and contributes negatively to Sr retention in the perovskite phase in accordance with Lux-Flood chemistry and to product phase purity. The vMSS is therefore a way to obviate the more immediate restrictions of Lux-Flood chemistry in the liquid phase. LiCl-KCl eutectic can also be used to successfully synthesis core-shell LSCF-LSM nanoparticles in times as short as 1 hour at 600 °C. These results demonstrate the surprising versatility and flexibility of the MSS method to synthesize numerous potential energy relevant materials with greater ease than previously thought.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Phase Discovery and Selected Synthesis of Subvalent Niobium Tellurides Using a Polytelluride Flux Strategy

Transition metal subchalcogenides involve electron-rich metals and can facilitate an in-depth understanding of the relationships among quantum properties such as superconductivity, charge density wave, and topological band structures. However, effective experimental routes toward synthesizing transition metal subchalcogenides are still lacking, hindering the development of new quantum materials. Herein, we propose a eutectic polytelluride flux strategy as an excellent solution to address phase discovery and crystal growth in transition metal subtelluride systems. We report new phases easily and selectively synthesized using a eutectic “K 3 Te 4 ” polytelluride flux upon adjusting the ratio of Nb metal to flux in the starting materials (K/Nb/Te = 3:x:4). Using a high Nb content in the solvent (x = 2 and 1), crystals of KNb 3 Te 3 O 0.38 and K 0.9 Nb 3 Te 4 are obtained. Both subtellurides exhibit diverse Nb clusters, including face-sharing and edge-sharing Nb 6 octahedral columns and zig-zag Nb chains. Reducing the Nb content to x = 0.33 leads to the formation of a layered compound, K 1.06 NbTe 2 . This compound comprises a NbTe 6 trigonal prism with K intercalated between the layers. Single crystals of known binary Nb tellurides can also be grown using another eutectic flux “KTe 3 . 2 ”, and the obtained NbTe 2 exhibits a new polymorphism with extra trimerization along the b-axis in the Nb–Nb bonded double zig-zag cluster. Finally, precise control over the structural dimensionality and oxidation state, combined with the facile crystal growth process, makes our synthetic strategy an efficient route to explore quantum materials in transition metal subchalcogenides.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Enhanced Formation of Solvent-Shared Ion Pairs in Aqueous Calcium Perchlorate Solution toward Saturated Concentration or Deep Supercooling Temperature and Its Effects on the Water Structure

As a candidate of Martian salts, calcium perchlorate [Ca(ClO 4 ) 2 ] has the potential to stabilize liquid water on the Martian surface because of its hygroscopicity and low freezing temperature when forming aqueous solution. These two properties of electrolytes in general have been suggested to result from the specific cation–anion–water interaction (ion pairing) that interrupts the structure of solvent water. To investigate how this concentration-dependent and temperature-dependent ion pairing process in aqueous Ca(ClO 4 ) 2 solution leads to its high hygroscopic property and the extreme low eutectic temperature, we have conducted two sets of experiments. First, the effects of concentration on aqueous calcium perchlorate from 3 to 7.86 m on ion pairing were investigated using Raman spectroscopy. Deconvolution of the Raman symmetric stretching band (ν 1 ) of ClO 4 – showed the enhanced formation of solvent-shared ion pairs upon increasing salt concentration at room temperature. We have confirmed that the low tendency of forming contact ion pairs in concentrated solution contributes to the high hygroscopicity of the salt. Second, the near eutectic samples were studied as a function of temperature by both combined differential scanning calorimetry–Raman spectroscopic experiments and in situ X-ray diffraction. The number of solvent-shared ion pairs was found to increase with decreasing temperature when cooled below the temperature of maximum density of the solution, driven by a change in water toward an ice-like structure in the supercooled regime. The massive presence of solvent-shared ion pairs in turn limits the development of the long-range order in the tetrahedral networks of water molecules, which is responsible for the extremely low eutectic point and deep supercooling effects observed in the Ca(ClO 4 ) 2 –H 2 O system.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

High-Pressure Melting Experiments of Fe 3 C and a Thermodynamic Model of Fe-C Liquids for the Earth's Core

Melting experiments of Fe 3 C were conducted to 85 GPa in laser-heated diamond anvil cells with in situ X-ray diffraction and post-experiment textural observation. From the determined pressure-temperature conditions of the melting curve for Fe 3 C, together with literature data on the melting point of diamond and eutectic point of the system Fe-Fe 3 C/Fe 7 C 3 under high pressures, we established a self-consistent thermodynamic model for high-pressure melting of the system Fe-C including the mixing parameters for liquids. The results show that mixing of Fe and C liquids is negatively nonideal from 1 bar to the pressure at the center of the Earth. The departure from ideal mixing becomes progressively larger with increasing pressure, which leads to greatly stabilized liquids under core pressures. The modeled carbon content in eutectic melts under core pressures is 3.3–4.4 wt%. From the Gibbs free energy, we derived an internally consistent parameters for Fe-C outer cores which included the crystallizing points at their bottoms, isentropic thermal profiles, and densities and longitudinal seismic wave speeds (Vp). While the addition of carbon in excess of the eutectic melt composition effectively reduces the density of iron liquid, the Vp of iron liquid is not greatly changed. Therefore, the low density and high Vp of PREM relative to pure iron cannot be reconciled by an Fe-C liquid. Therefore, the Earth's core cannot be approximated by the system Fe-C and should include another light element.

58 GEOSCIENCES↗
DOE OSTI2021

Production and Chemical Analysis of NaCl-KCl-UCl 3 Salts

Thermochemical and thermophysical properties of several molten salt systems of interest to molten salt reactor (MSR) developers are being measured at Argonne National Laboratory. These properties are needed by developers to design, license, and operate their reactors. Thermochemical properties being measured at Argonne include thermal transitions, phase behavior, and heat capacity. Thermophysical properties being measured at Argonne include liquid density, surface tension, volumetric thermal expansion of the liquid phase, thermal diffusivity, thermal conductivity, and viscosity. Properties of the ternary NaCl-KCl-UCl 3 system that may be used as a fuel bearing salt are being measured. Low-melting eutectic salt compositions are of interest to developers owing to the thermal efficiency. Using salts with low melting temperatures allow reactors to be operated at a lower temperature to improve operation and decrease safety concerns. Confirming low melting behaviors are maintained during composition changes that occur during fission is also an important factor. It is essential to any molten salt reactor developer utilizing a fuel bearing salt to have a complete understanding of the behavior of that fuel salt at a variety of compositions near the eutectic composition they plan to use. This kind of stability is indicated on a phase diagram by low liquidus slopes at compositions bordering the eutectic. Therefore, a thermodynamic model was constructed to determine what compositions of the ternary NaCl, KCl, and UCl 3 system should be studied to support MSR developers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 1

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore fuel product. The long-term EnCore ® fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Westinghouse Accident Tolerant Fuel Phase 2B with Higher Enriched and Higher Burnup Add-On Project Final Technical Report Deliverable Volume 2

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program funded by Westinghouse and the US Department of Energy (DOE) utilized chromium (Cr) coated zirconium alloy cladding with doped UO 2 (ADOPT(TM)) and uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. ADOPT is a Cr 2 O 3 +Al 2 O 3 doped UO 2 pellet with increased oxidation resistance and density, and increased resistance to fission gas release. Due to the issues with U 3 Si 2 fuel reaction in pressurized water reactor (PWR) environments, uranium nitride (U 15 N) has been substituted for U 3 Si 2 as a long-term fuel option. Cr coated cladding with ADOPT fuel is the near-term Westinghouse EnCore ® fuel product. The long-term EnCore fuel product is SiGA ® SiC/SiC composite cladding with high density/high thermal conductivity UN fuel. In 2020, the higher burnup and higher enriched and (HBHE) program was integrated into the ATF program. The objective of this expanded program is to extend ATF burnups to at least 75 MWd/kgU to economically facilitate 24-month cycles in PWRs. The ATF program now includes ADOPT fuel with enrichments >5% 235 U. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in Westinghouse Churchill autoclaves and the Massachusetts Institute of Technology (MIT) reactor. Additionally, the Cr coated cladding has been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and in LTR programs at the Doel 4 and Bryon 2 commercial reactors. High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried out to determine the time and temperature limits for the Cr coated zirconium claddings. These tests indicate that Cr coated Zr cladding can take temperatures up to about 1500°C for short periods of time without becoming totally oxidized. The main issue is the formation of the Cr-Zr eutectic at 1333°C resulting in the migration of this eutectic inward with the formation of ZrCr 2 which rapidly oxidizes to ZrO 2 on the outside of the tube. The Cr coated Zr cladding also delays the bursting of the tube, reduces the burst area, and reduces hydriding and embrittlement of the Zr which is a major benefit for reducing fuel fragmentation, redistribution, and dispersal (FFRD), the major licensing issue faced by HBHE. The manufacturing parameters for the SiC have been found to have a significant effect on the corrosion rate of the SiC in light water reactor (LWR) conditions and significant improvements have been made in the yield, quality, and oxidation resistance of the SiC cladding by identifying and tightening manufacturing process parameters. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year can be achieved which meets corrosion requirements under normal operating conditions. High temperature steam oxidation tests at the state-of-the-art Westinghouse facilities in Churchill, PA and at Karlsruhe Institute of Technology (KIT) have been carried and determined that SiC cladding can withstand temperatures up to about 1800°C to 1900°C before excessive corrosion reactions begin, and that the SiC will not balloon and burst. ADOPT fuel pellets were incorporated into the Westinghouse ATF program because of their increased density and oxidation resistance. A topical report supporting the use of ADOPT pellets has been submitted to the NRC and will be approved shortly. ADOPT pellets have been a product in Europe for over 15 years. The additives work to make larger grain sizes which appear to reduce fission gas release during transients as well as increase the density of the pellet. Fuel rod and assembly design in preparation for the lead test rod (LTR) and lead test assembly (LTA) programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that both Cr coated cladding and SiC composite cladding can increase the time to fuel rod loss of geometry for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher oxidation resistance of the SiC and Cr coated cladding. These two hours can be used to implement additional response options instituted through the Diverse and Flexible Mitigation Capability (FLEX) program by reactor operators. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely if the modest water flow provided by FLEX continues. Experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Methods for mitigating or stopping the Cr-Zr eutectic are also needed to further improve the oxidation resistance of Cr coated Zr. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 MWd/kgU were completed but post irradiation examinations were not done since U 3 Si 2 has been replaced by UN. UN has three issues that are being addressed. The first is increasing the oxidation resistance so that there is not excessive reaction up to ~1500°C. While UN increases the pellet density and additives to the UN postponed the temperature at which rapid oxidation occurred by ~100°C to 200°C, this is not enough, and pellet coating options are now being pursued. The second is developing a method to manufacture that does not require the multi-step process of UF 6 to UO 2 to UC to UN. Efforts are underway looking at UF 6 to UN 2 to UN and UF 6 to UF 4 to UN 2 to UN reactions. Finally, an economically acceptable process for potentially enriching the 15 N content of the nitrogen used to make UN to >95% 15 N was identified though not experimentally validated. In addition to the work supported by the DOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan, and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings were organized in 2018 and 2019 by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. No meetings were held in 2020 and 2021 due to the COVID-19 related restrictions on traveling and gatherings. Since SiC, coated cladding, and high-density fuel options are not currently used in the nuclear industry; support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/UO 2 fuel. Several meetings have been held with the NRC to generate a fast-track approach to licensing ATF using a combination of atomic modeling, in-rod sensors, and in-reactor testing. This approach was memorialized in an “Accelerated Fuel Qualification White Paper” prepared by the Accelerated Fuel Qualification Working Group lead by General Atomics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Lower-length-scale modeling of chemical additions, corrosion and fission products in select MSR base salts

Molten salts play a crucial role in advancing Generation IV nuclear energy technology, with chloride-based salts like NaCl-UCl 3 garnering significant attention due to their distinctive properties. However, the corrosive nature of molten salts causes the dissolution of chromium (Cr), leading to the formation of CrCl 2 and CrCl 3 species in molten chlorides. Moreover, the radioactive decay of nuclear fuel gives rise to fission products, including Cs, Sr, and I chlorides. The first part of this report presents a comprehensive study utilizing ab initio molecular dynamics (AIMD) simulations to investigate the properties of eutectic NaCl-UCl 3 molten salt in the presence of corrosion products (CrCl 2 and CrCl 3 ) and fission product (CsCl). The study focuses on essential structural and thermophysical properties such as density, mixing energy, coordination numbers (CN), and Radial Distribution Functions (RDF) of the salts with varying compositions of corrosion products and fission product, covering a range from 0% to 13.5%. The results offer valuable insights into the behavior of corrosion and fission products in uranium-based molten salts, providing essential data that can be used as input to the MTDB-TC and MSTDB-TP property databases being developed by the NEAMS program. Due to their favorable characteristics such as low melting points, high boiling points, and low costs, MgCl 2 +NaCl+KCl (MNK) eutectic salts have recently attracted attention as high-temperature heat transfer fluids. The incorporation of LiCl into MNK salts can further reduce their melting points and increase their specific heat capacities, which is de- sirable for high-temperature heat transfer applications. The second part of this report presents the development and validation of a new shell-model potential for the MgCl 2 +NaCl+KCl+LiCl system, which captures the polarization of Cl anions. The extensive comparison with experimental data and AIMD simulations demonstrates the accuracy and reliability of the potential. Furthermore, using this potential, we elucidate the intricate network structure in MNK eutectic salts that is formed through polyhedron sharing. This research contributes to a fundamental understanding of the atomic structures and thermophysical properties of multi-component molten salts, which is crucial for future development of heat transfer fluids for applications in molten salt reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Investigations into the Ternary NaF-KF-UF4 Salt System – Phase A

A knowledge gap exists in the data and understanding of fresh fuel salt and irradiated multicomponent fuel salt systems thermophysical properties. Quantifying these properties is necessary for the design and construction of test reactors, as well as the licensing of future commercial molten-salt reactors. To facilitate thermal property determination on a proposed fuel salt composition for Seaborg Technologies, several samples containing depleted uranium tetrafluoride (UF4), sodium fluoride (NaF), and potassium fluoride (KF) were blended, and a melt temperature analysis was performed. From the melting temperature analysis, it was determined that sample Seaborg-7, a ternary salt composition of 26.4UF4-24.7KF-48.9NaF (mol%), was very near a ternary eutectic point. Therefore, thermal properties such as melting temperature, salt stability, density, heat capacity, thermal diffusivity, and viscosity were experimentally determined on the Seaborg-7 salt. These measurements document the baseline properties of fresh fuel salt as a function of temperature, where future experiments on irradiated fuel salt will provide a holistic perspective on the change of thermophysical properties during reactor operations. Several precision instruments were used to collect property data, and instrument calibrations and data collection were performed and documented in a standardized and reproducible manner with meticulous detail. This process ensured that the measurement procedures and resulting data can readily be duplicated elsewhere. The Seaborg-7 salt was shown to be stable at temperatures up to 900°C, as no mass change was observed upon repeated heating and cooling. The peak melting temperature was determined to be 547°C (557°C endset). The enthalpy of fusion (??H?_fus^o) was determined to be 167.5 ± 2.7 J/g while the enthalpy of crystallization (??H?_c^o) was determined to be -147.8 ± 13.3 J/g. In addition to the eutectic melting peak, upon heating, several pre eutectic peaks were observed, occurring at 470°C (onset) and 499°C (peak). Specific heat capacity measurements showed a slightly increasing trend with respect to temperature in the solid phase, while the liquid-specific heat capacity showed a somewhat flat trend with an average value of 106.1 ± 1.24 J/mol·K between 600 to 800°C. Three independent trials using the Seaborg-7 salt determined the density to be ?(T) = 4.908 – 0.000363·T(°C), validated between 32 to 200°C, and ?(T) = 4.808 – 0.00113·T(°C), validated between ~575 to 850°C. Thermal diffusivity was determined for the liquid state and is represented by the linear equation y = 0.1581 + 0.000207·T(°C) between 550 to 850°C. The viscosity was determined from 600 to 800°C and is represented by the exponential fit equation, ? (mPa·s) = 736.58e^(-0.006·T(°C)). This report documents the conclusion of fuel salt thermophysical property measurements for the Seaborg SPP, Phase A project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Property Measurements of LiF-NaF-KF Molten Salts Doped with Corrosion Products and Oxygen

Measurements of thermophysical properties of molten salts are needed for modeling and simulation activities that support the development of molten salt reactor (MSR) technologies. Properties of interest including transition temperatures, phase behavior, heat capacity, density, volumetric thermal expansion, surface tension, viscosity, thermal diffusivity, thermal conductivity, and vapor pressure are being performed at Argonne. Results of these property measurements are suitable for use in evaluating reactor performance during startup and the early operating life of the reactor. Ingressions of oxygen and moisture into the fuel salt are expected to occur at different times during the operating life of the reactor due to system leaks, maintenance, and refueling activities. The presence of these environmental contaminants induces corrosion of structural materials. The introduction of corrosion-derived species, oxygen and moisture is expected to affect the physical and chemical properties of the salt and operation of the reactor. Previous work performed at Argonne evaluated the effects of fission product dopants on the thermal properties of eutectic LiF-NaF-KF (FLiNaK). Properties of FLiNaK are commonly used to represent those of fluoride-bearing fuel salts. Metallic corrosion products such as chromium and nickel ions together with dissolved oxygen are expected to affect system redox differently than the accumulation of fission products. Work summarized in this report was performed to measure the effects of corrosion products and dissolved oxygen on the phase transitions, and specific heat capacity of FLiNaK. Thermophysical property measurements were made using four salts that were prepared by doping aliquots of a eutectic mixture of FLiNaK with surrogate corrosion products. Controlled additions of CrF 3 and NiF 2 were used as surrogates for corrosion product contamination anticipated during extended reactor operations in which fuel salt is in contact with steel reactor components. Controlled additions of UF 4 from two sources containing known amounts of UO 2 at different concentrations were used to represent oxygen contamination. The phase transitions and specific heat capacities of the four salts were measured by using differential scanning calorimetry. Measurements were made at temperatures spanning the range of 500−900 °C, which is the expected operating range of MSRs. Measured property values were compared to values measured previously with eutectic FLiNaK without dopants. Differences between property values measured for the doped and non-doped salts were compared with the uncertainties of the measurements to determine the significance of the effect of corrosion products and oxygen on salt properties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

In Situ Prediction of Microstructure and Mechanical Properties in Laser-Remelted Al-Si Alloys: Towards Enhanced Additive Manufacturing

Laser surface remelting of aluminum alloys has emerged as a promising technique to enhance mechanical properties through refined microstructures. This process involves rapid cooling rates ranging from 10 3 to 10 8 °C/s, which increase solid solubility within aluminum alloys, shifting their eutectic composition to a larger value of silicon content. Consequently, the resulting microstructure combines a strengthened aluminum matrix with silicon fibers. This study focuses on the laser scanning of Al-Si aluminum alloy to reduce the size of aluminum matrix spacings and transform fibrous silicon particles from micrometer to nanometer dimensions. Analysis revealed that the eutectic structure contained 17.55% silicon by weight, surpassing the equilibrium eutectic composition of 12.6% silicon. Microstructure dimensions within the molten zones, termed ‘melt pools’, were extensively examined using Scanning Electron Microscopy (SEM) at intervals of approximately 20 μm from the surface. A notable increase in hardness, exceeding 50% compared to the base plate, was observed in the melt pool regions. Thus, it is exemplified that laser surface remelting introduces a novel strengthening mechanism in the alloy. Moreover, this study develops an in situ method for predicting melt pool properties and dimensions. A predictive model is proposed, correlating energy density and spectral signals emitted during laser remelting with mechanical properties and melt pool dimensions. This method significantly reduces characterization time from days to seconds, offering a streamlined approach for future studies in additive manufacturing.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Synthesis and Thermophysical Property Determination of NaCl-PuCl3 Salts

Currently, a knowledge gap exists in the available data and understanding of thermophysical properties relating to fresh fuel salts, especially those containing plutonium. Determination of this data is necessary for the design and construction of test reactors, as well as the licensing of future commercial molten salt reactors. Thermophysical properties such as melting temperature, salt stability, density, and heat capacity were determined on synthesized eutectic NaCl-PuCl3 and a more sodium rich composition containing 25 mol% PuCl3. These measurements document the baseline properties of the salt as a function of temperature for future experiments on irradiated fuel salt which will provide a holistic perspective on the change of thermophysical properties during reactor operations. It was determined that the NaCl-PuCl3 ingot synthesized for this study contained 63.4 mol% NaCl, 36.3 mol% PuCl3, and was 99.7% pure. Upon heating it was shown that the NaCl- PuCl3 eutectic was stable at temperatures of 800°C. The onset of melting occurred at 541°C and the enthalpy of fusion was determined to be 140.7 ± 8.4 J/g. Specific heat capacity measurements showed a slightly decreasing trend with respect to temperature in the liquid phase ranging from 0.67 to 0.57 J/g·K, with an average value of 0.637 ± 0.03 J/g·K (104 ± 5 J/mol·K) between 500 to 720°C. Three independent trials of the molten NaCl-PuCl3 eutectic salt found the density to be ?(T) = 3.8589 – 9.5342·10-4 T(°C), validated between 500 to 800°C. In addition to salt synthesis and experimentally determining thermodynamic properties, Ab Initio molecular dynamic (AIMD) simulations were used to calculate density and heat capacity values.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of Heat Treatment on Microstructure and Mechanical Properties of Stellite 12 Fabricated by Laser Additive Manufacturing

Microstructure evolution and wear resistance of Stellite 12 produced by laser additive manufacturing were studied in untreated and heat-treated conditions. In untreated conditions, cobalt matrix dendrites and inter-dendritic eutectic carbides were the primary microstructure, blocky and lamellar eutectic carbides can be found, and M{sub 7}C{sub 3} and M{sub 23}C{sub 6} are the primary carbides. Heat treatment can modify the microstructure of this alloy. Primary carbides were dissolved into the cobalt matrix after solution treatment. Aging treatment resulted in the precipitation of carbides, the transformation of M{sub 7}C{sub 3} into M{sub 23}C{sub 6} carbides, and the increase in microhardness. Solution and aging treatment can also cause the precipitation of carbides, but the morphology and distribution of carbides were absolutely different. The wear resistance of untreated samples at room temperature (RT) and 600 °C is better than that at 200 and 400 °C due to the high hardness at RT and the formation of oxide film at 600 °C. The wear resistance of heat-treated samples was tested at 600 °C; it was dominated by the formation and flake-off of oxide film. The sample that underwent solution plus aging treatment has a superior wear resistance.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Soluble Graphene Nanosheets from Recycled Graphite of Spent Lithium Ion Batteries

Soluble graphene nanosheets are fabricated from recycled graphite of spent lithium ion batteries through a modified Hammers process followed by deoxygenation with NaOH–KOH eutectic. Ultrasonic exfoliation in N-methyl-pyrrolidone indicates the loosened graphene layers in recycled graphite are prone to exfoliation. Reduction of the exfoliated graphene oxide sheets was conducted in molten NaOH–KOH eutectic at different temperatures. The results show that molten NaOH–KOH effectively eliminates the unsaturated oxygen-containing moieties from the exfoliated graphene oxide sheets while creating more hydroxyl functional groups. Higher temperature treatment is more prone to remove hydroxyls while producing the shrinkage on the surface of graphene sheets. Graphene sheet with a good solubility is produced when the graphene oxide is heat-treated at 220 °C for 10 h. After reduction, the graphene oxide sheets exhibit excellent dispersibility or solubility in water, ethanol and other polar solvents, therefore being highly desirable for solution processing of graphene materials. Such study not only identifies a high-quality stockpile to prepare soluble graphene but also paves a feasible alternative of graphite recycling from spent lithium batteries.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Fiber Laser Welding–Brazing Characteristics of Dissimilar Metals AZ31B Mg Alloys to Copper with Mg-Based Filler

Fiber laser welding–brazing of 1-mm-thick AZ31B Mg alloys to 1.5-mm-thick copper (T2) with Mg-based filler was performed in a lap configuration. The weld appearance, interfacial microstructure and mechanical properties were investigated with different heat inputs. The results indicated that processing windows for optimizing appropriate welding parameters were relatively narrow in this case. Visually acceptable joints with certain strength were achieved at appropriate welding parameters. The maximum tensile-shear fracture load of laser-welded-brazed Mg/Cu joint could reach 1730 N at the laser power of 1200 W, representing 64.1% joint efficiency relative to AZ31Mg base metal. The eutectic structure (α-Mg + Mg{sub 2}Cu) and Mg-Cu intermetallic compound was observed at the Mg/Cu interface, and Mg-Al-Cu ternary intermetallic compound were identified between intermetallics and eutectic structure at high heat input. All the joints fractured at the Mg-Cu interface. However, the fracture mode was found to differ. For laser power of 1200 W, the surface was characterized by tearing edge, while that with poor joint strength was almost dominated by smooth surface or flat tear pattern.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Development of Graphene Nanoplatelet-Reinforced AZ91 Magnesium Alloy by Solidification Processing

It is a challenging task to effectively incorporate graphene nanoplatelets (GNPs) which have recently emerged as potential reinforcement for strengthening metals into magnesium-based matrices by conventional solidification processes due to their large surface areas and poor wettability. A solidification processing which combines mechanical stirring and ultrasonic dispersion of reinforcements in liquid matrix was employed to develop AZ91 magnesium alloy matrix composites reinforced with 0.25 and 0.5 wt.% GNPs. The microstructural studies conducted with scanning and transmission electron microscopes revealed that fairly uniform distribution and dispersion of GNPs through the matrix were achieved due to effective combination of mechanical and ultrasonic stirring. The GNPs embedded into the magnesium matrix led to significant enhancement in the hardness, tensile strength and ductility of the composites compared to those of unreinforced AZ91 alloy. The strength enhancement was predominantly attributed to the grain refinement by the GNP addition and dislocation generation strengthening due to the coefficient of thermal expansion mismatch between the matrix and reinforcement. The improved ductility was attributed to the refinement of β eutectics by transforming from lamellar to the divorced eutectics due to the GNP additions. In addition, the strengthening efficiency of the composite with 0.25 wt.% GNP was found to be higher than those of the composite with 0.5 wt.% GNP as the agglomeration tendency of GNPs is increased with increasing GNP content. These results were compared with those of the GNP-reinforced magnesium composites reported in the literature, indicating the potential of the process introduced in this study in terms of fabricating light and high-performance metal matrix composites.

36 MATERIALS SCIENCE↗