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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Voltammetry measurements in lithium chloride-lithium oxide (LiCl–Li 2 O) salt: An evaluation of working electrode materials

Instrumentation to provide process monitoring (PM) and safeguards for the oxide reduction (OR) step in the electrochemical processing of used oxide nuclear fuel is necessary to ensure equipment is operated as declared. Cyclic voltammetry (CV) has been proposed for real-time monitoring of the operation of an OR process for safeguards purposes. In this study, different materials including 316 stainless steel (SS), tantalum (Ta), molybdenum (Mo), tungsten (W), platinum (Pt), and iridium (Ir) were evaluated as potential working electrode (WE) materials based on their chemical inertness, corrosion resistance, and sensitivity in detecting Li 2 O and other chloride salts in the OR electrolyte. Of the electrodes evaluated, 316 SS, Ir, and Pt all performed reasonably well in the LiCl-Li 2 O electrolyte. Here, stainless steel was operated in the cathodic potential and had reasonable corrosion resistance and is relatively inexpensive. Iridium could be operated in both the cathodic and anodic potentials and was the most corrosion resistant of those evaluated. Platinum is limited to the anodic potential range, had reasonable corrosion resistance, and was the most sensitive to Li 2 O concentrations in the salt. In the development of a stand-alone safeguards instrumentation, our recommendation is a CV sensor fitted with both Ir and Pt WEs. The results in this study may also be helpful for anode material selection in other electrometallurgy industries, such as molten salt electrolysis.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Theoretical Design of Lithium Chloride Superionic Conductors for All-Solid-State High-Voltage Lithium-Ion Batteries

The development of solid electrolytes (SEs) is a promising pathway to improve the energy density and safety of conventional Li ion batteries. Several lithium chloride SEs, Li 3 MCl 6 (M = Y, Er, In, and Sc), have gained popularity due to their high ionic conductivity, wide electrochemical window, and good chemical stability. This study systematically investigated 17 Li 3 MCl 6 SEs to identify novel and promising lithium chloride SEs. Calculation results revealed that 12 Li 3 MCl 6 (M = Bi, Dy, Er, Ho, In, Lu, Sc, Sm, Tb, TI, Tm, and Y) were stable phase with a wide electrochemical stability window and excellent chemical stability against cathode materials and moisture. Li-ion transport properties were examined using bond valence site energy (BVSE) and ab initio molecular dynamics (AIMD) calculation. Li 3 MCl 6 showed the lower migration energy barrier in monoclinic structures, while orthorhombic and trigonal structures exhibited higher energy barriers due to the sluggish diffusion along the two-dimensional path based on the BVSE model. AIMD results confirmed the slower ion migration along the 2D path, exhibiting lower ionic diffusivity and higher activation energy in orthorhombic and trigonal structures. For the further increase of ionic conductivity in monoclinic structures, Li-ion vacancy was formed by the substitution of M 3+ with Zr 4+ . Zr-substituted phase (Li 2.5 M 0.5 Zr 0.5 Cl 6 , M = In, Sc) exhibited up to a fourfold increase in ionic conductivity. This finding suggested that the optimization of Li vacancy in the Li 3 MCl 6 SEs could lead to superionic Li 3 MCl 6 SEs.

25 ENERGY STORAGE↗
DOE OSTI2023

Fluorine-Substituted Lithium Chloride Solid Electrolytes for High-Voltage All-Solid-State Lithium-Ion Batteries

Lithium ternary halides are promising solid electrolytes, owing to their high ionic conductivity and reasonably high oxidative and chemical stability. Recently, fluorine substitution in Li 3 MCl 6 has been suggested as a promising approach for further enhancing oxidation stability. Accordingly, this study outlines a material design strategy for F-substituted Li 3 MCl 6 through systematic theoretical analyses. Calculations reveal that the mixing limit of F in Li 3 MCl 6–x F x is in the range of 0.5–1.5, and the resulting Li 3 MCl 6–x F x phases can retain ionic conductivity above 1 mS/cm up to x = 1.0. Additionally, the calculations also predict that the formation of F-containing passivating phases could increase the oxidation potential for Li3MCl5F to ~6.3 V. The proposed material design strategy is validated through the synthesis of Li 3 YCl 5 F, which is confirmed to show both high ionic conductivity and enhanced oxidation stability. The design guidelines presented herein can accelerate the potential use of halide-based electrolyte chemistries in high-voltage all-solid-state batteries.

25 ENERGY STORAGE↗
DOE OSTI2023

Crystal structures of polymerized lithium chloride and dimethyl sulfoxide in the form of {2LiCl·3DMSO} n and {LiCl·DMSO} n

Two novel LiCl·DMSO polymer structures were created by combining dry LiCl salt with dimethyl sulfoxide (DMSO), namely, catena-poly[[chlorido­lithium(I)]-μ-(dimethyl sulfoxide)-κ 2 O:O-[chlorido­lithium(I)]-di-μ-(dimethyl sulfoxide)-κ 4 O:O], [Li 2 Cl 2 (C 2 H 6 OS) 3 ]n, and catena-poly[lithium(I)-μ-chlorido-μ-(dimethyl sulfoxide)-κ 2 O:O], [LiCl(C 2 H 6 OS)]n. The initial synthesized phase had very small block-shaped crystals (<0.08 mm) with monoclinic symmetry and a 2 LiCl: 3 DMSO ratio. As the solution evaporated, a second phase formed with a plate-shaped crystal morphology. After about 20 minutes, large (>0.20 mm) octa­hedron-shaped crystals formed. The plate crystals and the octa­hedron crystals are the same tetra­gonal structure with a 1 LiCl: 1 DMSO ratio. These structures are reported and compared to other known LiCl·solvent compounds.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Direct Processing of Lithium Chloride Based Waste Salt in a Ceramic Waste Form

Ceramic waste form materials were fabricated with a LiCl based salt mixture to demonstrate the suitability of direct processing of waste salt from an all LiCl pyroprocessing flow sheet on the formation, microstructure, and durability of ceramic waste forms. Materials were successfully synthesized by using the simplified direct processing technique with salt loadings of 5, 7.5, and 10 wt% at the laboratory scale. Materials fabricated with LiCl based salt indicated full conversion of zeolite to sodalite occurred to form products with low open porosity. This indicates that the Na 2 O content of NBS4 glass is suitable for processing salt chemistries which do not contain NaCl. Generated sodalite was microencapsulated by the glass binder, halite occlusions were detected within sodalite domains at all salt loadings, and a Cs-rich phase believed to be Cs-pollucite was detected in the material made with the lowest salt loading (5 wt%). Salt inclusions were encapsulated within the binder glass at all loadings, and found to consist primarily of NaCl with Cs present. A small amount of salt was found at the surface of the wasteform made with the highest salt loading of 10 wt%, which may indicate an upper limit to the amount of salt that can be accommodated or incomplete mechanical mixing of the reagents. Durability testing by using the ASTM C1308 method indicated that initial rapid dissolution of exposed halite inclusion phases was directly correlated to salt loading. Wasteform degradation during all tests occurred primarily due to dissolution of the sodalite phase. Durability of the binder glass was found to increase as the waste salt loading increased. Results for the release of cations from the salt indicate the changes in wasteform durability and halite inclusion chemistry are strongly influenced by ion exchange phenomena between the salt and the glass. This appears to be the result of the relatively stronger affinity for the glass phase of Li, compared to that of other cations such as Na and Cs. The effects from cation exchange were diminished at lower salt loadings.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

A Molten Salt Electrochemical Process for the Preparation of Cost-Effective p-Block (Coating) Materials

Solar energy applications rely heavily on p-block elements and transition metals. Silicon is, by far, the most commonly used material in photovoltaic cells and accounts for about 85% of modules sold presently. Of late, thin film photovoltaic cells have gained momentum because of their higher efficiencies. Most of these thin film devices are made out of just five elements, namely, cadmium, tellurium, selenium, indium, gallium and copper. The present manuscript describes an elegant and inexpensive molten salt-based electrolytic process for fabricating a tellurium-coated metallic substrate. A three-electrode set up was employed to coat iridium with tellurium from a molten bath containing lithium chloride, lithium oxide and tellurium tetrachloride (LiCl-Li 2 O-TeCl 4 ) at 650 °C for a duration ranging from 30 to 120 min under a galvanostatic mode. The tellurium coating was observed to be thick, uniform, smooth and homogeneous. Additionally, the deposited tellurium did not chemically react with the iridium substrate to form intermetallic compounds, which is a good feature from the standpoint of the device’s performance characteristics. The present process, being generic in nature, shows the potential for the manufacture of both the coated substates and high-purity elements not just for tellurium but also for other p-block elements.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Removal and Deactivation of Bond Sodium from Fast Reactor Blanket Materials

The disposition of sodium-bonded spent nuclear fuel and blanket materials in a repository is complicated by the presence of sodium metal that is used as a thermal bond between the uranium metal fuel and blanket slugs and their cladding. The concern is that the metallic sodium could react with water, producing explosive hydrogen gas, or could exhibit a pyrophoric character. Thus, experimental studies were performed to investigate and demonstrate the removal and deactivation of bond sodium from blanket material in a dry environment. Specifically, bond sodium was removed from unirradiated Fermi-1 blanket elements and an assembly via a melt-drain-evaporate process using elevated temperature and reduced pressure. The effectiveness of sodium metal removal from the blanket materials and their associated cladding was =99.9998%, based on post-test quantitative analyses. The separated sodium metal was collected and subsequently deactivated by reacting it in a molten state with a controlled addition of ammonium chloride particles atop a molten salt medium. In this process sodium chloride is formed and assimilates into the salt pool. The subsequent deactivation of the bond sodium produced a solid ingot of sodium chloride, potassium chloride, lithium chloride, and cesium chloride that was devoid of sodium metal based on post-test analyses of the salt product. Both the sodium removal and deactivation operations were conducted within a dry inert atmosphere enclosure. The results of this study substantiate a path forward for the disposition of sodium-bonded blanket materials, including 34 metric tons heavy metal in irradiated Fermi-1 blanket material currently stored at Idaho National Laboratory.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Synthesis and characterization of uranium trichloride in alkali-metal chloride media

Given a growing interest in uranium salts for pyrochemical processing of used fuel and uranium-fueled molten salt reactors, the synthesis of uranium trichloride in alkali-metal chloride media was investigated in a series of four experiments. Specifically, uranium metal powder and uranium hydride powder were prepared and separately blended with ammonium chloride and lithium chloride – potassium chloride eutectic in two runs, while the same powders were separately blended with ammonium chloride and sodium chloride in two additional runs. Each of the lithium chloride – potassium chloride containing blends was slowly heated to 923 K, while those containing sodium chloride were heated to 1123 K. During each heat up, the ammonium chloride sublimed into gaseous ammonia and hydrogen chloride, leading to the chlorination of uranium metal or uranium hydride and the formation of molten salt solutions of the respective chlorides. Experimental conditions were incorporated in the runs to promote formation of uranium trichloride over uranium tetrachloride in the respective media. Molten samples of each run product were taken and characterized via chemical analyses, diffractometry, and microscopy. The final products from each run were dark dense ingots of the respective salt systems with uranium concentrations ranging from 44 to 51 wt%. Finally, chemical analyses and diffractometry identified the predominant presence of uranium trichloride in these systems; however, a possible minor presence of uranium tetrachloride could not be conclusively dismissed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Synthesis and characterization of uranium trichloride in alkali-metal chloride media

Given a growing interest in uranium salts for pyrochemical processing of used fuel and uranium-fueled molten salt reactors, the synthesis of uranium trichloride in alkali-metal chloride media was investigated in a series of four experiments. Specifically, uranium metal powder and uranium hydride powder were prepared and separately blended with ammonium chloride and lithium chloride – potassium chloride eutectic in two runs, while the same powders were separately blended with ammonium chloride and sodium chloride in two additional runs. Each of the lithium chloride – potassium chloride containing blends was slowly heated to 923 K, while those containing sodium chloride were heated to 1123 K. During each heat up, the ammonium chloride sublimed into gaseous ammonia and hydrogen chloride, leading to the chlorination of uranium metal or uranium hydride and the formation of molten salt solutions of the respective chlorides. Experimental conditions were incorporated in the runs to promote formation of uranium trichloride over uranium tetrachloride in the respective media. Molten samples of each run product were taken and characterized via chemical analyses, diffractometry, and microscopy. The final products from each run were dark dense ingots of the respective salt systems with uranium concentrations ranging from 44 to 51 wt%. Chemical analyses and diffractometry identified the predominant presence of uranium trichloride in these systems; however, a possible minor presence of uranium tetrachloride could not be conclusively dismissed.

D Herrmann, Steven↗
DOE OSTI2024

Anionic Effects on Concentrated Aqueous Lithium Ion Dynamics

In this article, the structural dynamics, chemical reactivity, anisotropy, diffusivity, viscosity, and density were measured for concentrated lithium salt solutions, including lithium chloride (LiCl), lithium bromide (LiBr), lithium nitrite (LiNO 2 ), and lithium nitrate (LiNO 3 ), with methyl thiocyanate as an infrared vibrational probe molecule, using two-dimensional infrared spectroscopy (2D IR), nuclear magnetic resonance (NMR) spectroscopy, and viscometry. 2D IR, NMR, and viscosity results show that LiNO 2 exhibits longer correlation times, lower diffusivity, and nearly four times greater viscosity when compared to the other lithium salt solutions of the same concentration, suggesting that nitrite anions may strongly facilitate structure formation via strengthening water-ion interactions, directly impacting bulk solution properties at sufficiently high concentrations. Additionally, the LiNO 2 and LiNO 3 solutions show significantly reduced chemical reactivity with respect to lithium cations coordinating with the methyl thiocyanate when compared to the lithium halide salts.

2D IR↗
DOE OSTI2022

Different interfacial reactivity of lithium metal chloride electrolytes with high voltage cathodes determines solid-state battery performance

A deep understanding of the interaction of the surface of cathode materials with solid electrolytes is crucial to design advanced solid-state batteries (SSBs). This is especially true for the new class of lithium metal chloride (Li-M-Cl) solid electrolytes which are receiving rapidly growing attention due to their very high oxidative stability (>4 V) in combination with good ionic conductivity that can enable long cell cycle life. While Li-M-Cl electrolytes typically contain resource-limited metals (M) such as indium or rare earths, work has focused on substituting M with more abundant elements such as zirconium. Via operando resistance measurements using intermittent current interruption we explore the dynamic evolution of the interphase at the surface of Ni-rich NCM85 or NCM111 cathode particles inside a working SSB with three different Li-(M 1 ,M 2 )-Cl catholytes (Li 3 InCl 6 , Li 2 Sc 1/3 In 1/3 Cl 4 and Li 5/2 Y 1/2 Zr 1/2 Cl 6 ) to reveal the impact of the cationic metal substitution on the interfacial chemistry. We show that the metal plays a critical role in determining high voltage stability, contrary to prior assumptions. Using a combination of cyclic voltammetry and ultraviolet photoelectron spectroscopy measurements of the electronic band structure to assess oxidative stability; coupled with DFT calculations and ToF-SIMS to evaluate products formed at the interface at different upper cutoff potentials and degrees of delithiation, we are able to differentiate between electrochemical and chemical degradation. We find that Li 2 Sc 1/3 In 1/3 Cl 4 yields the highest (and Li 3 InCl 6 the lowest) stability against electrochemical oxidation, while Li 5/2 Y 1/2 Zr 1/2 Cl 6 undergoes a detrimental chemical reaction with oxygen released from Ni-rich NCM85 at high potentials, resulting in fast capacity fading. Altogether, our work establishes a platform for the metrics and an approach that can be utilized to efficiently evaluate the stability of new halide SEs in SSB cells.

25 ENERGY STORAGE↗
DOE OSTI2021

Halogenation of used aluminum matrix test reactor fuel – a bench-scale demonstration with surrogate materials

In this work, experiments with surrogate materials were performed at bench scale to demonstrate a halogenation technique applicable to treatment of used aluminum matrix test reactor fuel. The technique involves dissolution and separation of aluminum from used aluminum matrix test reactor fuel in molten-halide salt systems prior to treatment and disposition of the fuel’s uranium and fission products. Demonstration of the halogenation technique was performed with neodymium metal as a non-radiological surrogate for uranium metal. Experiments involved blending forms of aluminum and neodymium metal with ammonium and lithium chloride or ammonium and lithium bromide, which upon heating decomposed into ammonia gas and the respective hydrogen chloride or bromide gas. The latter reacted with the metals to form the respective aluminum and neodymium halides. At elevated temperatures, aluminum halides gasified away from the respective neodymium halides, which fused with their respective lithium halides. Samples of fused and distillate salts were collected and analyzed, yielding extents of aluminum removal that ranged from 94.5–98.2% for chlorination runs and 91.4–97.8% for bromination runs. No neodymium was detected in the distillate fractions. Some experiments were repeated with excess reactants, and a portion of aluminum chloride distillate was processed into a consolidated waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Computing chemical potentials with machine-learning-accelerated simulations to accurately predict thermodynamic properties of molten salts

The successful design and deployment of next-generation nuclear technologies heavily rely on thermodynamic data for relevant molten salt systems. However, the lack of accurate force fields and efficient methods has limited the quality of thermodynamic predictions from atomistic simulations. Here we propose an efficient free energy framework for computing chemical potentials, which is the central free energy quantity behind many thermodynamic properties. We accelerate our simulations without sacrificing accuracy by using machine learning interatomic potentials trained on density functional theory (DFT) data. Using lithium chloride as our model system, we compute chemical potentials with DFT-accuracy for solid and liquid phases by transmuting ions into noninteracting particles. Notably, in the liquid phase, we demonstrate consistency whether we transmute one ion pair or the entire system into ideal gas particles. By locating the temperature where the chemical potential of solid and liquid phases cross, we predict a melting point of 880 ± 18 K for lithium chloride, which is remarkably close to the experimental value of 883 K. With this successful demonstration, we lay the foundation for high-throughput thermodynamic predictions of many properties that can be derived from the chemical potentials of the minority and majority components in molten salts.

Gibson, Luke D. [Oak Ridge National Laboratory (OR↗
DOE OSTI2020

Fabrication and Characterization of Composite Membranes for the Concentration of Lithium Containing Solutions Using Forward Osmosis

In this study, an energy and highly time efficient, cost effective, and scalable forward osmosis (FO) composite polyvinylidene fluoride (PVDF) membrane is developed to concentrate lithium chloride recovered from naturally occurring geothermal brine resources. Dense thin robust defect-free Kynar-PVDF membranes are fabricated on porous PVDF hollow fiber supports. Various membrane fabrication parameters such as membrane thickness, coating solution concentration, cross-linker concentration, and curing temperature are studied to develop a defect-free dense membrane. Advanced characterization tools such as FIB-SEM, Raman, FTIR spectroscopy, and XRD are used to study the surface and cross-sectional morphology, chemistry of crosslinking, crystallinity and chemical structure of the dense PVDF membrane. FO is utilized to concentrate LiCl in the solution from 36 to 175 g L -1 with an average water transfer rate of 0.450 L m -2 h -1 (LMH) at 85 °C. Furthermore, the membrane showed stable performance for more than 550 h while maintaining the purity of the concentrated lithium chloride solution of >99.99 wt%. Overall, the results suggest that the FO-based technology is an innovative, time and energy efficient approach compared to the conventional technologies for concentration of battery grade LiCl for application in lithium ion batteries and other clean energy technologies.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

4.3.1.1 TCF: Recovery of Lithium from Geothermal Brine with Lithium-Aluminum Layered Double Hydroxide Chloride Sorbents

Lithium has gained importance as a near-critical element for various energy storage applications, and efforts are ongoing to develop simple and cost-effective methods for lithium recovery. For instance, in a typical 50 MW geothermal power plant in the Salton Sea, United States, approximately 15,000 metric tons of Li 2 CO 3 or LiOH salts can be produced annually by recovering and converting LiCl from geothermal plant waste solutions. Currently, approximately 390 MW of geothermal power is generated in the Salton Sea Known Geothermal Resource Area, which corresponds to potential Li 2 CO 3 production of approximately 120,000 metric tons per year. This makes lithium recovery from geothermal brines a promising, potentially low-cost source of lithium, contributing significantly to the global supply chain.

15 GEOTHERMAL ENERGY↗
DOE OSTI2025

Demonstration of Electroreduction Technology to Convert GNF Uranium Oxide Powder to Metal

Global Nuclear Fuel – Americas LLC (GNF) has partnered with Argonne National Laboratory to demonstrate electroreduction of uranium oxides produced by deconversion of UF 6 to uranium metal through the Gateway for Accelerated Innovation in Nuclear (GAIN) program under the U.S. Department of Energy to accelerate the domestic production of metallic advanced reactor fuels. Electroreduction of uranium oxide was first demonstrated and patented by Argonne in the early 2000s as a technology to convert used oxide nuclear fuel from light water reactors to metal for further fuel reprocessing. More recently, electroreduction has been proposed as a front-end technology for metallization of uranium oxides produced by deconversion of UF 6 and for scrap recovery of oxide materials. During the electroreduction process, UO 2 powder is contained in a stainless-steel mesh basket with a cathode lead located in the center of the UO 2 bed. The basket is immersed in lithium chloride molten salt electrolyte containing 1 wt% lithium oxide along with a platinum anode and a nickel/nickel oxide (Ni/NiO) reference electrode. Current is applied to the cell between the cathode and anode to reduce the UO 2 to metallic uranium via a solid-state reduction reaction. Oxide ions released from the UO 2 during reduction are transported through the salt to the anode where oxygen gas is evolved. Once reduction is complete, the basket containing the metallicized uranium is removed from the salt and can be processed to remove the salt and consolidate the uranium into an ingot for use in metallic fuel fabrication. This project was performed to provide evidence of the electroreduction technology readiness level for metallization of UO 2 powder, identify and retire technical risks for industrialization of electroreduction, and accelerate the path to commercialization for metallic fast reactor fuel production. To that end, five electroreduction tests were performed with UO 2 provided by GNF and the resulting product was analyzed for the extent of conversion to metal and for impurity contents of the metal product to verify that electroreduction does not introduce impurities that would prevent use of the product in metallic fuel fabrication.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗