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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 1,063 records · Page 59
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Niowave: Flow measurement development - FY2022

It is critical to maintain proper flow rate of lead bismuth eutectic (LBE) for the operation of the Niowave converter. The high temperature of the LBE (> 200 °C) makes it difficult to purchase simple off the shelf flow meters. One method that can be used to measure the flow rate indirectly, is to use a venturi, with a known throat diameter and differential pressure measurement across the throat. LANL designed the venturi and after numerous iterations with vendors, narrowed down a differential pressure transducer to use with the venturi.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Application of Liquid Metals in Nano Particles and Vitrimers (Summer Quarter Report)

The melting points of pure elemental metals can be decreased by forming liquid metal alloys at the eutectic point composition. Utilizing this unique characteristic, the desirable properties of electron rich metals can be incorporated into new materials and applications. Liquid metals can be made into nanoparticles (LMNP), allowing for increased surface areas of a highly tunable and functional metal oxide skins. Careful characterization is important for understanding LMNP with many methods available to not only investigate size and distribution, but also surface chemistries. This is a review article written to fulfill the UO KCGIP Masters Program requirements.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Thermophysical Property Measurements of NaCl-UCl3

Describe the research (summary of Scope of Work and principal objectives of the CRADA): Argonne produced and measured thermochemical and thermophysical properties of eutectic NaCl-UCl 3 salt at several temperatures in the range of interest. (Note measurements made with UCl 3 instead of PuCl 3 at participant’s request.)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Neptunyl and Uranyl Peroxide Chemistry in Molten Salts, Uranyl Structures from Nature, and Thermodynamic Studies Extending into the Transuranium Elements (Final Report)

This project was initially funded in 2007 and the current technical report is for 2019-2022. The specific objectives of this project (2019-2022) were: (1) Synthesis of actinide peroxide compounds in molten salt eutectics, (2) Studies of new uranyl minerals with fascinating structures and compositions, (3) Drop-solution calorimetric studies of actinide compounds extending into the transuranium elements, with an emphasis on the uranyl-sulfate system, (4) Continuation of studies of uranyl vanadate clusters synthesized using ionic liquids, (5) The training of Ph.D. graduate students in actinide chemistry, and to introduce undergraduate students to actinide research through providing research assistantships and supervision. The report lists 67 archival journal papers describing the work, of which 23 were since 2019. Highlights of the work are summarized for the period 2019-2022.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Conceptual design for a blanket tritium extraction test stand

Tritium breeding is fundamentally required for a sustainable fusion fuel cycle, yet the technological readiness of blanket technology lags far behind other fusion systems. Breeder concepts are divided into solid and liquid media, where solid breeders typically rely on a sweep gas, such as helium, to carry away tritium from lithium containing ceramic materials, and liquid breeders produce tritium from lithium containing eutectics (e.g., PbLi) or molten salts (FLiBe). In each case, tritium must be harvested from the breeding medium. A promising method for tritium extraction is through a vacuum permeator, in which a concentration gradient from the tritium- containing fluid promotes diffusion through a membrane with high hydrogen permeability to the vacuum. This technology has been demonstrated for hydrogen gas systems using Pd and PdAg permeators, but relatively little work has been done to test tritium extraction from PbLi. A Tritium Extraction eXperimental (TEX) loop is being designed to test tritium extraction in a vacuum permeator configuration. The system design is such that it will allow the testing of tritium extraction from both helium and PbLi. A phased approach is being taken that will allow testing of small specimens for fundamental permeation measurements, to multi-meter component testing at near-prototypic conditions. A molten PbLi loop is challenging due to the toxic and explosive nature of Pb and Li, respectively, radiological concerns by introducing tritium, and high temperatures involved in such a system. In addition, PbLi corrosion is a significant issue at high temperatures (>400C). The TEX system will not employ a neutron source for volumetric production of tritium. Herein we present the design and methods used for 1) pumping PbLi, 2) introducing deuterium or tritium into the PbLi, 3) quantifying the amount of deuterium in the loop, 4) extracting deuterium and tritium, and 5) quantifying the total amount of extracted deuterium or tritium from the permeator. In addition, the safety design for operating such a system will be discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

High Energy Aqueous Lithium-Ion Batteries (Final Report)

Lithium-ion batteries (LIBs) are a widely used energy storage device. For decades, researchers have been studying aqueous electrolytes due to their non-flammability and environmental friendless. However, the hydrogen evolution reaction (HER) and oxygen evolution reaction (OER) of water greatly limit the voltage window of the aqueous electrolyte. The recent concept of ‘solvent-in-salt’ electrolytes, namely ‘water-in-salt’ electrolytes (WISE) for aqueous batteries and ‘super-concentrated’ electrolytes for non-aqueous batteries, greatly expands the electrochemical window of these battery systems and consequently their energy density. This expansion occurs because the large number of salt aggregates in ‘solvent-in-salt’ electrolytes enables preferential salt decomposition over solvent decomposition to form robust interphase layers on electrodes, thus enhancing their stability in a wide range of operating voltages. However, even when the salt concentration reached 63m, regardless of cost, the cathodic potential of the WISE expanded only to 1.75V with an overall electrochemical stability window of <3.25V. To design the next generation of high energy aqueous LIBs, electrolytes that possess a wider electrochemical stability window and lower cost but still maintain the merit of non-flammability are much needed. In this project, we designed the aqueous LIBs with widened voltage window. We reported a cheap and non-flammable ternary eutectic electrolyte that can form a robust solid electrolyte interface (SEI)so that the cathodic limiting potential can be decreased to <1.5 V for Li 4 Ti 5 O 12 anodes, while maintaining both a high ionic conductivity of >10 -3 S cm –1 and a low viscosity. We demonstrated the super-stable2.5V LiMn 2 O 4 ||Li 4 Ti 5 O 12 pouch cell with mass loading of 2.5 mAh/cm. Further, LiCl-LiBr-Graphite/Li full cells were designed with high energy density. Finally, we use a bi-layer design to enlarge the voltage window of aqueous electrolyte, which can endow the stable cycling of aqueous Lithium metal batteries.

25 ENERGY STORAGE↗
DOE OSTI2023

Electrorefining Bismuth Using AC Superimposed DC Waveforms

Bismuth has been successfully electrorefined using both direct current (DC) and alternating current (AC) superimposed DC waveforms on a kilogram scale at Brigham Young University (BYU) in Provo, Utah. Finding a suitable surrogate for Pu Electrorefining has always been a technical challenge. This work allows the possibility to improve Pu electrorefining with nonradioactive material at universities, which greatly reduces the cost at Lawrence Livermore National Laboratory (LLNL) for improving Pu Electrorefining. Several possible surrogates (Ce, In, Sn, Zn, and Bi) were selected based on their ability to mimic the Pu electrorefining process. Ce and In electrorefining experiments were conducted at the same temperature as Pu electrorefining in CaCl 2 while Sn, Zn, and Bi, electrorefining experiments were conducted at a lower temperature in a eutectic molten salt composed of LiCl, KCl, and CaCl 2 . The electrorefining experiments using Ce with a Ga impurity did not produce a cathode ring. While the remaining Ce anode after electrorefining was significantly less than what was initially added, Scanning Electron Microscopy with Energy Dispersive X-Ray (SEM-EDX) analysis showed that the Ce had been oxidized and formed a colloid with the surrounding molten salt. This is supported in the literature by analysis of Ce-rich CaCl 2 mixtures. All the electrorefining experiments except for the Ce electrorefining experiments did not introduce an impurity to the system. Electrorefining experiments with In showed that the InCl 3 volatilized out of the molten salt to an extent that only a low current (<<1 A) could be supported without decomposing the molten salt. Zn electrorefining experiments had marginal success, however an easily separable product was not formed. The Zn had to be rinsed from the salt after the experiment. Of the identified potential surrogates. Investigated, only Bi and Sn yielded an easily separable product ring, which enables quantitative analysis of yields and coulombic efficiencies. Bismuth was first identified as a surrogate at BYU, while Sn was identified as a surrogate at LLNL later. Thus, the electrorefining experiments conducted at BYU using AC superimposed DC waveforms used Bi as the metal to be electrorefined.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Low-temperature electroplating of zirconium: Ionic mixture methods

Non-aqueous ionic mixture electroplating allows for room-temperature or near-room-temperature applications of thin metal films, typically metals for which aqueous solvents cannot be used. While there are many methods of plating zirconium, each has disadvantages that ionic liquid plating could correct. However, ionic liquid plating requires further research and development, as a new class of solvents were only developed a few years ago. Current ionic liquids cost several thousand dollars per liter, yet produce films with less than 90% surface coverage as well as deposition of salt species that enable interdiffusion through the Zr. Film coverage needs to be increased while maintaining approximate room temperatures and lower solvent costs. Current ionic liquid technology has suffered from poor surface wetting and limited Zr diffusivity leading to dendrite formation which inhibits total surface coverage. Zr electroplating can be improved by application of room-temperature ionic mixtures as the plating medium. We have successfully demonstrated the capability to plate high-quality metallic zirconium layers using deep eutectic solvents (DES) and DES-ionic liquid (IL) mixtures. Plating thicknesses of up to 6 microns have been achieved, and the coatings are conformal to the substrate and not dendritic, with low contamination from the plating solution. Our current method uses the DES ethaline (a 1:2 ratio of choline chloride and ethylene glycol) to complex zirconium (IV) ions (present via the addition of ZrF4). We pulse plate the Zr from a bath with an overabundance of LiF, in accordance with prior literature. We have also had success in mixing the DES with ILs, such as Triethylsulfonium bis(trifluoromethylsulfonyl)imide, 1-Butyl-3-methylpyridinium bis(trifluormethylsulfonyl)imide, Methyl-trioctylammonium bis(trifluoromethylsulfonyl)imide, and Diethylmethyl(2-methoxyethyl)ammonium bis(trifluoromethylsulfonyl)imide.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Property Measurements of NaCl-UCl 3 and LiF-NaF-KF Molten Salts Doped with Surrogate Fission Products

Knowing thermophysical and thermochemical property values of salts over the expected range of operating temperatures is essential to modeling and simulation efforts that support the commercialization of molten salt reactor (MSR) technologies. Salt properties being measured at Argonne include thermal transitions, phase behavior, heat capacity, density, surface tension, volumetric thermal expansion, thermal diffusivity, thermal conductivity, and viscosity. The properties of several salts of interest including NaCl-UCl 3 and LiF-NaF-KF (FLiNaK) have been measured at Argonne and reported previously. These property measurements are suitable for use in evaluating reactor performance during startup and the early operating life of the reactor. The physical and chemical behavior of the fuel salt is altered by the buildup of fission products and the resulting changes in heat transfer and other properties must be understood by reactor developers. This report summarizes melting behavior, heat capacity, and thermal diffusivity values that were measured for mixtures of FLiNaK and NaCl-UCl 3 salts that have been doped with surrogate fission products. Additional heat capacity measurements were also performed on a separate mixture of FLiNaK without dopants. Measured properties are suitable for incorporation into the Molten Salt Thermal Properties Database–Thermochemical (MSTDB–TC) and the Molten Salt Thermal Properties Database–Thermophysical (MSTDB–TP) to support the development of MSRs. Salts were prepared by doping eutectic mixtures of NaCl-UCl 3 and FLiNaK with surrogate fission products. These salts were prepared for use in salt spill experiments being performed at Argonne and were subsequently used for thermal property measurements. Thermal transitions and heat capacities were measured by using differential scanning calorimetry (DSC) and thermal diffusivity was measured by using laser flash analysis (LFA) at temperatures spanning the typical MSR operating range. The measured property values of the doped salts were compared to measured property values of salts without fission products to quantify the effects of the doped fission products.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Optimal Chloride Salt Mixture for a Fusion Blanket

Deuterium-tritium fusion reactors cannot operate for a significant period without a closed tritium fuel cycle, according to a recent National Academies of Sciences (NAS) report on bringing fusion reactors to the US electrical grid. This fact places breeder blankets as one of the foundational systems for self-sustained fusion reactor operation. The typical functional requirements for a breeder blanket system include producing tritium, absorbing kinetic energy, transporting thermal energy, and being environmentally attractive. State-of-the-art research on liquid blankets has converged to primarily focus on (LiF) 2 and BeF 2 (FLiBe) molten salts and a metallic eutectic of lead and lithium (PbLi), but “virtually all of the technologies related to the tritium fuel cycle are at a low technological readiness level”. This work sought to explore optimum blanket configurations as it aligned with the Oak Ridge National Laboratory (ORNL) FY 2023 Laboratory Directed Research and Development Program’s research priority of developing and expanding the current understanding of fusion blanket science and technology. This purpose of this work was to address ORNL research priorities and NAS recommendations by investigating novel liquid blanket materials that could provide self-sustaining operation and draw on experience from research on molten salts used for advanced fission reactors, concentrated solar, and thermal energy storage. The hypothesis when proposing this research was that there could be chloride-based blanket designs that can exceed the tritium breeding ratios of (FLiBe) molten salt blankets while reducing the use of Be (FLiBe), avoiding the generation of HF (FLiBe), and minimizing magnetohydrodynamic (MHD)-perturbed flow fields (PbLi). The fastest and most cost-effective path to deploying liquid fusion breeder blankets could be from maximizing the synergistic technological overlap between fusion, fission, concentrated solar, and thermal energy storage industries.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Thermal Property Measurements for an LEU-Fueled Molten Salt Reactor (Final CRADA Report)

Argonne performed thermodynamic modeling to predict eutectic compositions in the LiF-NaF-UF4 and LiF-NaF ThF4 systems. Three compositions each of LiF-NaF-UF4 and LiF-NaF-ThF4 molten salts were produced and the thermal properties between 400 and 600 °C were measured. One composition from each salt system was selected for measurements of heat capacity and thermal diffusivity across temperature ranges suggested by FEI experts as beneficial to industry.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Abstract for CRADA between National Energy Technology Laboratory and Western Michigan University

NETL will team up with Western Michigan University (WMU) on designing, developing, and testing the integration of enzyme/Deep Eutectic Solvent (DES) treatment with kraft pulping with a focus on optimizing kraft pulp conditions and testing pulp strength, and optical, and barrier properties. NETL will perform pulping experiments by combining Kraft pulping with DES to allow lower temperature, and reduced chemical use thus improved yield. The main purpose of this study is to investigate different heating methods during chlorine dioxide bleaching of kraft pulps. This study also compares the methods with one another with respect to pulp properties and fiber morphologies. In addition, the energy consumption and quality of bleached pulps under different bleaching conditions are going to be analyzed to provide a solid basis for scale up microwave bleaching process.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Savannah River National Laboratory – General Fusion 2023 INFUSE Report (Rev.1)

This report describes the results from an INFUSE research project, where Savannah River National Laboratory (SRNL) in collaboration with General Fusion (GF) used process modeling to understand and optimize commercial power plant (CPP) fuel cycle designs based on parameters provided by GF. The study primarily focused on two candidate fuel cycles with different blanket materials, one with a lead lithium eutectic (LLE) blanket and the other with a pure lithium (Li) blanket. LLE benefits from a low melting point, favorable neutronics, and lower reactivity, but liquid lithium has the potential for higher tritium breeding ratios (TBR) and does not poison the plasma as a high Z contaminant. It was found that the main differences between LLE and Li designs are the extraction technologies required to remove tritium from the blanket and the amount of tritium and its distribution within the facility. More than 80% of the in-process tritium inventory for the LLE design is contained in the isotope separation system, while for the Li design, over 60% of the in-process tritium inventory is contained within the blanket material. This is due to significant tritium retention by Li. For the Li blanket, the burden of tritium processing rests on the blanket extraction technology rather than the traditional exhaust processing route. Thus, the blanket extraction technology is a main driver of tritium inventory in the Li system and determines the subsequent interface with the tritium processing plant.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Exploring the Nature of f-Element Soft Donor Interactions Using Electronically Tunable Azolate Ionic Liquids

This project was undertaken to advance the understanding of how f-elements interact with moderately soft donors, a heavily investigated yet open question which is of prime importance in spent nuclear fuel processing and fundamental inorganic chemistry. During the course of the project, based on exciting results, a stretch goal was developed to study the hydrolysis products of transuranic actinide metals, a somewhat understudied field even with its significance in nuclear fuel cycle and impacts in environmental chemistry. The stretch goal was to take our serendipitous discovery of an easy route to isolation of crystalline multinuclear ƒ-element hydrolysis products, and apply it to gaining a mechanistic understanding of Pu(III/IV) colloid formation. The simplicity of our techniques should lend themselves to the remote handling required for study of many transuranic elements. We developed several methodologies using azolium azolate chemistry to overcome ƒ-element hydrolysis problems that prohibit the isolation of ƒ-element soft donor complexes and to build a crystallographic library of ƒ-element N-donor complexes as a means to understand the fundamental differences between actinide and lanthanide interactions with moderately soft donor ligands. Our next major endeavor will be to transfer this chemistry from 4ƒ elements to transuranic elements, particularly in the study of hydrolysis of Pu(III/IV). While our work is fundamental in nature, applications of the knowledge we are generating should be felt in such diverse fields as catalysis, separations in general, nuclear waste remediation specifically, and many other applications in f-element magnetic and luminescent properties. The potential ramifications of the consistent and robust formation of hydrolysis controlled hexanuclear lanthanide structures are enormous, with future uses being catalyst formation, higher-nuclearity structure synthesis using our hexanuclear motif as a fundamental building block, Pu waste remediation, separations, and many other potential applications resulting from characteristic magnetic and luminescent properties of lanthanide polynuclear structures. Three synthetic methodologies (direct mixing with variable stoichiometries, use of volatile solvent, metathesis) were developed starting with 7 acidic and 6 basic azoles to obtain ionic liquids suitable for f-element coordination. Proton transfer by acidic/basic azole combination led to suitable low melting salts and two cocrystals. Acid/base reaction of azoles with soft-donor permanent cations of ([X 4444 ][OH] (where [X 4444 ] + = tetrabutylammonium [N 4444 ] + or tetrabutylphosphonium [P 4444 ] + ) with weakly acidic azoles including imidazole, 1,2,3-triazole, 1,2,4-triazole, 5-aminotetrazole, 4,5-dicyanoimidazole, and 2-amino-4,5-dicyanoimidazole) revealed several suitable low-melting salts. Metathesis reactions of Na(azolate) were conducted by first using weakly acidic azoles including 4,5-dicyanoimidazole, 2-amino-4,5-dicyanoimidazole, 5-aminotetrazole, and 1,2,4-triazole to form sodium or lithium salts using group(I) hydroxides in methanolic solutions. The best results were obtained by reacting the basic and acidic azoles directly in 1:1 or 3:1 ratios at elevated temperatures. Twenty-two azole mixtures were identified which are either low melting solids or room temperature liquids. Each of the low melting solids was confirmed to be a new solid phase, each of which is being further characterized. The liquids and solids are anticipated to be ILs, eutectics, or partially ionized systems, all of which will be suitable for the dissolution of f-element salts. Five new synthetic methodologies were developed to finding suitable crystallization conditions for f-element complexation with the goal of finding simple one pot reaction syntheses and crystallization strategies that could be used under the demanding conditions of transuranic chemistries. These synthetic methods yield many new crystalline phases which were studied by single crystal X-ray diffraction.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

ECAR-6588 Chemical Compatibility of MARVEL Components: Project # 33526

This ECAR serves to demonstrate the compatibility of MARVEL Reactor materials in contact with the two molten metal coolants, sodium potassium eutectic and lead. Operational requirements are determined, and recommendations for material treatment or replacement are presented.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

SAM Code Enhancements for Modeling of Liquid Metal-Cooled Fast Reactor Concepts

The SAM code is under development and supported by DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. These advanced reactor concepts incorporate novel and improved approaches to achieve safety and economic feasibility. This report summarizes two major efforts in addressing the modeling gaps in SAM for liquid-metal-cooled fast reactor (LMFR) applications, i.e. thermal mixing and stratification phenomena in large pools and corrosion-oxidation of components in flowing lead. A new one-dimensional model for thermal mixing and stratification effects in large pools and enclosures is developed and implemented. Thermal mixing and stratification occur when fluid enters a pool at a temperature different than the bulk fluid itself, a scenario often encountered during transients in pool-type systems. These phenomena are critical for the safety of reactors, impacting phenomena like natural circulation, which is essential for passive cooling. The improved model in SAM addresses limitations of state-of-the-art approaches by combining one-dimensional (1D) channels, representing the coolant jet flow, with lumped-parameter zero-dimensional (0D) pools, representing the rest of coolant in the tank. Energy exchange between the 1D jet and the 0D pools is based on heat transfer correlations calibrated against 3D simulations. It is verified that this model can handle various flow configurations, including hot jets in colder pools, cold jets in hotter pools, and the presence of features like ceilings, free surfaces, and obstacles. Additionally, validation against experimental data demonstrates the ability of the model to capture mixing and stratification effects in a wide range of conditions. The flexibility and improved accuracy of the new model make it a valuable tool for reactor safety analysis, allowing for the simulation of different geometries encountered in advanced reactors. A system-level corrosion modeling capability is developed and implemented in SAM to support Lead Fast Reactor (LFR) development. Although the initial focus of this capability will be on LFR application, this can later be expanded to include other liquid metals such as Lead-Bismuth Eutectic (LBE) and PbLi. This report summarizes the common corrosion mitigation strategies and outlines the progress on implementing and validating a corrosion-oxidation model in SAM. Verification and validation of the corrosion-oxidation portion of the model was performed using analytical solution and measured data from samples tested in the non-isothermal pumped lead loop at IPPE Obninsk. The iron transport and corrosion/precipitation portion of the model was assessed using an analytical model and measured corrosion depths from a natural convection lead loop experiment performed at CEA. It is demonstrated that the model implemented in SAM performed well in these assessments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Cesium Removal from Surrogate Pyroprocessing Salt by Electrodeposition

Active metals in used nuclear fuel dissolve into the salt during pyroprocessing and are not removed by electrorefining or drawdown operations. The buildup of 137 Cs over time increases the heat load and ionizing radiation level of the salt such that it must be replaced frequently, resulting in a significant amount of salt waste. An effective means of managing cesium in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. A previous report summarized issues that must be addressed when developing a removal strategy and assessed the suitability of existing methods and remaining technological gaps to their application (Rose and Thomas 2023). Cesium is extremely stable in molten salt as a chloride—even more stable than the LiCl-KCl eutectic base salt used for pyroprocessing fuel—which makes removing cesium a challenge. However, sufficiently strong atomic interactions occur between active metal species and liquid metals that make the electrodeposition of active metal fission products into liquid metal electrodes energetically favorable. The feasibility of recovering cesium from LiCl-KCl pyroprocessing salt through electrodeposition into liquid metals is eing assessed by identifying potentially effective liquid metals and performing tests to determine the effectiveness of electrodepositing cesium from a LiCl/KCl salt into these liquid metals. Previous studies investigating the electrodeposition of Sr 2+ , and Ba 2+ into zinc, cadmium, bismuth, lead, tin and antimony have shown that alkali and alkaline earth metals can be electrodeposited at liquid metal cathodes (Kim et al., 2018). The removal of Ba 2+ and Sr 2+ was measured to be more efficient than the removal of monovalent cations due to the greater thermochemical driving force for alloying those elements with the liquid metal (Jang et al. 2022). Because the equilibrium potentials are dependent on the interactions of the active metal in the liquid metal, it is likely that the other alkali metals Li + , and K + , will deposit from LiCl/KCl salt with the Cs + . Therefore, application of this method to recover active metals from pyroprocessing salt will benefit from the use of a liquid metal and set of operating conditions that sufficiently increase the reduction potential of cesium to remove cesium from the waste salt with an acceptable amount of co-deposited lithium and potassium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Translation of "Contribution a l'etude du diagramme pression-temperature du beryllium pur" (Contribution to the Study of the Pressure-Temperature Diagram of Pure Beryllium)

An unusual feature of beryllium’s phase diagram is the transition from a hexagonal close packed (hcp, or α) to a body-centered cubic (bcc, or β) structure ~20 K below the normal melting temperature of T melt ≈ 1560 K. An early thermal analysis of 99.7% pure Be found two arrests1 : one at 1533-1536 K and another at 1551-1557 K. Without ruling out an allotropic transformation in Be itself, Sloman attributed the first of these to solidification of a Be-BeO eutectic. Losano observed only the higher of the two in 99.962% Be, seeming to confirm the notion that the lower feature was an artifact of impurity. Teitel (99.4%) also measured two arrests and attributed the lower to impurity, but questioned whether oxygen was the culprit.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗