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

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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

CHALLENGES IN THE DEVELOPMENT OF THE ELECTROREFINING PROCESS AT Y-12

In order to ensure future capabilities of the Y-12 site as older buildings retire, a new process was developed, the Metal Purification Process. The purpose of this new process is to provide a simpler and more efficient uranium metal purification that the previous multi-staged, complex chemical processes. The basis for this technology currently exists in the US DOE complex, but it has not been utilized on a large scale for uranium at higher enrichments. The system requires larger-geometry vessels and furnaces in order to meet the through-put requirements. The crux of the process is the electrorefining cell, consisting of several concentric liners and an unfavorable geometry cylindrical crucible. Within the crucible, a molten Li-CL-KCL-UCL3 salt electrolyte is present. ‘Dirty’ metal is placed into a loading basket and loaded into an electrorefining cell. Here, the metal reacts into the molten salt within the cell to form UCL3, while ‘clean’ metal simultaneously plates out from the salt electrolyte in the form of metal dendrites. These dendrites are then collected and removed from the cell. After removal from the cell, the dendrites are taken to a furnace to remove adhered salts. Then, the salt-free dendrites are moved to a separate furnace for consolidation into the final product form. The development of the Criticality Safety Evaluation for this complex system brought with it many NCS Challenges and lessons learned. These challenges include: design decisions for ensuring subcriticalty in the electrorefining cell during abnormal conditions, the interface of the main glovebox system with auxiliary systems such as the purification system and designated storage, implementation of mass tracking, and concerns from production and operation regarding the movement of material within the system.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Electrochemical Separation of Ag 2 S and Cu 2 S from Molten Sulfide Electrolyte

The production of precious metals from Cu-rich sources such as ore products or secondary sources is slow and complex largely due to limited solubility in aqueous electrolytes. This results in sequential processing with various electrolytes and chemistries, where first Cu is electrorefined, followed by Ag, followed by Au and the platinum group metals. These are separate processes, often conducted in separate electrorefining and electrowinning facilities. The chemical properties of molten sulfides, and their ability to operate at a temperature where liquid metal cathodes are used, suggest the possibility of an alternative, streamlined processing route for Cu and precious metals. Unfortunately, little thermodynamic or electrochemical information is available regarding the behavior of Cu and precious metal sulfides in molten sulfide electrolytes. Herein, the relative activity of the Cu 2 S-Ag 2 S pseudobinary dissolved in a BaS-La 2 S 3 supporting electrolyte is measured at 1523 K. It was found that the supporting electrolyte favors mixing with Ag 2 S over Cu 2 S. Molten sulfide electrolysis of Cu and Ag was conducted, with results in good agreement with the thermodynamic model. It is found that the Ag-Cu cathode chemistry will influence the electrochemical selectivity in the Ag-Cu-Ba-La-S system.

25 ENERGY STORAGE↗
DOE OSTI2024

Electrolysis in Chloride Molten Salts for Sustainable Critical Metals Production and Recovery

Critical materials, such as rare-earth metals, are essential to numerous applications, including clean energy; however, the present industrial practices for producing rare-earth metals involve environmentally damaging and thus unsustainable chemical and electrochemical processes. An alternative moderate-temperature chloride-based molten salt electrolysis process can address these issues, providing energy efficient and sustainable metal production. While it is being developed presently for rare-earth electrowinning, one can easily envision its broader application to rare-earth electrorefining and the electrolytic production of high-volume metals like Fe and Al. Presently, these high-volume metals industries account for nearly 10% of global greenhouse gas emissions. Furthermore, the chloride MSE process presents a huge opportunity for truly achieving sustainability if it is developed further for producing Fe, Al, Ti, Mg, and other commodity metals.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Management of Alkali and Alkaline Earth Fission Products in Used Pyroprocessing Salt

Pyroprocessing of spent nuclear fuel (SNF) involves dissolving metallic fuel into a molten salt electrolyte (typically eutectic LiCl-KCl) and then preferentially depositing actinides onto inert cathodes. Subsequent operations include drawdown of residual actinides and lanthanides from the electrolyte prior to re-using the salt. The recovered actinides are recycled and the recovered lanthanides are disposed as waste. Alkali and alkaline earth metal fission products in the fuel, such as Cs, Sr and Ba, dissolve into the salt during electrorefining. The concentrations of these elements buildup over time in the molten salt electrolyte, which may change the freezing point. The radioactive decay of 137 Cs and 90 Sr (half life 30 and 29 years) generates significant heat and produces strong ionizing radiation fields (β and γ). The increasing heat load and radioactivity as these elements build up in the molten salt requires frequent replacement and disposal of the electrolyte salt. Alternatively, the salt can be treated to remove these and other elements and then recycled to the electrorefiner. An effective strategy to manage these alkali and alkaline earth metal fission products in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. Alkali and alkaline earth metal fission products are extremely stable in molten salt as chlorides--even more stable than the LiCl-KCl eutectic base salt--making them challenging to remove. They are not removed during drawdown operations to recover residual actinides and lanthanides and a separate operation is required to sufficiently purify the salt for reuse. This work is focused on selecting a method for separating Cs, Sr and Ba from the salt recovered from the lanthanide drawdown operation prior to recycling the cleaned salt back to the electrorefiner. Not addressed in this work is the management of the waste stream produced by the separation. This report summarizes the issues to be addressed when developing removal strategies for cesium, strontium, and barium and reviews existing methods to identify suitable methods and any technological gaps in their application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Mk-IV Salt Crystallization Hot Finger Apparatus for Partitioning Used Electrorefiner Salt

Electrorefining is a controlled redox process used to regulate the behavior of ionic species. Through this process, metals can be deposited onto a cathode from an electrolyte solution in a controlled manner. The Mk-IV electrorefiner (Mk-IV ER) at Idaho National Laboratory is an engineering-scale, molten salt-based electrorefining cell that has been used for decades to recover metallic uranium from spent fuel. As a result, highly stable fission product chlorides have accumulated in the electrolyte. This accumulation results in changes to the salt’s properties, such as melting temperature, thermal conductivity, and density, as well as elevated product impurity and fissile materials criticality margin. These factors prompt the need for a salt regeneration process, such as melt-crystallization and species drawdown. This work focuses on providing a conceptual design to regenerate ER salt from used Mk-IV-ER salt in-situ, while minimizing salt waste volumes by concentrating the fission products in a final processed salt heal. We propose using a hot-finger crystallization apparatus design to fractionally crystallize salt in the Mk-IV-ER head space (or baffle space), allowing the collection of solid and liquid fractions. By using a cup-drain design, the used salt will be allowed to slowly solidify on the walls of a stainless-steel cup. The apparatus drain plug will then open to allow the liquid salt phase to drain to a lower cup, effectively separating the liquid phase from the solid phase. Under the hypothesis that the liquid phase salt concentrates the fission products, which is under examination in the accompanying work package, this separation allows the recovered solid salt to be reused while minimizing the high-level salt waste volume of used ER salt.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Electrorefiner Speciation and Phase Model for Prediction of Operation Lifetime

The Mk IV electrorefiner (ER) at Idaho National Laboratory has recovered uranium metal from used nuclear fuels for 27 years. Thus, large quantities of fission product chlorides more electropositive than uranium have accumulated in the ER salt, increasing the complexity of phase equilibria and salt speciation. Importantly, the melt temperature of the salt increases with ER operation and will ultimately exceed design limits without intervention. Methods to remove fission products from molten chloride salt are known, but their implementation at scale is not yet demonstrated. In this work, historical ER composition and differential scanning calorimetry (DSC) are utilized to develop coupled speciation and chloride salt solution thermochemistry models, which enable accurate present-day analysis of ER salt as well as forecast its future disposition. The developed thermochemical models are applicable to chloride salts generally and have been made available in the Molten Salt Thermal Properties Database – Thermochemical.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Fabrication and Evaluation of Large Alumina Crucibles by Vat Photopolymerization Additive Manufacturing for High-Temperature Actinide Chemistry

Additive manufacturing (AM) offers opportunities to advance the design and function of ceramic tooling in high temperature actinide pyrochemistry. In technical ceramics such as alumina, conventional forming techniques often restrict design flexibility and can limit experimental progress. In this study, we investigate the use of vat photopolymerization (VP) with commercial resins to fabricate large-scale alumina crucibles, reaching dimensions up to 125 mm, which is significantly larger than typically reported for dense VP ceramics. Notably, these additively manufactured components are produced using consumer-grade hardware, which limits process control, but offers significant upside in scalability and accessibility. Using microscopy and X-ray computed tomography, the VP alumina parts have high bulk densities above 95%, but also the prevalence of AM-induced artifacts and surface defects. Mechanical testing showed these defects to significantly reduce flexural strength and compromise part reliability. Electrorefining trials under sustained exposure to molten salts and metals reveal mixed results, with the AM material exhibiting high chemical compatibility, but mechanical failures due to the reduced strength were prevalent. Our findings illustrate both the promise and current limitations of AM ceramics for actinide chemistry, and point toward future improvements in process optimization, design strategies, and part screening to enhance performance and reliability.

Materials science↗
DOE OSTI2020

Actinide and rare earth drawdown system for molten salt recycle

A method for recycling molten salt from electrorefining processes, the method having the steps of collecting actinide metal using a first plurality of cathodes from an electrolyte bath, collecting rare earths metal using a second plurality of cathodes from the electrolyte bath, inserting the collected actinide metal and uranium into the bath, and chlorinating the inserted actinide metal and uranium. Also provided is a system for recycling molten salt, the system having a vessel adapted to receive and heat electrolyte salt, a first plurality of cathodes adapted to be removably inserted into the vessel, a second plurality of cathodes adapted to be removably inserted into the vessel, an anode positioned within the vessel so as to be coaxially aligned with the vessel, and a vehicle for inserting uranium into the salt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Actinide and rare earth drawdown system for molten salt recycle

A method for recycling molten salt from electrorefining processes, the method having the steps of collecting actinide metal using a first plurality of cathodes from an electrolyte bath, collecting rare earths metal using a second plurality of cathodes from the electrolyte bath, inserting the collected actinide metal and uranium into the bath, and chlorinating the inserted actinide metal and uranium. Also provided is a system for recycling molten salt, the system having a vessel adapted to receive and heat electrolyte salt, a first plurality of cathodes adapted to be removably inserted into the vessel, a second plurality of cathodes adapted to be removably inserted into the vessel, an anode positioned within the vessel so as to be coaxially aligned with the vessel, and a vehicle for inserting uranium into the salt.

Willit, James L.↗
DOE OSTI2025

Small-scale production of bespoke accelerated aging plutonium alloy

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

Aging theory↗
DOE OSTI2025

Assessment of Active Metal Removal from Pyroprocessing Salt by Electrodeposition

Active metals present in used nuclear fuel dissolve into the salt during pyroprocessing, but are not recovered during electrorefining or drawdown operations due to their chemical stability in the salt. The accumulation of 137 Cs and 90 Sr in process salt 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 removing active metals from the molten salt to enable recycle of the electrolyte (e.g., LiCl/KCl) would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Traveling Molten Zone Refining Process Development for Innovative Fuel Cycle Solutions

This ARPA-E ONWARDS project successfully developed and demonstrated an innovative immiscibility zone refining process for separating actinides from active fission products in metallic nuclear fuel applications. The technology addresses a critical challenge in advanced nuclear fuel cycles by providing a high-throughput, low-waste alternative to conventional electrorefining.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2025

Field Testing of Safeguards Technologies in the Hot Fuel Examination Facility

Recent developments in nuclear fuel reprocessing techniques have yielded more efficient processes and fuel cycle options that strengthen the nuclear industry and production of clean energy. One such area of interest is pyroprocessing of used oxide fuel. However, with these advances in the back end of the nuclear fuel cycle, advances in safeguards instrumentation, measurements, and approaches are needed to ensure special nuclear material (SNM) is accounted for according to regulatory requirements. As a high-level overview of a nominal pyroprocessing approach, used oxide fuel from commercial light water reactors (LWR) is mechanically removed from the metallic cladding. Then the fuel is crushed and randomized representative samples are taken and sent to an analytical lab for analysis. The analytical results of the feed material are used for input accountancy into the rest of the process. The crushed oxide fuel is then moved to the oxide reduction (OR) furnace where it is reduced from an oxide to metallic form. The OR product is distilled to remove salt and then moved to an electrorefiner (ER), where it is immersed in a eutectic mixture of lithium chloride potassium chloride (LiCl-KCl) that typically ranges between 450-550 ?. Within the ER, the usable uranium is electrochemically transported through the molten salt from the anode to the cathode, and then subsequently removed as a relatively pure U product.. A simplified model of pyroprocessing techniques with added emphasis on the safeguards can be seen below in Fig. 1

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Gamma-ray spectra analyses of molten salts in spent nuclear fuels pyroprocessing facilities for mass measurement

To evaluate radioactive tracer dilution and gamma spectroscopy as a safeguards technique for monitoring the mass of liquid salt in molten salt systems, the gamma spectroscopy data of salt samples from the pyroprocessing facilities in Idaho National Laboratory were acquired and analyzed. The primary focus is the uncertainty analysis of the gamma radioactivity of 154 Eu isotope in the molten salts. The facilities include an electrorefiner for pyroprocessing spent oxide fuels (SOF-ER) and Mark-IV ER for processing the metallic fuels from Experimental Breeder Reactor-II (EBR-II). The relative gamma radioactivity uncertainty was consistently at 3% for 137 Cs for both Mark-IV ER and SOF-ER salts. However, for 154 Eu, it was 8% for SOF-ER salt and 7% for Mark-IV ER salt. Here, the main reason for the higher uncertainty for 154 Eu is believed to be the lower counting statistics in gamma spectroscopy. To improve the gamma radioactivity uncertainty of isotope 154 Eu, a longer data acquisition time (12 h instead of 4 h) and a higher count rate gamma detector for gamma spectroscopy were tested and it was found that the uncertainty for 154 Eu was significantly improved from 7% or 8–3%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Operation of Argonne's Liquid Salt-Liquid Metal Separation Testbed for U/TRU Product Processing

Argonne National Laboratory has constructed a liquid salt-liquid metal separation testbed for use in the development and advancement of cathode processing of U/TRU co‑deposits generated by pyroprocessing of used nuclear fuel. The U/TRU product recovered from the electrorefiner contains adhered and entrained salt that must be removed prior to consolidation of the U/TRU alloy for use in advanced reactor fuel fabrication. The bottom pour operation utilizes the low melting points of U/TRU co‑deposits and higher densities of molten metals compared to molten salts to separate and consolidate the U/TRU product. Argonne’s testbed is designed to support the development and optimization of bottom-pouring configurations for batch and semi-continuous operations, integration of process monitoring and control technologies, and determination of operational requirements for implementing in an industrial setting. Scoping tests were performed to demonstrate operational aspects of the testbed, including operation using single-pour spout and dual-pour spout configurations, effectiveness of salt containment and extent of salt vaporization, and the use of sensor probes to detect the location of the interface between the metal and salt phases during pouring. Recommendations for process optimization testing for further development of bottom pour processing to separate U/TRU alloys from adhered salt were made based on the results of scoping tests. Completing the recommended activities will increase the technical readiness level (TRL) of the liquid salt-liquid metal separation operation and consolidation of U/TRU alloys to support industrialization of pyroprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗