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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Evaluation of Iron‐Phosphate Glass–Ceramic Waste Form for Electrorefiner Salt Waste Simulant Dechlorinated With Phosphoric Acid

The importance of glass and glass–ceramic nuclear waste forms has been reaffirmed in recent years by the growing interest in nuclear power as a reliable energy source. Determination of processing methods for the disposal of halide-containing wastes will be essential for the advancement of nuclear technologies such as non-aqueous fuel reprocessing. Phosphate-based dechlorination and subsequent vitrification of radioactive salt waste into an iron-phosphate waste form have been identified as a potential processing scheme for electrochemical processing waste. The impact of H 3 PO 4 -based dechlorination of complex salt mixtures on the vitrification process and structure of the final iron-phosphate waste form has not yet been investigated. In this work, iron-phosphate glass–ceramics were made from simulant salt waste (48LiCl–33KCl–19NaCl mol%) dechlorinated with the H3PO4-based method. The glass-forming region was compared to that of traditionally prepared Na 2 O–Fe 2 O 3 –P 2 O 5 systems. For a candidate glass-forming composition, the processing scheme presented here was determined to favor Fe 3+ species. The O/P molar ratio was consistent for the candidate composition when dechlorinated at 400°C and 600°C in air and argon environments, indicating glass network connectivity was maintained despite variations in processing parameters. The results presented here validate processing schemes requiring iron-phosphate waste form synthesis following H 3 PO 4 -based dechlorination.

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

Final Design for Thermal/Epithermal eXperiments (TEX) with Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility

One of the main goals of the Thermal/Epithermal eXperiments (TEX) project is to use existing Nuclear Criticality Safety Program (NCSP) assets to create critical experiment plutonium and uranium test beds for materials important to criticality safety that have insufficient benchmark evaluations. The plutonium test bed experiments were completed in 2018 and are published in the 2020 edition of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The uranium test bed assemblies were measured in 2020, with benchmark preparation underway.

42 ENGINEERING↗
DOE OSTI2025

Developments in Processing Criteria for Iron Phosphate Waste Form Design for Treating Electrorefiner Salt Wastes

This report describes work done under a Nuclear Energy University Partnership (NEUP) project that included a collaborative effort between Pacific Northwest National laboratory, University of Nevada Reno, University to Utah, and Missouri University of Science and Technology. In this project, the main goals were the following: (1) to assess H3PO4 dechlorination parameters for halide removal from salt simulants, (2) to evaluate crucible compatibility with iron phosphate glass melts, and (3) to evaluate waste form properties within a ternary system of P2O5 (added as H3PO4), Fe2O3 and fission product simulants.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

TEX-Chlorine: Highly Enriched Uranium with Chloride Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility

This report documents the final benchmark of the TEX-Chlorine (IER-499) Thermal Epithermal eXperiments (TEX) with highly enriched uranium with chlorine absorbers and high-density polyethylene reflectors and moderators. TEX-Chlorine is a variation of the TEX-HEU baseline assembly with the addition of sodium chloride absorber plates. This evaluation contains three experimental configurations that were performed on Comet at NCERC between July and August 2024. The three configurations, which were acceptable as benchmark cases, spanned from thermal (first two configurations) to fast (third configuration). All three cases were reviewed and accepted by the ICSBEP TRG in April 2025 and was submitted to the ICSBEP in August 2025 after receiving subgroup approval.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Final Design for Thermal/Epithermal eXperiments (TEX) with Lithium Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility (IER 575, CED-2 Report)

One of the main goals of the Thermal/Epithermal eXperiments (TEX) project is to use existing Nuclear Criticality Safety Program (NCSP) assets to create critical experiment plutonium and uranium test beds for materials important to criticality safety that have insufficient benchmark evaluations. The plutonium test bed experiments were completed in 2018 and are published in the 2020 edition of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The uranium test bed assemblies were completed in 2023 and accepted in the 2024 edition of the ICSBEP Handbook.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1981

Electrorefining process for lunar free metal - Space and terrestrial applications and implications

An electrochemical refining process is proposed for the separation and recovery of principal and trace elements from reduced metallic particles found in lunar soils. A process variation is presented for purification and recovery of chromium and manganese from electrodeposited impure iron available from lunar silicate and other minerals. The process involves anodic dissolution of impure metal and cathodic deposition in divided cells using aqueous chloride solutions. The anolyte is withdrawn and separated using ion exchange techniques.

Waldron, R. D.↗
DOE OSTI2024

SCALE Demonstration for Sodium-Cooled Fast Reactor Fuel Cycle Analysis

In support of the US Nuclear Regulatory Commission non-light-water reactor fuel cycle demonstration project, SCALE 6.3.1 capabilities for radionuclide characterization, criticality, and shielding were demonstrated for scenarios in the sodium-cooled fast reactor (SFR) nuclear fuel cycle. Three postulated accident scenarios were selected for analysis in this work. As a basis for all scenarios, irradiated fuel inventories were generated using SCALE/ORIGAMI. To cover multiple SFR design choices, two different types of SFR fuel, uranium/transuranic-loaded and U-based fuels, were considered. For the first scenario, SCALE/MAVRIC was used to calculate the radiation shielding and dose rates inside and outside of the containment building due to a drop of a spent fuel assembly from the fuel-handling system during unloading inside the containment building. For the second scenario, potential critical configurations in an electrofiner were investigated through criticality calculations with SCALE/CSAS. For the third scenario, the activity of the waste salt from an electrorefiner was evaluated using SCALE/ORIGEN. The dose rate produced by the analyzed SFR assemblies is similar to that produced by a typical pressurized water reactor (PWR) fuel assembly with a discharge burnup of 50 GWd/MTU, with the same cooling time of 10 days. The criticality analyses suggested that the different electrorefiner configurations have a large margin to criticality. The activity analysis of the electrofiner waste revealed that shielding and cooling may be required for the waste salt that contains transuranics and fission products produced by the electrorefiner because of the high activity of the waste salt. In general, the application of various capabilities in the SCALE code system for SFR fuel inventory generation, criticality, and shielding was successfully demonstrated for the selected scenarios in the SFR nuclear fuel cycle. Additional analyses can be performed to provide more accurate results when more details of the SFR nuclear fuel cycles are available, for example, the dimension of the electrorefiner and the salt compositions during reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Salt Sampling FY21 Technical Report

The goal of the salt sampling program at Argonne is to develop and deploy automated molten salt sampling approaches to enable high-precision in-process salt sample analysis to improve the timeliness of sampling-based accountancy measurements. Tools currently under development in support of this goal include (1) a modular vacuum sampler with an accompanying sample handling method for coupling vacuum sampling with high-precision at-line sample analysis, (2) a pneumatic sample generator that enables high-throughput sample analysis to improve the precision of existing analytical techniques, and (3) a windowless flow cell to enable on-line optical analysis of molten salt in a sampling loop. Compared to point sampling approaches (i.e., dip probes), vacuum sampling systems and on-line sampling loops facilitate access to a larger cross-section of a process fluid. This is known to improve the characterization of the process fluid by producing more representative samples and by enabling the analysis of a larger cross section of the fluid. A vacuum sampling approach for molten salts eliminates the risk of dross contamination of samples and avoids the use of moving parts in the salt. In FY21, two methods for integrating a vacuum sampler with a pneumatic sample generator were tested. These included direct fluidic coupling and coupling using a solid salt transfer mechanism. Solid salt transfer was ultimately selected over fluidic coupling, primarily to enable the transport of samples over longer distances to support automated at-line integration with high-precision techniques (such as microcalorimetry) that cannot withstand the extreme conditions near an electrorefining process. To facilitate rapid solid salt coupling, new mechanisms were developed for rapidly charging and discharging salt sample tubes at the vacuum sampler and pneumatic sample generator, respectively. While the charging mechanism will be deployed in FY22, the tube transfer method and discharge mechanism were tested in FY21. These were deployed at one of Argonne’s engineering-scale electrorefiners to implement at-line high-throughput pneumatic micro-sample generation capabilities. The method was used to generate precise uranium- and lanthanide-bearing electrorefiner micro-samples with the specific dimensions requested by researchers at Los Alamos National Laboratory for use in testing their novel microcalorimeter x-ray techniques. The solid salt transfer mechanism proved not only to be an effective means of integrating the precision sample generator with vacuum sampling, but also improved the performance of the sampler generator. Because the modular sampling approach described here eliminates the need for new high-radiation sample handling capabilities, salt-wetted seals, salt-wetted moving parts, and heated transfer lines outside the electrorefiner, it will address most of the remaining technical challenges for the at-line deployment of high-precision analytical techniques. This will enable significant reductions in the time delay for sampling-based accountancy measurements by eliminating the need for manual off-line sample processing and analysis. On-line optical analysis of molten salt in a sampling loop would provide complementary information to at-line and in-situ techniques. In FY21, an open-aperture molten salt gravity flow cell with windowless optical access to flowing salt was successfully demonstrated. Future work should include the refinement and performance testing of the on-line and at-line sampling tools, integration of additional analysis techniques, stakeholder outreach and collaboration, evaluation of the integrated methods, and analyses to determine how the various tools might fit into an integrated safeguards monitoring system of unattended near real time monitoring tools.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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

Application of Advanced Materials Processing to Enable Direct Production of Fast Reactor Fuel Alloys

Argonne National Laboratory (the Contractor), located in Lemont, IL, and Oklo, Inc. (the Participant), headquartered in Sunnyvale, CA, entered into a Cooperative Research and Development Agreement (CRADA) to integrate advanced electrorefining co-deposition and molten-salt monitoring technologies to produce a uranium-transuranic (U/TRU) alloy within a controlled composition range that can be used as fast reactor fuel for Oklo, Inc.’s advanced reactor technology. Argonne performed the integration of electrorefining and process monitoring technologies, determined operating parameters for producing U/TRU alloys, and developed optimized design and operating parameters for co-deposition of U/TRU alloys. Oklo, Inc. worked with Argonne and the Nuclear Regulatory Commission (NRC) to provide the necessary process documentation to begin implementing pyroprocessing technology in the fast reactor fuel production process. Outcomes of this project included a pilot-scale co-deposition cathode design and technical basis for the operation of the co-deposition electrorefiner to produce U/TRU alloys with controlled composition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Study of Potentiometry for Monitoring Activity of GdCl 3 in Molten LiCl-KCl Salt

In pyroprocessing spent nuclear fuels by electrorefining in molten LiCl-KCl salt, it is desired to monitor in real time the UCl 3 concentration in the salt for safeguards purposes. Current chemical analysis of the highly radioactive salt for electrorefining by an inductively coupled plasma technique is inconvenient and usually time-consuming in generating the salt composition results. In this paper, we evaluated whether a simple potentiometry approach can be used for real-time monitoring the concentration of GdCl 3 , which was used as a surrogate for UCl 3 , in LiCl-KCl-GdCl 3 salt by measuring the open circuit potential of a Gd metal electrode with respect to a Ag/AgCl reference electrode (RE) when GdCl 3 salt was incrementally added to the LiCl-KCl salt. Additions of LaCl 3 , CeCl 3 and NdCl 3 salts were used for evaluating the effects of other chloride salts on the selectivity of the Gd metal electrode vs Ag/AgCl RE. While using potentiometry to determine GdCl 3 concentrations, Gd metal was unexpectedly observed to be unstable and dissolved in LiCl-KCl salt when GdCl 3 is present.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development, feasibility, and uncertainty of radioactive 22 Na tracer dilution and gamma spectroscopy for mass determination of molten salt for pyroprocessing spent nuclear fuels

Accurate knowledge of the total mass of molten salts both for pyroprocessing spent nuclear fuels and for molten salt reactors is necessary for safeguards purposes. However, it is challenging to know or measure the total mass of molten salt due to the complicated shapes as well as the compositional—and thus density—changes that occur over time during operation. Here, we investigated radioactive tracer dilution (RTD) as a potential safeguards technique for application to the total mass measurement of LiCl–KCl–UCl 3 salt used for uranium electrorefining, with a focus on the feasibility, uncertainty, and fission product effect of RTD for salt mass measurement. To this end, experimental-scale RTD tests (600–700 g of total mass) were initially carried out by adding 22 Na tracer salt into LiCl–KCl salt containing 0, 5, and 28wt% radioactive salts from an electrorefiner (ER) for refining uranium. Upon the completion of the experimental-scale RTD tests, we performed a scale-up RTD test by directly adding 22 Na tracer salts to the ER, which uses about 12 kg of LiCl–KCl–UCl 3 salt during normal operation. This paper reports the main features of RTD for total mass measurement of molten salt and summarizes the results of the experimental-scale and scale-up RTD tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Electrochemical behavior of SnCl 2 and influence of Cu and Ni ions in molten LiCl−KCl−CaCl 2 eutectic

Reliable transport and thermodynamic data for multivalent ions in complex molten salts are scarce, limiting model fidelity for electrorefining and impurity control. Here, we report a comprehensive electrochemical characterization of SnCl₂ in LiCl–KCl–CaCl₂ (50.5–44.2–5.3 mol%) at 685 K, including the effects of Ni 2+ and Cu + impurities. Using cyclic voltammetry (CV), chronoamperometry (CA), and chronopotentiometry (CP), we quantified Sn 2+ and Ni 2+ diffusion with exceptional agreement across methods: Sn 2+ averaged (1.03 ± 0.10) × 10 −5 cm 2 s −1 , and Ni 2+ averaged (0.75 ± 0.19) × 10 −5 cm 2 s −1 . The tight confidence-interval overlap across CV, CA, and CP strengthens confidence in these values and is uncommon in molten chloride studies. Open-circuit-potential measurements provided standard apparent reduction potentials that closely match LiCl–KCl literature, indicating minimal shift with CaCl₂ present. The Sn 2+ /Sn couple behaves as a reversible two-electron soluble–insoluble process at 685 K; the Sn 4+ /Sn 2+ couple transitions to soluble–soluble behavior near 788 K, which may correlate with the decomposition of surface bound chlorostannates, though direct characterization remains to be established. In mixed systems, Cu+/Cu overlaps Sn 2+ /Sn, limiting Cusingle bondSn electroseparation, whereas the larger potential gap between Ni 2+ /Ni and Sn 2+ /Sn supports selective Ni removal. These internally consistent transport and thermodynamic data establish a validated basis for process modeling and optimization of Sn electrorefining and impurity management in LiCl–KCl–CaCl₂.

Berzins-Delahay↗
DOE OSTI2023

Thermodynamic description of molten salt systems: KCl-LiCl-NaCl and KCl-LiCl-NdCl 3

The mixture of KCl and LiCl has been used as electrolyte in the electrorefining process to recover uranium from used nuclear fuels due to the low melting point. However, lanthanides and sodium in the reactor waste continuously dissolve into it and thus alter its thermodynamic properties. To understand the thermodynamic behavior evolution of the electrolyte with the accumulation of impurities, thermodynamic modeling for KCl-LiCl-NaCl and KCl-LiCl-NdCl 3 and four constituent binary systems in the entire composition space was performed using the CALPHAD (CALculation of PHAse Diagrams) approach. The ionic liquid was described by the two-sublattice model, where neutral species were introduced to consider short-range ordering (SRO) within the melt, whereas the solid solution was modeled based on the Compound Energy Formalism. Literature data on phase equilibria and thermochemical properties were critically evaluated and used during the optimization of thermodynamic parameters for KCl-LiCl-NaCl and KCl-LiCl-NdCl 3 and their subsystems. The calculated phase diagrams and mixing enthalpies are in good agreement with the experimental data. The thermodynamic modeling for the KCl-LiCl-NdCl 3 system was carried out for the first time. To fill the gap in experimental measurement, enthalpy of mixing for the KCl-LiCl-NdCl 3 melt was estimated using the surrounded-ion model. These data then served as critical inputs for thermodynamic optimization. Furthermore, the present study can provide insights into thermodynamic property evolution of the electrolyte and solubility limit of various impurities during the electrorefining process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

In-situ anodic precipitation process for highly efficient separation of aluminum alloys

Electrorefining process has been widely used to separate and purify metals, but it is limited by deposition potential of the metal itself. Here we report in-situ anodic precipitation (IAP), a modified electrorefining process, to purify aluminium from contaminants that are more reactive. During IAP, the target metals that are more cathodic than aluminium are oxidized at the anode and forced to precipitate out in a low oxidation state. This strategy is fundamentally based on different solubilities of target metal chlorides in the NaAlCl4 molten salt rather than deposition potential of metals. The results suggest that IAP is able to efficiently and simply separate components of aluminum alloys with fast kinetics and high recovery yields, and it is also a valuable synthetic approach for metal chlorides in low oxidation states.

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↗