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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
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 The anolyte is withdrawn and separated using ion exchange techniques.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 OSTI2016

Development and Optimization of Normal Pulse Voltammetry for Real Time Analysis of Electrorefiner Salt with High Concentrations of MgCl{sub 2} and UCl{sub 3}

Electro-refiners can accumulate large holdup of fissile actinides from used nuclear fuel including plutonium and will need to be well safeguarded in a commercial facility. Idaho National Laboratory has operated the Fuel Conditioning Facility, an engineering scale demonstration facility, with two electro-refiners for two decades using mass tracking models supplemented with destructive analysis for the requirements of domestic safeguards. This approach is known to be insufficient of meet international safeguards timeliness requirements. Since there are non-nuclear weapons states that are interested in commercial electro-refiners, there is a need to develop a real time concentration measurement technique. In addition to being useful for safeguards, such a real time measurement would be functional for maintaining optimal process control. Normal pulse voltammetry (NPV) is a powerful electrochemical analysis method that has been identified to have great potential for measuring concentrations of various ions in a given salt mixture. While cyclic voltammetry (CV), chronoamperometry (CA), and chronopotentiometry (CP) result in progressive growth of the working electrode surface area, NPV potentially can circumvent this problem via periodically stripping deposited metals from the electrode (WE) surface. With each successive pulse, the potential is dropped to a more negative level. The corresponding current is a summation of partial currents from each of the reducing species. The ultimate objective of this work is to develop correlations between limiting currents measured via NPV with known concentrations. This is potentially a more accurate method than CV and other methods, because it should keep the WE surface area from growing significantly. Previous work on using NPV for actinide concentration determination in eutectic LiCl-KCl has been reported by M. Iizuka et al. However, only low uranium concentrations were studied in that paper. The maximum concentration of uranium tested was 2.48 wt%. Limiting currents from NPV were found to be proportional to their concentrations up to about 1.7 wt%. They were unable to maintain double plateaus (U and Pu) at uranium concentrations above 1.69 w%. This issue was resolved during our high concentration study. This result represents a significant breakthrough for optimizing voltammetry for analysis of actinides in molten salt.

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

Thermal analysis of projected molten salt compositions during FFTF and EBR-II used nuclear fuel processing

This work studies the change in liquidus temperature of the salt used in the Mk-IV electrorefiner at the Idaho National Laboratory during processing campaigns of sodium-bonded driver fuels from Experimental Breeder Reactor II and the Fast Flux Test Facility reactor. Modeling and Simulation Tool for Electrochemical Recycling Systems (MASTERS), an INL proprietary pyroprocessing flowsheet simulation tool, was used to simulate the processing campaigns and determine the resulting composition changes of the Mark-IV electrorefiner salt. Surrogate salt samples simulating the Mk-IV electrorefiner salt during the fuel processing campaigns were prepared, and the thermal properties were measured via differential scanning calorimetry. Results from this study indicate that after processing approximately 2,150 kg of uranium metal from used fuel, the liquidus temperature of the molten salt exceeds 520°C, which is a significant increase from 352°C (the melting temperature of eutectic LiCl-KCl). It was also discovered that the liquidus temperature is strongly dependent on the amount of NaCl accumulating in the Mk-IV electrorefiner salt. Thus, the assessment of the liquidus temperature during the subsequent fuel processing campaigns can be simplified by considering the ternary LiCl-KCl-NaCl system only.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

First-Principles-Based Computational Study on Nucleation and Growth Mechanisms of U on Mo(110) Surface Solvated in an Eutectic LiCl-KCl Molten Salt

In nuclear engineering and industry electrochemical methods have been intensively employed in pyro-processing technology to recycle used nuclear fuels for substantial alleviation of the concern with the safe treatment of high-level radioactive nuclear wastes. Pyro-processing is based on sequential electrochemical reactions at high temperature (typically above 700 K at ambient pressure) to enhance the efficiency for decoupling nuclear fuels from unfavorable fission fragments. Using a molten salt in electrorefining process, at which actinides can be selectively deposited in solid cathode from ionic states, gives several advantages over an aqueous solution: higher thermal conductivity (it is liquid phase at high temperature, molten salt plays a role as medium at high temperature over 773 K in electrochemical reaction, lower kinetic barrier in ionic diffusion mobility and better stability against undesired side reactions with radioactive materials. Therefore, the volume of high-level nuclear wastes to be, otherwise, disposed into secured sites is dramatically reduced. Most of all it conforms to nuclear non-proliferation treaty. Electrochemical reactions in a molten salt medium are, however, much more complicated than in aqueous solution rendering the accurate understanding of thermodynamic and kinetic properties formidably challenging. It is represented by extensive researches in experiments and computations to identify underlying mechanisms on how U and other fission fragments behave in the electrorefining system. Regardless of the researches there is a general agreement that the initial stage of the nucleation and growth process of the U deposit is critical for determining the properties of final product and the electrorefining efficiency. In spite of the focused studies it still remains unclear. Theoretical and computational researches have been often carried out with various empirical parameter for interatomic potential energy, which are unknown variables until numerical fits to experimental measurements or just assumption is made. Thus, empirical simulations sometimes predicted outcomes are inconsistent depending on materials of interest with experimental measurements. This may be largely because the accuracy of an empirical interatomic potential energy models critically depends on the input parameters and thus, can be limited in transferability of the results. Furthermore, it may be devoid of fundamental physics or chemistry. This situation requires special caution when an electrorefining system is studied to recycle spent nuclear fuels because of the extreme difficulty in reliable acquisition of thermodynamic and kinetic data from direct experimental measurements. In this presentation, we extensively utilized first principles density functional theory (DFT) calculations and ab-initio molecular dynamic (AIMD) simulations to figure out underlying mechanisms of the initial stage of nucleation and growth behaviors of U on Mo(110) surface in a eutectic LiCl-KCl molten salt at T = 773 K. To rigorously investigate the effect of the molten salt electrolyte we compared the results with those calculated under vacuum condition.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Fundamental Electrochemical Properties of Liquid Metals in LiCl-KCl for Separation of Alkali/Alkaline-Earths (Cs, Sr, and Ba)

In electrorefiner, uranium is recovered from used nuclear fuel using an electrorefining process in which a metallic used nuclear fuel anode is oxidized into molten LiCl-KCl-UCl 3 electrolyte and pure U is preferentially reduced onto an inert cathode. While electrorefiner systems facilitate the recycling of substantial amounts of uranium from used nuclear fuel, they also contribute to the production of nuclear waste due to the build-up of fission products such as 90 Sr and 137 Cs in the molten salt electrolyte as they are electrochemically more active than U. The accumulation of alkali/alkaline-earth elements (Ba, Sr, Cs) in the electrolyte presents a problem as Sr and Cs isotopes have high heat densities and produce large amounts of highly ionizing radiation; these hazards combined with difficulty in removing the highly stable alkali/alkaline-earth elements from the electrolyte necessitates frequent replacement and disposal of the electrolyte, which then increases the overall volume of nuclear waste. This research project focuses on evaluating the viability of using liquid metal electrodes for electrochemical separation of alkali/alkaline-earths from molten salts utilizing the strong atomic interactions between candidate liquid metals and alkali/alkaline-earths. The strength of chemical interactions was quantified by determining the thermodynamic properties (e.g., activity) in liquid metals of Bi, Sb, and Pb. Thermodynamic properties, including activities, partial molar entropies, and partial molar enthalpies, were determined using electromotive force measurements for the Sr-Bi, Sr-Sb, Sr-Pb, Ba-Bi, and Ba-Sb binary systems in order to elucidate the strength of interactions between the alkaline-earth elements and each liquid metal, and to develop a comprehensive understanding of their behavior. Activities as low as a Sr = 10 -13 at x Sr = 0.04 at T = 888 K were measured as well as liquid state solubility as high as 40 mol% at 988 K in Sr-liquid metal systems; activities as low as aBa = 10 -15 at xBa = 0.05 at T = 888 K with liquid state solubility as high as 30 mol% at 988 K in Ba-liquid metal systems. Experimental data was used as input data towards computational efforts involving first-principles calculations as well as the CALPHAD (CALculation of PHAse Diagram) technique in the case of the Sr-Sb and Ba-Bi systems to develop improved phase diagrams and provide further basis for the use of computational models in elucidating strongly interacting binary systems. Attempts to remove Sr and Ba from molten salt electrolyte using an electrochemical cell with liquid metal cathodes were successful, with post-mortem elemental analysis of the electrodes confirming significant quantities of Sr (6.5 mol%) and Ba (12.8 mol%) deposited into Bi. Furthermore, deposition results correlated well with the deposition behavior predicted from the aforementioned electromotive force measurements, inviting the possibility of using liquid metal electrodes for selectively removing alkali/alkaline-earth elements from molten LiCl-KCl electrolyte to recycle the process salt in electrorefiner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2016

Phase Equilibria in Systems Relevant to Pyroprocessing

Pyroprocessing has many advantages for the recovery and recycle of actinides, especially the transuranic elements, in used nuclear fuel. It has several unique features that distinguish it from other used fuel treatment technologies such as the ability to handle short-cooled fuel, a small process footprint, and production of metallic uranium and uranium - transuranic (U/TRU) products that can be directly fabricated into metallic fuel for advanced reactors. Recovery of the uranium and U/TRU products from the used fuel occurs in an electrorefining step. Uranium and transuranics in the used fuel (metal or reduced oxide) are dissolved at the anode in a molten chloride electrolyte, and electrodeposited at two different cathodes - one for uranium and another for the uranium - transuranic product. During this process, fission products less noble than uranium and transuranics (i.e., lanthanides, alkali and alkaline earth) anodically dissolve into the electrolyte, while only uranium and U/TRU are electrodeposited at the respective cathodes. Thus, there is a buildup of active metal and lanthanide fission products in the molten salt electrolyte. The goal of this work is to improve the fundamental understanding of the chemistry of actinides and transuranics in this complex, multicomponent solution. The fundamental thermodynamic properties of the multi-component molten salt electrolytes are complicated due to the formation of ternary and more complex compounds. As the actinides anodically dissolve at the anode and deposited at the cathode, a gradient in the concentration develops at the anode/electrolyte and the cathode/electrolyte interfaces. Ternary phase formation could impact the performance of the electrorefining system by influencing the chemical activity and thus the redox potentials of the actinide chlorides in the molten salt solution. Literature data on the LiCl-KCl-UCl{sub 3} system is limited, sometimes conflicting, and may reflect the use of older analytical instrumentation, which was less accurate than that currently available. In addition, there has not been any investigation of the low UCl{sub 3} concentration region of the system, which is the region relevant to electrorefining. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2018

Waste form evaluation for RECl 3 and REO x fission products separated from used electrochemical salt

The work presented here is based off the concept that the rare earth chloride (RECl3) fission products mixture within the used electrorefiner (ER) salt can be selectively removed as RECl3 (not yet demonstrated) or precipitated out as REOCl through oxygen sparging (has been demonstrated). This paper presents data showing the feasibility of immobilizing a mixture of RECl3’s at 10 mass% into a TeO2-PbO glass and it shows that this same mixture of RECl3’s can be oxidized to REOCl at 300°C and then to REOx by 1200°C. When the REOx mixture is heated at temperatures >1200°C, the ratios of REOx’s change. The mixture of REOx was then immobilized in a LABS glass at a high loading of 60 mass%. Both the TeO2-PbO glass and LABS glass systems show good chemical durability. The advantages and disadvantages of tellurite and LABS glasses are compared.

Riley, Brian J. (ORCID:0000000277456730)↗
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↗