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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Safeguards in Pyroprocessing: an Integrated Model Development and Measurement Data Analysis

Pyroprocessing is an electrochemical method based on the molten salt electrolyte, mainly the LiCl-KCl eutectic molten salt, to recycle the used nuclear fuel. For a conceptual design of commercial pyroprocessing facility, tons of special nuclear materials, namely U and Pu, may be involved, which could be used for non-peaceful purposes if they are diverted. Effective safeguards approaches have to be developed prior to the development and construction of a pyroprocessing facility. Present research focused on two main objectives, namely calculating the properties of nuclear species in LiCl-KCl molten salt and developing integrated model to safeguard a pyroprocessing facility. Understanding the characteristics of special nuclear materials in LiCl-KCl eutectic salt is extremely important to understand their behaviors in an electrorefiner. The model development for the separation processes in the pyroprocessing, including electrorefining, actinide drawdown, and rare earth drawdown benefits the understanding of material transport and separation performance of these processes under various conditions. The output signals, such as potential, current, and species concentration contribute to the material balance closure and provide safeguards signatures to detect the scenarios of diversion. U and Pu are the two main elements concerned in this study due to our interest in safeguards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

An Initial Study of Uranium Morphology in Fused LiCl-KCl

Molten salt electrorefining is one method by which reusable constituents within used nuclear fuel can be recovered for recycle and especially radiotoxic isotopes removed for incorporation into tailored waste forms. In these processes, irradiated metallic uranium or uranium alloys are dissolved at an electrochemical anode and crystals of purified uranium are collected at a cathode. The morphology of the deposited crystals can impact a variety of important factors, including the adhesion characteristics of the deposit to the cathode surface and the amount of salt which is macroscopically and microscopically occluded with the deposit. In addition, future applications such as direct creation of nuclear fuels by electrochemical deposition might be enabled if the science and fundamentals of uranium deposition morphology were thoroughly understood. A reasonable amount of experience has been collected from uranium electrorefining within the range of typical plant conditions. However, only limited laboratory study has been performed to define and understand the range of uranium crystal morphologies under repeatable laboratory conditions. The present study was an initial step to define these morphologies and transitions. The deposit from each deposition potential was primarily composed of a different morphology with little crossover of that morphology into the other deposition potentials. The spacing of these experiments (overpotentials of 25, 150 and 450 mV) is wide enough that additional dominant morphologies could exist. In addition, while the majority of the deposit from each potential was comprised of a particular morphology, minor morphologies were also found within the deposit. This may indicate the presence of an additional dimension of complexity, namely uranium concentration. Additional experimental and theoretical studies would be valuable to elucidate this subject. Additional results will be presented and discussed. (authors)

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 OSTI2020

Electrochemical Salt Wasteform Development: A Review of Salt Treatment and Immobilization Options

Electrochemical reprocessing, also referred to as pyroprocessing, is a technique for recycling actinides from used nuclear fuel (UNF) to produce fuel for future reactors. Here, UNF is dissolved in a molten salt (e.g., LiCl-KCl eutectic) within an electrorefiner. After UNF dissolution, fission products are released into the electrolyte salt and converted to chlorides. This paper discusses wasteform options for processing the base electrolyte salt with the fission product salts as well as just the rare-earth fission products (as RECl 3 , REOCl, or REO x ) with the intent of finding optimal methods for reducing total waste salt volumes or partitioning the salt for alternate wasteform options. Furthermore, two of the more detailed partitioning options discussed herein include halide removal from the salt (dehalogenation), which accounts for more than half of the salt on a molar basis, and RE fission product removal for wasteforms with high-RE loadings. Wasteform properties are compared with emphasis on wasteform volume starting from a given amount of (1) total salt cations or (2) RE cations. Comparisons are also made of wasteform chemical durabilities, with the data available from like testing methods. A main conclusion from this work is the justification of subsequent salt processing after electrorefiner operations for achieving significant wasteform volume reduction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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↗
DOE OSTI2016

State-of-the-art in Manufacturing Pyrochemical Processing Crucibles

Engineers at Los Alamos National Laboratory have been purifying plutonium through pyrochemical processing for many years. The harsh environments involved with the pyrochemical processes include high temperatures (700-900 °C), molten salts and reactive metals (MgCl 2 , PuCl 3 , NaCl-KCl, CaCl 2 , Pu, Ca), and chlorine gas. Because of these corrosive conditions, designing and developing the pyrochemical processing equipment has always been an intriguing technical challenge. In this article I focus on the design and development of the pyrochemical processing crucible for the electrorefining (ER) process, applying design for manufacturability principles throughout the process. I review the current crucible design and electrorefining process and explore possibilities of making the crucible as one part, building the anode and cathode into the crucible, and various material of construction improvements. Materials of construction that should be reviewed further include alumina, silicon nitride, and a tantalum carbon alloy. I briefly review the application of three-dimensional (3D) printing to the process of producing the ER crucible, including thermoplastic 3D printing and mechanical property concerns. I finish by summarizing current plans and possibilities of improving the ER crucible.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Water Solubility of Plutonium and Uranium Compounds and Residues at TA-55

Understanding the water solubility of plutonium and uranium compounds and residues at TA-55 is necessary to provide a technical basis for appropriate criticality safety, safety basis and accountability controls. Individual compound solubility was determined using published solubility data and solution thermodynamic modeling. Residue solubility was estimated using a combination of published technical reports and process knowledge of constituent compounds. The scope of materials considered includes all compounds and residues at TA-55 as of March 2016 that contain Pu-239 or U-235 where any single item in the facility has more than 500 g of nuclear material. This analysis indicates that the following materials are not appreciably soluble in water: plutonium dioxide (IDC=C21), plutonium phosphate (IDC=C66), plutonium tetrafluoride (IDC=C80), plutonium filter residue (IDC=R26), plutonium hydroxide precipitate (IDC=R41), plutonium DOR salt (IDC=R42), plutonium incinerator ash (IDC=R47), uranium carbide (IDC=C13), uranium dioxide (IDC=C21), U 3 O 8 (IDC=C88), and uranium filter residue (IDC=R26). This analysis also indicates that the following materials are soluble in water: plutonium chloride (IDC=C19) and uranium nitrate (IDC=C52). Equilibrium calculations suggest that PuOCl is water soluble under certain conditions, but some plutonium processing reports indicate that it is insoluble when present in electrorefining residues (R65). Plutonium molten salt extraction residues (IDC=R83) contain significant quantities of PuCl 3 , and are expected to be soluble in water. The solubility of the following plutonium residues is indeterminate due to conflicting reports, insufficient process knowledge or process-dependent composition: calcium salt (IDC=R09), electrorefining salt (IDC=R65), salt (IDC=R71), silica (IDC=R73) and sweepings/screenings (IDC=R78). Solution thermodynamic modeling also indicates that fire suppression water buffered with a commercially-available phosphate buffer would significantly reduce the solubility of PuCl 3 by the precipitation of PuPO 4 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Experimental Study of Codeposition Electrochemistry Using Mixtures of ScCl 3 and YCl 3 in LiCl-KCl Eutectic Salt at 500°C

Cyclic voltammetry and chronopotentiometry tests were applied to molten LiCl-KCl eutectic at 500 °C including amounts of ScCl 3 and YCl 3 . The purpose of the testing was to observe the effect of applied electrical current on the codeposition of scandium and yttrium, which were chosen as surrogate elements for uranium and plutonium, respectively. Features of the work were to vary the concentration of ScCl 3 (at relatively low concentrations) as well as varying the applied current, all with a fixed concentration of YCl 3 . Results of the experiments could provide insight of uranium electrorefining and may provide evidence, which suggests the electrorefiner could be operated at lower UCl 3 concentration whereby codeposition (U and Pu) could be more effectively controlled.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Extraction, Separation, and Production of High Purity Rare Earth Elements and Critical Minerals from Coal-Based and Related Resources

The general objectives of this project are to develop concepts for rare earth metal and critical mineral production from coal-based and related (minerals associated with coal) resources and incorporate them into a Technical Research Plan with an associated overall flow sheet. The project team has extensive expertise in market evaluation, mineral separation, leaching, chemical separations, alternative metallothermic reaction technology, electrowinning, and electrorefining that was critical to the success of the proposed project. The project included critical industrial partners needed for project success. The project team has extensive expertise in market evaluation, mineral separation, leaching, chemical separations, alternative reduction technology, electrowinning, and electrorefining as well as appropriate pilot-scale facilities to enable this project. The project encompassed broader opportunities for developing domestic resources of REE/CM materials for a more resilient, diverse, and secure supply chain for REE/CM materials with built-in redundancies and appropriate resource stockpiles. The production of REE/CM materials will help to revitalize and rebuild world-class American manufacturing capacity and the related workforce through new jobs and infrastructure. Furthermore, due to the nature and location of the production sites, economic growth in diverse communities of color and economically distressed areas will be cultivated. Finally, this technology can be applied to reuse and remediate coal waste tailings for REE/CM production.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2023

Application of Principal Component Analysis to Electrochemical Reprocessing PM and NMAC

In this report, data from an electrorefiner (ER) for nuclear fuel reprocessing is evaluated for process monitoring (PM) conclusions. This data comes from tests performed at the Idaho National Laboratory in 2022. Multivariate approaches utilizing methods of Principal Component Analysis (PCA) is applied. This is based off established work in process monitoring for fault detection in industrial facilities. This report will discuss the background, methods, and results of the application and some of the conclusions and applications that can be drawn from them. PCA is applied to two different electrorefiner (ER) operations that occurred at Idaho National Laboratory between August and October 2022. The first operation occurred with little incident while the second had several noted faults in the equipment in operational logs. The data from the first run was used to train the data for “normal” operations and applied to both sets of data to determine when operations were in an “off-normal” condition and identify where the fault occurs through PCA. PCA was able to identify off-normal events and identify the cause for off-normal operations. These identified off-normal events matched with the events and their causes in the operational logs. However, small amounts of variance in the data led to false detection of “off-normal” events. Thus, careful selection of training data and a-posteriori conclusions based off operator assessments will both be required for application of PCA to PM applications. This work demonstrated that multivariate approaches and latent variables are applicable to pyroprocessing PM applications and can be further expanded in future work as quality variables such as salt concentration from sensors and sampling become available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Deliquescence of eutectic LiCl-KCl diluted with NaCl for interim waste salt storage

Molten eutectic LiCl-KCl salt is a widely used electrolyte for electrorefining uranium from spent nuclear fuel. Due to the hygroscopic nature of this salt, careful handling and storage of solidified waste electrorefiner (ER) salt is required to avoid deliquescence and corrosion of container materials. This study investigated a potential processing path for reducing the deliquescence through dilution with NaCl. The hydration behavior of LiCl-KCl salts diluted with NaCl was evaluated in terms of mass gain due to mass of water absorption, visual evidence of deliquescence, and visual evidence of corrosion to stainless steel containers in a humid air environment. Pure eutectic LiCl-KCl exhibited a 50 mass% increase due to water absorption and showed signs of deliquescence after 24 hours. 89 mass% NaCl was required in order to prevent deliquescence. Dilution with 89% NaCl was also found to protect corrosion to stainless steel crucibles. Corrosion, thus, appears to require deliquescence. While NaCl dilution greatly decreases steady state hydration and eliminated deliquescence, storage volume is increased ~10x.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Synthesis of americium trichloride via chlorination of americium oxide using zirconium tetrachloride in LiCl-KCl molten salt

To better inform electrorefining operations of used nuclear fuel, a fundamental understanding of the electrochemical properties of each actinide in the molten salt electrolyte is needed. Access to actinide chlorides is thus required to support measurement of these properties. In this work, we synthesized americium trichloride through a novel pathway comprising chlorination of Am2O3 with ZrCl4 in a LiCl-KCl molten salt at 500 degrees C. The formation of AmCl3 was confirmed by cyclic voltammetry and gamma spectroscopy. This chlorination method does not produce mixed hazardous-radioactive waste and the byproduct, ZrO2, is not electrochemically active and does not affect the actinide electrodeposition reaction.

actinide chloride synthesis↗
DOE OSTI2024

Experimental determination of the electrochemical properties of bismuth chloride in eutectic LiCl–KCl and LiCl–KCl–CaCl 2 molten salts

Bismuth has been investigated as a potential liquid electrode for molten salt electrorefining, but the electrochemical behavior of Bi 3+ /Bi redox couple has received scant attention in LiCl–KCl and LiCl–KCl–CaCl 2 eutectics. Electroanalytical techniques were used to determine properties of Bi 3+ /Bi. Diffusivities of 6.71 and 8.93 × 10 –6 cm 2 s –1 and standard apparent potentials of 0.1166 and 0.1187 V versus Ag/AgCl (4.75 mol%) were recommended for BiCl 3 in LiCl–KCl and LiCl–KCl–CaCl 2 eutectics, respectively, at 680 ± 3.1 K. The ICP-MS analysis of BiCl 3 content limited the accuracy of the results. Furthermore, the reliability of the results improved with a more thorough drying procedure for CaCl 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Direct Observation of Pure Cu and Cu-Ag Anode Passivation in H{sub 2}SO{sub 4}-CuSO{sub 4} Aqueous Solution by Channel Flow Double Electrode and Optical Microscopy

The dissolution and passivation of pure Cu and Cu-5 wt pct Ag anodes in H{sub 2}SO{sub 4}-CuSO{sub 4} electrolyte were investigated by a direct observation method that combined the techniques of optical microscopy and channel flow double electrode. Linear sweep voltammetry of the anodes showed that the dissolution of Cu transited from the charge transfer-controlled reaction to the mass transfer-controlled reaction, followed by the passivation of the electrodes. The direct observation of the pure Cu anode revealed that Cu particles were generated on the surface and the particles fell away during passivation. On the other hand, a slime layer of Ag particles that adhered to the surface was generated during the dissolution of the Cu-5 wt pct Ag anode. The Cu-5 wt pct Ag anode was passivated with a lower current density than the pure Cu anode, which suggested that the morphology and adhesive characteristics of the slime on the anode affected the passivation. The direct observation method described herein is useful for understanding reactions on electrodes that undergo drastic changes in their surface morphology. Information obtained from this method can help with the development of new processes for the effective utilization of limited natural resources and energy, such as the recycling of Cu by electrorefining.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Synthesis and characterization of super occluded LiCl-KCl in zeolite-4A as a chloride salt waste form intermediate

Here, this paper reports the hygroscopic properties of eutectic LiCl-KCl after absorption into zeolite-4A, up to salt loadings of 75 wt%. Samples of the salt occluded zeolite were hydrated in a humidity chamber at constant temperature and relative humidity for up to 100 h. At up to 45 wt% salt loading, the un-occluded phase of salt consisted primarily of NaCl, which forms when the Na + ions present in the zeolite framework exchange with Li + and K + ions from the eutectic LiCl-KCl. This results in minimal water absorption and corrosion of contacted stainless steel. At greater than 45 wt% salt loading, water absorption and corrosion progressively worsened. The mixture has a significant amount of excess LiCl-KCl, making it highly hygroscopic. This study reveals an option for the intermediate treatment of waste salt from spent nuclear fuel electrorefiners that could facilitate it to be stored in a non-inert atmosphere for extended periods of time before final conversion into a permanent waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗