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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Electroanalytical Measurements of Oxide Ions in Molten CaCl 2 on W electrode

The electrochemical interaction of oxide ions with tungsten electrodes in molten calcium chloride (CaCl 2 ) was analyzed by combining electroanalytical techniques with X-ray photoelectron spectroscopy. During a cyclic voltammetry (CV) scan, the oxide ions appear to interact with the tungsten working electrode via a multi-step oxidation and reduction process. The overall redox peaks behave reversibly up to 750 mV s -1 . This electrochemical process enables the oxide ion concentration to be correlated to CV oxidation peaks. Lastly, the resulting correlation agrees well (8.8% difference) with back titration measurements and can be used to monitor oxide content in the salt in real-time during CaCl 2 based pryoprocesses.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Effect of iodides on thermal behavior and phase partitioning in LiCl-KCl

Liquid-fueled molten salt reactors (MSRs) are designed to operate with fissile materials and, ultimately, fission products dissolved in the primary molten salt coolant. Understanding the speciation and transport of iodine—a high-yield fission product—is essential because this element’s accidental release poses significant environmental concerns due to its capacity to be readily absorbed by the human thyroid gland. Here, we report the impact of iodide species (LiI and KI) on phase transitions, phase distribution, and phase stability in LiCl–KCl-eutectic salt mixtures. The study employed a combination of computational and experimental techniques, including thermodynamic FactSage calculations, differential scanning calorimetry, and high-temperature X-ray diffraction. The results indicate that the presence of iodide (10–25 wt%) significantly alters the melting behavior of the LiCl–KCleutectic system. Adding 10 wt% LiI has a more-pronounced effect than 10 wt% KI, as LiI converts to KI, leading to formation of LiCl, thereby, altering the LiCl-KCl ratio which significantly affects the melting temperature of the mixture. Furthermore, the evolution of crystalline structure, solid-fraction composition, and the dynamics of mixed-halide solid–liquid partitioning as a function of temperature indicate the potential for selective iodide separation from chloride-salt mixtures via solid–liquid separation techniques. Overall, the presented findings provide valuable insights that are beneficial for the design and operation of MSRs, as well as for the safe handling and effective processing of used nuclear fuel using advanced pyrochemical techniques.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Progression to Compatibility Evaluations in Flowing Molten Salts

Molten salt compatibility with structural alloys has been identified as a key issue for the development of Generation 3 concentrating solar power (CSP) systems with thermal storage. To accelerate this evaluation to pumped systems, the goal of this project was to conduct thermal convection loop (TCL) experiments with a peak temperature of ≥700°C and a typical temperature gradient of ~100°C. The experiments indicated that conventional ~16wt.%Cr Ni-based alloys are compatible up to 700°C with purified (i.e. low O) or dried (low H₂O) industrial-sourced Mg-K-Na chloride salt with <10 μm/yr loss. This two-year project was conducted based on the experimental and mechanistic understanding developed more than 60 years ago at Oak Ridge National Laboratory (ORNL). The first TCL experiment met the <15 μm/yr corrosion metric for this project with specimens of Ni-based alloy 600 exposed at 580°-700°C for 1000 h and post-exposure room temperature tensile tests showed minimal degradation. The second TCL experiment successfully deployed an electrochemical sensor from Argonne National Laboratory and had a peak temperature of 750°C but only ran for ~110 h due to a furnace failure. Both experiments used highly purified industrial-sourced salt with an O content of ~3 μg O/g salt and a Mg addition of 0.04 wt.%. In the second year, the Chloride Collective developed a more economical drying procedure such that the O content was much higher (>20,000 μg O/g salt). A third TCL experiment was conducted with a sensor and a peak temperature of 700°C using dried salt from the same industrial source and increased NaCl content (~20 wt.%). In addition to a 0.05% Mg addition to the salt, a Mg coupon was added in the coldest part of the loop which dissolved during the experiment. Again, small mass changes were noted for specimens of alloys 600 and C276 but the values were slightly higher than those measured in purified salt with Mg. A thin non-continuous and non-adherent oxide layer was deposited on most specimens containing Mg, Si and Al but Cr depletion also was observed. Both years included facilities qualification crucible experiments and then capsule experiments to confirm a baseline isothermal reaction rate. The first year capsule experiment led to the conclusion that the two-stage ORNL purification process using NH₄Cl and CCl₄ left the salt with a high Cl potential and a Mg addition (~0.05wt.%) was needed to lower the potential. The second year capsule experiments at 600° and 700°C found little difference in depth of attack for 0-0.25%Mg additions. In general, the complex reactions where salt can be trapped in the porous surface layer of metal indicated that mass change is an unreliable metric and average depth of Cr depletion is a better metric for assessing the extent of attack. These results have created a new baseline that is contrary to the recent published literature for chloride salts where mass losses have been reported that can be extrapolated to very significant annual metal loss rates. Chloride salt corrosion can be controlled and the results also indicate that salt purification to low O levels may not be necessary. However, additional TCL experiments are needed at different times and temperatures to generate important engineering information such as temperature dependent corrosion rates and reaction rate laws for extrapolation to long-term behavior and isolate the effect of salt additives and impurities.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Direct heat exchanger for molten chloride fast reactor

A molten chloride fast reactor (MCFR) includes a plurality of reflectors defining a central core having a core geometric center. A flow channel fluidically connected to the central core. The flow channel includes an outlet flow channel downstream of the central core and an inlet flow channel upstream from the central core. A primary heat exchanger (PHX) disposed outside the central core and between the outlet flow channel and the inlet flow channel. The MCFR also includes a decay heat heat exchanger (DHHX). At least a portion of the DHHX is disposed above the core geometric center, and a fuel salt is configured to circulate at least partially through the outlet flow channel, the DHHX, the PHX, the inlet flow channel, and the central core.

Abbott, Ryan↗
DOE OSTI2021

Unraveling Local Structure of Molten Salts via X-ray Scattering, Raman Spectroscopy, and Ab Initio Molecular Dynamics

In this work, we resolve a long-standing issue concerning the local structure of molten MgCl 2 by employing a multimodal approach, including X-ray scattering and Raman spectroscopy, along with the theoretical modeling of the experimental spectra based on ab initio molecular dynamics (AIMD) simulations utilizing several density functional theory (DFT) methods. We demonstrate the reliability of AIMD simulations in achieving excellent agreement between the experimental and simulated spectra for MgCl 2 and 50 mol % MgCl 2 + 50 mol % KCl, and ZnCl 2 , thus allowing structural insights not directly available from experiment alone. A thorough computational analysis using five DFT methods provides a convergent view that octahedrally coordinated magnesium in pure MgCl 2 upon melting preferentially coordinates with five chloride anions to form distorted square pyramidal polyhedra that are connected via corners and to a lesser degree via edges. This is contrasted with the results for ZnCl 2 , which does not change its tetrahedral coordination on melting. Although the five-coordinate MgCl 5 3– complex was not considered in the early literature, together with an increasing tendency to form a tetrahedrally coordinated complex with decreasing the MgCl 2 content in the mixture with alkali metal chloride systems, current work reconciles the results of most previous seemingly contradictory experimental studies.

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

Determination and Comparison of Parameters from Cyclic Voltammetry of LaCl 3 in Eutectic LiCl-KCl

Electrochemical measurements of LaCl 3 were obtained in eutectic LiCl-KCl at 773 K. Cyclic voltammetry data were acquired using a tungsten electrode LaCl 3 amounts in the molten salt in the range of 0.5 to 3.0 wt%. Both cyclic voltammetry and Tafel analyses of the data resulted in calculation of values for the charge transfer coefficient (0.262), diffusion coefficient (1.67 × 10 -5 cm 2 s -1 ), exchange current density (0.0148 to 0.0473 A cm -2 ), and charge transfer resistance (0.404 to 1.99 Ω) related to lanthanum ions in the molten salt. Calculated values were compared to those available in the literature. The values of parameters found in literature varied significantly in comparison to those in the present study due to dissimilar experimental conditions, e.g., electrode material, temperature, and LaCl 3 concentration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Parametric Study of Used Nuclear Oxide Fuel Constituent Dissolution in Molten LiCl-KCl-UCl 3

Prior work identified dissolution of used nuclear oxide fuel constituents from a uranium oxide matrix into molten LiCl-KCl-UCl 3 at 500°C, prompting a subsequent series of three progressive studies (including an initial scoping study, an electrolytic dissolution study, and a chemical-seeded dissolution study) to further investigate associated parameters and mechanisms. Thermodynamic calculations were performed to identify possible reaction mechanisms and their propensities in used oxide fuel constituent dissolution. Used nuclear oxide fuels with varying preconditions from fast and thermal test reactors were separately immersed in the subject salt system to assess fuel constituent migration from the bulk fuel matrix to the salt phase in an initial scoping study. Dissolution of expected fuel constituents, including alkali, alkaline earth, lanthanide, and transuranium oxides, into the chloride salt phase varied widely, ranging from 12% to 99% in the initial study. Uranium isotope blending between the salt phase and bulk fuel matrix was also observed, which was attributed to reducing conditions in the fuel matrix. Electrolytic and chemical-seeded dissolution studies were subsequently performed to effect reducing conditions in the fuel. Other parameters, including temperature (at 500°C, 650°C, 725°C, and 800°C) and uranium trichloride concentrations (at 6, 9, and 19 wt% uranium), were investigated in the latter two studies, resulting in fuel constituent dissolution above 90%. Extents of dissolution were based on initial and final fuel constituent concentrations in the oxide fuels following operations in the salt and subsequent removal of the salt via distillation. Finally, in this series of progressive studies, oxide fuel preconditioning and in situ reducing conditions, along with elevated temperature and uranium trichloride concentrations, were the primary parameters promoting used nuclear oxide fuel constituent dissolution in accordance with identified reaction mechanisms.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Controlled Conversion of Sodium Metal From Nuclear Systems to Sodium Chloride

A series of three bench-scale experiments was performed to investigate the conversion of sodium metal to sodium chloride via reactions with non-metal and metal chlorides. Specifically, batches of molten sodium metal were separately contacted with ammonium chloride and ferrous chloride to form sodium chloride in both cases along with iron in the latter case. Additional ferrous chloride was added to two of the three batches to form low melting point consolidated mixtures of sodium chloride and ferrous chloride, whereas consolidation of a sodium-chloride product was performed in a separate batch. Samples of the products were characterized via X-ray diffraction to identify attendant compounds. The reaction of sodium metal with metered ammonium chloride particulate feeds proceeded without reaction excursions and produced pure colorless sodium chloride. The reaction of sodium metal with ferrous chloride yielded occasional reaction excursions as evidenced by temperature spikes and fuming ferrous chloride, producing a dark salt-metal mixture. This investigation into a method for controlled conversion of sodium metal to sodium chloride is particularly applicable to sodium containing elevated levels of radioactivity—including bond sodium from nuclear fuels—in remote-handled inert-atmosphere environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development and Assessment of Deployed Sensors and Technologies Supporting Molten Salt Loop Operations

During FY24, Argonne conducted a variety of activities to develop and assess new monitoring and control technologies towards enabling long-duration operations of molten salt reactor systems. The first objective of the work completed this year was to deploy technologies capable of reliably and rapidly monitoring the operational health of a molten salt loop. To achieve this, the project focused on specific tasks, including: (1) electroanalytical technique development for use in deployed sensors, and (2) operations of sensors on the Facility to Alleviate Salt Technology Risks (FASTR) loop at Oak Ridge National Laboratory. The second objective of the work this year focused on creating a pumped actinide flow loop to enable long-duration salt chemistry and corrosion studies using uranium-bearing fuel salts. Toward that end, we designed and procured a loop capable of being installed into a radiological glovebox at Argonne. A corrosion control system was also designed for integration into this loop. These combined systems will enable the feasibility of corrosion control and management systems to be investigated with complex chloride fuel salt mixtures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

A Holistic Approach for Elucidating Local Structure, Dynamics, and Speciation in Molten Salts with High Structural Disorder

To examine ion solvation, exchange, and speciation for minority components in molten salts (MS) typically found as corrosion products, we propose a multimodal approach combining extended X-ray absorption fine structure (EXAFS) spectroscopy, optical spectroscopy, ab initio molecular dynamics (AIMD) simulations, and rate theory of ion exchange. Going beyond conventional EXAFS analysis, our method can accurately quantify populations of different coordination states of ions with highly disordered coordination environments via linear combination fitting of the EXAFS spectra of these coordination states computed from AIMD to the experimental EXAFS spectrum. In a case study of dilute Ni(II) dissolved in the ZnCl 2 +KCl melts, our method reveals heterogeneous distributions of coordination states of Ni(II) that are sensitive to variations in temperature and melt composition. These results are fully explained by the difference in the chloride exchange dynamics at varied temperatures and melt compositions. Finally, this insight will enable a better understanding and control of ion solubility and transport in MS.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Technoeconomic Cost Analysis of NREL Concentrating Solar Power Gen3 Liquid Pathway: Preprint

The National Renewable Energy Laboratory is leading the liquid (molten salt) power tower pathway for the U.S. Department of Energy's concentrating solar power Gen3 initiative. The Gen3 liquid pathway required updated designs to three major components: the tower and receiver, the thermal energy storage (TES) tanks, and the power cycle. We assume a 100 MWe net system output and used the System Advisor Model (SAM) to complete a technoeconomic cost analysis of the Gen3 liquid pathway design and estimate its LCOE. This paper summarizes the methodology and results of that analysis.

chloride salt↗
DOE OSTI2022

Technoeconomic Cost Analysis of NREL Concentrating Solar Power Gen3 Liquid Pathway

The National Renewable Energy Laboratory is leading the liquid (molten salt) power tower pathway for the U.S. Department of Energy's concentrating solar power Gen3 initiative. The Gen3 liquid pathway required updated designs to three major components: the tower and receiver, the thermal energy storage tanks, and the power cycle. We assume a 100 MWe net system output and used the System Advisor Model (SAM) to complete a technoeconomic cost analysis of the Gen3 liquid pathway design and estimate its levelized cost of electricity. This paper summarizes the methodology and results of that analysis.

chloride salt↗
DOE OSTI2020

Voltammetry measurements in lithium chloride-lithium oxide (LiCl–Li 2 O) salt: An evaluation of working electrode materials

Instrumentation to provide process monitoring (PM) and safeguards for the oxide reduction (OR) step in the electrochemical processing of used oxide nuclear fuel is necessary to ensure equipment is operated as declared. Cyclic voltammetry (CV) has been proposed for real-time monitoring of the operation of an OR process for safeguards purposes. In this study, different materials including 316 stainless steel (SS), tantalum (Ta), molybdenum (Mo), tungsten (W), platinum (Pt), and iridium (Ir) were evaluated as potential working electrode (WE) materials based on their chemical inertness, corrosion resistance, and sensitivity in detecting Li 2 O and other chloride salts in the OR electrolyte. Of the electrodes evaluated, 316 SS, Ir, and Pt all performed reasonably well in the LiCl-Li 2 O electrolyte. Here, stainless steel was operated in the cathodic potential and had reasonable corrosion resistance and is relatively inexpensive. Iridium could be operated in both the cathodic and anodic potentials and was the most corrosion resistant of those evaluated. Platinum is limited to the anodic potential range, had reasonable corrosion resistance, and was the most sensitive to Li 2 O concentrations in the salt. In the development of a stand-alone safeguards instrumentation, our recommendation is a CV sensor fitted with both Ir and Pt WEs. The results in this study may also be helpful for anode material selection in other electrometallurgy industries, such as molten salt electrolysis.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
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 OSTI2022

Techno-Economic Analysis of Recuperated Joule-Brayton Pumped Thermal Energy Storage

This article describes a techno-economic model for pumped thermal energy storage systems based on recuperated Joule-Brayton cycles and two-tank liquid storage. Models have been developed for each component, with particular emphasis on the heat exchangers. Economic metrics such as the power and energy capital costs (i.e., per-kW and per-kWh capacity) and levelized cost of storage are evaluated by gathering numerous cost correlations from the literature, thereby enabling estimates of uncertainty. It is found that the use of heat exchangers with effectivenesses up to 0.95 is economically worthwhile, but higher values lead to rapidly escalating component size and system cost. Several hot storage fluids are considered; those operating at the highest temperatures (chloride salts) improve the round-trip efficiency but the benefit is marginal and may not warrant the additional material costs and risk when compared to lower-temperature nitrate salts. Cost-efficiency trade-offs are explored using a multi-objective optimization algorithm, yielding optimal designs with round-trip efficiencies in the range 59-72% and corresponding levelized storage costs of 0.12 0.03 and 0.38 0.10 $/kWhe. Lifetime costs are competitive with lithium-ion batteries for discharging durations greater than 6 h under current scenarios.

Carnot battery↗
DOE OSTI2025

Validation Testing for Molten Chloride Reactor Experiment Equipment Removal and Disposal Techniques

The Molten Chloride Reactor Experiment (MCRE) will be the first reactor featuring a fast-spectrum molten chloride circulating nuclear fuel system in the world. Planning for equipment removal and disposal (ERD) of MCRE has identified several technology gaps due to the unique environment of this nuclear experiment. Some of the gaps arise from the application of existing disassembly and/or sizing methods to novel material forms or in novel configurations. Others arise from unknown material behavior. This paper summarizes proposed test plans for ERD validation experiments to address these complicated or unknown equipment removal procedures. At the Waste Management Symposia in 2024, the Idaho National Laboratory (INL) MCRE ERD team presented the challenges associated with hosting multiple nuclear experiments in series with only brief transition periods between systems. Such difficulties include higher dose rates, the presence of radioisotopes infrequently encountered in reactor decommissioning and radioactive waste management, lack of intrinsic remote-operations infrastructure in the test bed, space constraints in the test bed, and contamination minimization requirements. To address these challenges, remote or semi-remote technologies are planned to be implemented in a non-hot cell environment with limited space availability. The team also discussed how a systems engineering approach is being used for conceptual development and design of equipment removal systems to address these challenges. For example, to reduce constraints for the removal of more difficult components, non-activated, noncontaminated elements are planned to be taken out first where possible. Still, there are complexities associated with the remaining components. In this work, the operational framework for MCRE ERD was reviewed for technical gaps and open questions, and test plans were drafted to address these areas. The tests plans were written for the following categories: vision systems, pipe cutting, drill/grout/filler, flush salt, and miscellaneous, with the miscellaneous group consisting of tests like techniques for removing bearings and reflector bricks. The test plans explore material, infrastructure, and staffing requirements needed for test execution. The test plans additionally focus on the evaluation of success. Determining the outcome of a test is imperative - as these explorative actions have the potential to rearrange or re-scope planned ERD activities. Success criteria identified thus far include required tool output, required area(s), debris production and mitigation, and repeatability. Test plans are an essential aspect of the systems engineering approach to MCRE ERD. They are used as the beginning steps in defining use cases for the ERD system. Performance of the validation tests is expected to begin in the summer of 2025 and will take approximately 9 to 12 months to complete. Execution of these plans will be expedited by specifying test needs ahead of time, facilitating efficient interactions with any subcontractors tasked with running the requested tests. Evaluating the outcomes of these tests will inform MCRE ERD procedures and timing and will also identify additional technical constraints for the MCRE ERD System. This upfront process optimization effort will help the project save time and resources at the end of the experiment.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

Powdermet CRADA for VELOCITI Voucher (CRADA Final Report)

NLR shall develop a model of an electric thermal energy storage (ETES) system in the System Advisor Model (SAM) framework that uses a salt and particle slurry developed by PowderMet as the thermal energy storage medium. The ETES shall consist of a recuperated sCO 2 Brayton cycle using 2-tank indirect thermal energy storage and an electric heater for heat input. The salt slurry uses chloride salt and operates at a maximum temperature of 500-720°C. The ETES uses dry air cooling. The salt slurry will be modeled as a sensible heat storage medium and as a latent heat storage medium. NLR will use the model to do parametric analysis of the factors affecting the performance (ex., round trip efficiency) of the ETES system. NLR will work with PowderMet to estimate the costs of system components and total system costs, with the levelized cost of electricity (LCOE) as the primary cost metric.

14 SOLAR ENERGY↗
DOE OSTI2021

DESIGN AND PROTOYPING OF A FISSILE-BEARING CHLORIDE SALT IRRADIATION EXPERIMENT

Limited data exists on the properties of irradiated fueled chloride-based salts for use in advanced nuclear reactors. The design and prototyping of a salt irradiation experiment is underway. Mechanical design will integrate the experiment into existing assemblies in TRIGA type cores. Neutronics modeling demonstrated how targeted power densities can be reached and provide insight on radionuclide source term generation within the experiment. Thermal-hydraulic and computational fluid dynamics (CFD) analysis indicated that temperature ranges in the salt can satisfy both safety and programmatic requirements. Initial prototyping tests provided confidence in the ability to extract the salt post-irradiation, and in the heater performance. Future planned prototyping will provide an integral assessment of the proposed design prior to insertion in the reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗