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

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Dehalogenation of electrochemical processing salt simulants with ammonium phosphates and immobilization of salt cations in an iron phosphate glass waste form

This study demonstrates the conversion of simulated pyroprocessing salt wastes [i.e., ER(SF) and ERV2] into chemically durable iron phosphate glass waste forms through reactions with ammonium hydrogen phosphate precursors [i.e., NH4H2PO4 and (NH4)2HPO4] and added Fe2O3. The reaction of the salt simulants with the phosphate precursors resulted in dehalogenation of the salt through the production of NH4Cl and NH4I (along with water), of which the NH4Cl can be used to produce UCl3 oxidant for use in the electrorefiner during pyroprocessing. These iron phosphate glasses contain a notably higher waste loading than the baseline glass-bonded sodalite ceramic waste form, which requires about 70 vol% sodalite to accommodate Cl from the salt, because they are dehalogenated prior to vitrification. The high waste loadings coupled with higher bulk densities (~3×103 kg/m3) than glass-bonded sodalite (~2.3×103 kg/m3) mean that these glasses have a higher storage volume, i.e., the volume required to immobilize a given mass of waste. The iron phosphate glass waste forms were characterized using a variety of techniques including physical property measurements and chemical durability testing. Following initial testing, a proof-of-concept demonstration was conducted to show that the NH4Cl generated from this process could be used to produce UCl3 using depleted uranium dendrites as a feedstock.

phosphate glass, pyroprocessing, waste forms, salt↗
DOE OSTI2021

Drying model of a high salt content cementitious waste form: Effect of capillary forces and salt solution

Highlights: • Drying model for a high salt content cementitious waste form is developed. • Water vapor diffusion and capillary liquid flow are distinguished. • Capillary and salt solution effects are considered in vapor-liquid equilibrium. A water transport model coupling capillary liquid flow with vapor diffusion is developed to describe the drying process for a cementitious waste form with high salinity porewater. Vapor-liquid equilibrium is formulated as the driving force for vapor diffusion and the model accounts for pore capillary and high salinity effects on water thermodynamic activity. Pore filling and porewater surface tension as a function of pore size distribution and water saturation have been quantified for the material. Geochemical speciation modeling is used to simulate porewater activity as a function of composition over the range of saturation. The theoretical relationship between relative humidity and water saturation generally agrees with experimental measurement, and the developed model is capable of predicting drying rates under various external relative humidity conditions. The model was developed to be incorporated into reactive transport models considering the effects of drying such as salt redistribution and efflorescence.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Re‐establishing the paradigm for evaluating halide salt compatibility to study commercial chloride salts at 600°C–800°C

Abstract Chloride salts are one candidate for a >700°C concentrating solar power (CSP) cycle, however, many reports from the literature suggest very high reaction rates between chloride salts and structural alloys. Historically, a specific methodology was established for evaluating halide salt compatibility based on solution kinetics. This study returned to that paradigm where the salts are purified and evaluated in sealed capsules before moving to a flowing experiment to determine a true corrosion rate in a temperature gradient for a commercial K–Mg–Na chloride salt. Isothermal testing focused on Ni‐based alloys 230 and 600 at 600°C–800°C. The results indicated there were promising combinations of salt chemistry, temperature, and alloy composition that reduce the extent of reaction. The results of the first monometallic thermal convection loop of alloy 600 run for 1,000 hr with a peak temperature of 700°C showed low attack with rates ≤9 µm/yr.

Pint, Bruce A.↗
DOE OSTI2018

Effects of Imide–Orthoborate Dual-Salt Mixtures in Organic Carbonate Electrolytes on the Stability of Lithium Metal Batteries

The effects of lithium imide and lithium orthoborate dual-salt electrolytes of different salt chemistries in carbonate solvents on the cycling stability of Li metal batteries were systematically and comparatively investigated. Two imide salts (LiTFSI and LiFSI) and two orthoborate salts (LiBOB and LiDFOB) were chosen for this study and compared with the conventional LiPF6 salt. The cycling stability of the Li metal cells with the electrolytes follows the order from good to poor as LiTFSI-LiBOB > LiTFSI-LiDFOB > LiPF6 > LiFSI-LiBOB > LiFSI-LiDFOB, indicating that LiTFSI behaves better than LiFSI and LiBOB over LiDFOB in these four dual-salt mixtures. The LiTFSI-LiBOB can effectively protect the Al substrate and form a more robust surface film on Li metal anode, while the LiFSI-LiBOB results in serious corrosion to the stainless steel cell case and a thicker and looser surface film on Li anode. Computational calculations indicate that the chemical and electrochemical stabilities also follow the order of LiTFSI-LiBOB > LiTFSI-LiDFOB > LiFSI-LiBOB > LiFSI-LiDFOB. The key findings of this work emphasize that the salt chemistry is critically important for enhancing the interfacial stability of Li metal anode and should be carefully manipulated in the development of high performance Li metal batteries.

Li metal batteries, dual-salt electrolyte, solid e↗
DOE OSTI2017

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

Riley, Brian J.↗
DOE OSTI2018

Sensitivity/uncertainty analyses comparing LR-0 reactor experiments containing FLiBe salt with models for molten-salt-cooled and molten-salt-fueled reactors

Here, critical experiments using an insertion zone with FLiBe salt performed at the LR-0 reactor at Research Centre Rež (RC Rež) have been compared to application models for solid-fueled, fluoride-salt-cooled high-temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts using sensitivity and uncertainty (S/U) analysis techniques. These experiments support FHR and MSR advanced reactor concepts by informing on neutron spectral effects and nuclear data uncertainties related to fluoride salts. The FLiBe salt in the LR-0 experiments is from the Oak Ridge National Laboratory Molten Salt Reactor Experiment and is enriched to greater than 99.99% 7 Li. This work is part of a broader collaboration between the United States and the Czech Republic on MSR and FHR technology. Results from the S/U analyses for FHR and MSR models indicated the most significant potential source of eigenvalue bias due to nuclear data within the FLiBe salt is radiative capture in 7 Li. Other smaller but potentially significant contributions come from 19 F, 6 Li, and other 7 Li reactions. Similarity comparisons of the RC Rež LR-0 experiments and FHR and MSR application models indicate that the LR-0 FLiBe experiments could be useful as candidate benchmarks for low-enriched uranium–fueled application models. However, sensitivity differences observed included both spectral effects, due to the LR-0 reactor being moderated by water instead of graphite and fluoride salt, and sensitivity magnitude effects driven by the amount of FLiBe inserted and its location. New experiments with refined designs to increase the volume and importance of salt in the system could provide improved data. Results also demonstrate that salt systems using thorium and/or 233 U fuel will require additional experiments with relevant driver fuels. Increased contributions from the graphite moderator and 19 F and covariance between reactions such as 233 U fission and 233 U radiative capture indicate higher uncertainty contributions from the salt and significant differences in contributions from the fuel species due to underlying uncertainties in their nuclear data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7 LiF + BeF 2 ) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in 7 Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2002

Plasma 24,25-dihydroxyvitamin D concentration of Dahl salt-sensitive rats decreases during high salt intake

Dahl salt-sensitive rats, but not salt-resistant rats, develop hypertension in response to high salt intake. We have previously shown an inverse relationship between plasma 25-hydroxyvitamin D (25-OHD) concentration and blood pressure of Dahl salt-sensitive rats during high salt intake. In this study, we report on the relationship between high salt intake and plasma 24,25-dihydroxyvitamin D (24,25-(OH)(2)D) concentration of Dahl salt-sensitive and salt-resistant rats. Rats were fed a high salt diet (8%) and sacrificed at day 2, 7, 14, 21, and 28. Plasma 24,25-(OH)(2)D concentrations of salt-sensitive rats were reduced to 50% of that at baseline at day 2-when blood pressure and plasma 25-OHD concentration were unchanged, but 25-OHD content in the kidney was 81% of that at baseline. Plasma 24,25-(OH)(2)D concentration was reduced further to 10% of that at baseline from day 7 to 14 of high salt intake, a reduction that was prevented in rats switched to a low salt (0.3%) diet at day 7. Exogenous 24,25-dihydroxycholecalciferol (24,25-(OH)(2)D(3)), administered at a level that increased plasma 24,25-(OH)(2)D concentration to five times normal, did not attenuate the salt-induced hypertension of salt-sensitive rats. Plasma 24,25-(OH)(2)D concentration of salt-resistant rats was gradually reduced to 50% of that at baseline at day 14 and returned to baseline value at day 28 of high salt intake. We conclude that the decrease in plasma 24,25-(OH)(2)D concentration in salt-sensitive rats during high salt intake is caused by decreased 25-OHD content in the kidney and also by another unidentified mechanism.

Non-NASA Center↗
DOE OSTI2025

The molten salt tritium transport experiment: A pumped fluoride salt loop for hydrogen isotope experimentation

Molten salt reactors (MSRs) and fusion reactors propose to use molten salts as coolants and breeder blanket materials, respectively. Tritium, however, poses safety concerns in both reactor types due to its ability to permeate through reactor materials and potential for environmental release. This manuscript addresses the tritium transport phenomena in molten salts and presents the design and analysis of the Molten Salt Tritium Transport Experiment (MSTTE). MSTTE is a forced-convection fluoride salt loop intended to measure hydrogen isotope permeation through structural materials in a flowing salt system. In the first phase, MSTTE will use FLiNaK salt and deuterium as surrogates for FLiBe and tritium, with future plans to utilize tritium and FLiBe. MSTTE couples a Copenhagen Atomics pumped salt loop with an external test section that introduces hydrogen isotopes into the loop and measures transport phenomena. The Hydrogen Injection System (HIS) controls hydrogen isotope introduction into the molten salt loop. Here, the permeation test section measures the permeation rate through stainless steel tubing in contact with flowing salt. Computational fluid dynamics (CFD) analysis ensures fully developed salt flow in the permeation test section. MSTTE is modeled with MELCOR-TMAP to predict the permeation rate as a function of experimental variables such as source term, salt flow rate, and salt temperature. Results indicate that the source term is the only parameter with a significant effect on the permeation rate. Pressure drop analysis suggests that the loop should operate below 200 LPM to maintain a pressure drop below 200 kPa. Additionally, finite-element analysis assesses thermal stress during loop operation to ensure the experiment's safe design. MSTTE will provide semi-integral data on tritium transport phenomena in molten salts and serve as a testbed for advancing molten salt technology.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Cryo-based Structural Characterization and Growth Model of Salt Film on Metal

Metal salt films play a critical role in high rate dissolution processes such as within an active pit. However, structural and compositional characterization of salt films is challenging owing to the dynamic nature of the dissolving interface. Here, the salt film formed on SS304 one-dimensional artificial pit surface was directly characterized using flash freezing and cryo-based focused ion beam/scanning electron microscopy. The salt film exhibits a porous structure. Ni is co-located with Fe in the salt film, indicating the formation of co-precipitated salt containing Fe2+ and Ni2+. Additionally, a salt film growth model is presented based on ion transport.

Salt film, Pitting corrosion, Cryogenic, Cryo-FIB/↗
DOE OSTI2025

Real-Time Characterization of Salt Aerosols Generated from Static and Sparged Molten Salt

The formation of radionuclide-bearing aerosols in the respirable size range has the potential to significantly influence offsite dose consequences and is, therefore, an important consideration in nuclear facility safety assessments. Molten salt reactor (MSR) developers will likely need to demonstrate an understanding of the conditions under which radionuclide-bearing aerosols may be generated from their reactor under normal operating and accident conditions, as well as the characteristics and transport behavior of these aerosols, to demonstrate to the U.S. Nuclear Regulatory Commission (NRC) that the facility can be operated safely. Recent reviews of the literature identified a lack of experimental data describing the mechanisms of formation and properties (size, concentration, and composition) of salt aerosol particles that are produced from molten salts. Experiments that identify the conditions that lead to radionuclide-bearing salt aerosol releases and quantify the characteristics of salt aerosols formed by different mechanisms are high-priority needs to support MSR licensing. This report describes tests that were conducted within the Argonne Salt Aerosol Test Stand (a sealed vessel and measurement system) to generate salt aerosols from static and sparged molten salts and measure their size and concentration in real-time. The results provide insight into salt aerosol formation by the vapor condensation and bubble bursting mechanisms and inform the potential radiological consequences of aerosol formation from molten fuel salt. Videos of the salt surface were taken during salt sparge tests to observe surface bubble behavior. The data in this report can be used to develop mechanistic source term and accident progression models for MSRs. The real-time salt aerosol characterization technique used in this study will be employed in future integral effects tests that are conducted at an engineering scale to simulate realistic MSR accidents and in future separate effects tests to address additional variables that may impact salt aerosol characteristics (e.g., presence of fission products in salt and humidity in atmosphere).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

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

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

36 - MATERIALS SCIENCE↗
DOE OSTI2017

Comparing Sensitivity/Uncertainty Analysis Results for LR-0 Salt Experiments with Salt Reactor Models

Recent reactor physics measurements conducted at Research Centre Rez (RC Rez) used the LR-0 low power nuclear reactor to investigate neutronic impacts and uncertainties of FLiNa (60 mol% LiF and 40 mol% NaF) and FLiBe (2 LiF + BeF{sub 2}) salt targets. This included using FLiBe originally from the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL), which contained enriched lithium (99.994 mol% {sup 7}Li). In collaboration with RC Rez, ORNL performed sensitivity/uncertainty (S/U) analyses of these LR-0 experiments with salt targets. Molten salt reactor technologies can be categorized into two general types: solid-fueled or liquid-fueled. Solid-fueled molten salt technologies are perhaps best represented by fluoride-salt-cooled high-temperature reactors (FHRs), which are solid-fueled and salt-cooled. They are potentially promising because they offer high-temperature outlet heat using a low-pressure, high-power-density system (relative to high temperature gas-cooled reactors). However, FHRs have never been built; therefore, no direct neutronic benchmarks exist for code or cross section validation work. Liquid-fueled molten salt reactors (MSRs) use a liquid salt as both fuel and coolant. MSRs can be designed to use a variety of fertile or fissile species as nuclear fuel, different neutron energy spectra, and a multitude of approaches for adding or removing material from the salt. The MSRE generated some operational data for fluoride salt MSRs, but no formal benchmarks currently exist based on the MSRE data, and the data available do not cover all operational conditions. Given these data limitations, ORNL performed S/U analyses using application models for FHR and MSR concepts to (1) identify potential sources of bias in FHR and MSR simulations resulting from cross section uncertainties and to (2) compare the sensitivity of neutron multiplication in FHR and MSR concepts to cross section data on an energy-dependent basis for specific nuclides against similar sensitivity results from LR-0 with a FLiBe salt target. In the future, these S/U analyses could inform the design of additional salt experiments or provide part of the basis for follow-up work to refine nuclear data. Please see the final project report for additional details beyond this paper.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Development and Demonstration of a Prototype Molten Salt Sampling System

Molten salt reactors (MSRs) offer potential operability and safety advantages when compared to commercial light water reactors (LWRs). However, operating experience with MSRs is sparse in comparison to what exists for LWRs. Further, the chemical and isotopic composition of the fuel and/or coolant salt is dynamic and difficult to characterize continuously, posing potential safety, operability, and safeguards unknowns that need to be addressed. A molten salt sampling system (MSSS) is regarded as a necessary subsystem within first generation MSRs used to obtain samples of salt for chemical and isotopic analysis in support of the need to monitor and control salt composition during operation. The MSSS is being developed using the Safety-in-Design (SiD) methodology, which incorporates incremental integration of safety analysis into the design process. The MSSS conceptual design emerging from the application of the early stages of the SiD methodology consists of a sample collection system and its housing, a freeze port, and inert gas control and delivery systems. This article describes the prototypes developed to test the functions of these MSSS subsystems, presents the results of testing in both dry and molten salt environments (including reliability data collection performed in accordance with the principles of SiD and the development of a semiquantitative fault tree model), and summarizes the opportunities for future design and testing enhancements based on the results of prototype testing.

molten salt reactor↗