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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 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Coupling coarse mesh CFD with fine mesh CFD for modeling for Molten Salt Reactors in the Virtual Test Bed VTB

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program aims at developing a simulation tool kit to accelerate the development and deployment of nuclear power technologies. NEAMS multiphysics tools have been designed to provide numerical simulation support for the design and licensing of GEN IV reactors. Pronghorn is NEAMS’s coarse-mesh computational fluid dynamics (CFD) tool, which is designed to run 3D core transients in GEN IV reactors at a reduced computational cost. To increase their accuracy, coarse-mesh CFD simulations require calibrated closure coefficients. One way of computing these coefficients is via the Nek5000, NEAMS’s high-fidelity CFD tool. This article discusses our current research lines in informing Pronghorn closure coefficients via Nek5000 to enable multiphysics simulations of the core cavity of the molten-salt fast reactor. We present an application in which Pronghorn mixing length turbulent viscosity has been calibrated from Nek5000 simulations. The resulting Pronghorn thermal-hydraulics model is then coupled to Griffin, the NEAMS neutron transport solver, to solve for the steady-state configuration of the molten-salt fast reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Online Molten Salt and Gas Monitoring Using Combined Spectroscopic Methods

The nonproliferation of nuclear material and the development of safeguards technologies to sample and analyze fuel salt is a challenge for molten salt reactor (MSR) developers. We seek online monitoring using combined spectroscopic methods of two distinct material phases from MSR’s. Molten salt transformed to aerosol phase allows for Light Induced Breakdown spectroscopy (LIBS) without adverse splashing (ARPA-E). Vaporized used nuclear fuel through chlorine-based volatility (CBV) allows for uranium separation based on boiling point differences

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Corrosion susceptibility and chromium loss in Austenitic steels and Nickel-based alloys in molten FLiNaK at 700 °C

A comparative study was conducted to evaluate the corrosion susceptibility of 316L, 316H, Ni 200, Inconel 625, and Hastelloy N in molten FLiNaK at 700 °C for 100 h. Top-view and cross-sectional scanning electron microscopy (SEM) imaging, combined with energy-dispersive X-ray spectroscopy (EDS) mapping, was performed to investigate microstructural and compositional changes. SEM images were further processed by introducing a contrast threshold to map cavity distribution. Using EDS mapping, intergranular and intragranular Cr loss were separately characterized. The ranking of mass loss after corrosion, from highest to lowest, is as follows: 316H > 316L > Inconel 625 > Hastelloy N > Ni 200. Cr loss was found to be correlated with Mo concentration, in agreement with many previous studies, with higher Mo content resulting in reduced Cr loss. On the other hand, there is no evidence that the high carbon content in 316H enhances corrosion resistance. The relatively low solubility of carbon in austenite at the testing temperature limits the amount of dissolved carbon. Therefore, the carbon-retarded vacancy diffusion, and consequently the reduced Cr diffusion, were not observed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Path Toward a Breeding, Proliferation-Resistant, Thermal-Spectrum Molten Salt Reactor

Liquid-fueled, thermal-spectrum molten salt breeder reactors (TS-MSBRs) offer the potential for affordable, safe, inexhaustible energy with minimal potential for nuclear material misuse and without significant actinide waste generation. Realizing the full set of TS-MSBR capabilities is only now becoming possible with the advent of advanced fuel-salt processing techniques, improved materials, and a more detailed understanding of fuel-salt properties. Additionally, modern higher-fidelity modeling and simulation methods enable a more detailed evaluation of TS-MSBR design options. TS-MSBRs, however, remain immature and will require substantial, sustained development resources.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Monitoring Noble Gases (Xe and Kr) and Aerosols (Cs and Rb) in a Molten Salt Reactor Surrogate Off-Gas Stream Using Laser-Induced Breakdown Spectroscopy (LIBS)

In this study with surrogate materials we show that laser-induced breakdown spectroscopy (LIBS) is a robust tool with promising capability toward monitoring gaseous (Xe and Kr) and aerosol (Cs and Rb) species in an off-gas stream from a molten salt reactor (MSR). MSRs will continually evolve fission products into the cover gas flowing across the reactor headspace. The cover gas entrains Xe and Kr gases, along with aerosol particles, before passing into an off-gas treatment system. Univariate models of Xe and Kr peaks showed a strong correlation to concentration indicated by their coefficients of determination of 0.983 and 0.997, respectively. Multivariate models were built for all four analytes using partial least squares regression coupled with preprocessing steps including normalization, trimming, and/or genetic algorithm derived filters. The models were evaluated by predicting the concentrations of the analytes in four validation samples, in which all calibration models were successfully validated at a confidence interval of 99.9%. Finally, pressure controllers were used to regulate the mass flow rate of Kr flowing into the measurement cell in sinusoidal and stepwise waveforms to test the real-time monitoring capabilities of the regression models. Both univariate and partial least squares Kr models were able to successfully quantify the gas concentration in the real-time evaluation. The root mean squared error of prediction (RMSEP) values for these real-time tests were calculated to be 0.051, 0.060, and 0.121 mol% demonstrating the measurement systems’ capability to perform online monitoring with acceptable accuracy.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Concept for an irradiation experiment to test a laser-induced breakdown spectroscopy off-gas sensor for molten salt systems

This paper focuses on laser-induced breakdown spectroscopy (LIBS) as an off-gas sensor and details a conceptual irradiation experiment to evaluate its performance in an environment that is more realistic of an operational molten salt reactor (MSR). MSRs are a promising advanced nuclear design that use high-temperature liquid salts as the coolant, fuel, or a combination of both. They will generate and subsequently evolve fission products into the reactor headspace during operation. These evolved fission products will necessitate an online off-gas system to treat the reactor cover gas before recycling it to the core. Such a system, especially during the early deployment phase, will benefit from online sensors to inform operators of the treatment system’s performance. Optical spectroscopy is well-suited to fit this task as it can be remotely deployed using fiber-optics and can detect most species at high frequencies. LIBS can detect nearly every element on the Periodic Table, stable or radioactive, making it an ideal candidate. The provided conceptual experiment discusses the facility requirements, salt capsule design, and detailed testing campaigns with corresponding simulated LIBS spectra.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

The effect of interfacial phenomena on gas solubility measurements in molten salts

The behavior of fission gases in molten fuel salt reactors governs activity transport from the reactor and can also affect the performance of the reactor itself. The gas solubility can be described thermodynamically by Henry’s law. However, the coupling of the condensed and gas phases depends on the interfacial area, which is difficult to measure or even to estimate. Surfaces of materials in the reactor will include disperse phases in the salt and porosity within the structural materials, covering a range of compositions and sizes. These attributes can affect measurements of fundamental properties such as gas solubility. Methods to obtain gas solubility, surface tension, interfacial energies, and bubble gas transport are reviewed. Recent data from manometric experiments are interpreted based on xenon sorption onto salt-wetted quartz.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Molten Chloride Reactor Experiment

MCRE will be the first fast spectrum, molten chloride reactor in operation. Small critical experiment to provide data for future reactors. - Collaboration between Southern Company Services, TerraPower, and Idaho National Laboratory (INL), as well as other partners. - Expected to be built and operated at the LOTUS testbed (ZPPR). - Authorized by the U.S. Department of Energy. - Technology development of salt synthesis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Molten Salts-Driven Discovery of a Polar Mixed-Anion 3D Framework at the Nanoscale: Zn 4 Si 2 O 7 Cl 2 , Charge Transport and Photoelectrocatalytic Water Splitting

Mixed-anion compounds widen the chemical space of attainable materials compared to single anionic compounds, but the exploration of their structural diversity is limited by common synthetic paths. Especially, oxychlorides rely mainly on layered structures, which suffer from low stability during photo(electro)catalytic processes. Herein we report a strategy to design a new polar 3D tetrahedral framework with composition Zn 4 Si 2 O 7 Cl 2 . We use a molten salt medium to enable low temperature crystallization of nanowires of this new compound, by relying on tetrahedral building units present in the melt to build the connectivity of the oxychloride. These units are combined with silicon-based connectors from a non-oxidic Zintl phase to enable precise tuning of the oxygen content. This structure brings high chemical and thermal stability, as well as strongly anisotropic hole mobility along the polar axis. These features, associated with the ability to adjust the transport properties by doping, enable to tune water splitting properties for photoelectrocatalytic H 2 evolution and water oxidation. This work then paves the way to a new family of mixed-anion solids

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Reciprocal Ternary Molten Salts Enable the Direct Upcycling of Spent Lithium‐Nickel‐Manganese‐Cobalt Oxide (NMC) Mixtures to Make NMC 622

Cathode active material is the most valuable component of spent lithium‐ion batteries, accounting for ≈30% of their overall value. Direct recycling of cathode materials involves recovering, regenerating, and reusing them without breaking down their chemical structure. This approach maximizes the added value of the cathode compound and reduces manufacturing costs by avoiding the need for virgin material production. However, one key challenge in scaling direct recycling from lab to industry is the requirement for highly purified cathode materials, contrasting with the low purity of black mass generated from battery shredding. No efficient separation process currently exists to isolate different lithium‐nickel‐manganese‐cobalt oxides (NMCs) from each other. Thus, direct recycling technologies that can operate with mixtures of multiple NMC stoichiometries will be best‐suited for industrial adoption. This study explores the direct recycling of NMC mixtures into NMC 622 using a “reciprocal ternary molten salts (RTMS)” system. Ionothermal relithiation and upcycling within the RTMS system successfully restore the layered structure, lithium content, and electrochemical performance of degraded NMCs, yielding results comparable to pristine NMC 622 (P‐NMC 622).

25 ENERGY STORAGE↗
DOE OSTI2026

Quantifying Uncertainties in Heat Capacity Measurements of Molten Salts Determined Using Differential Scanning Calorimetry

Uncertainty in specific heat capacity values of a molten salt determined by using differential scanning calorimetry (DSC) was assessed based on the precision of replicate measurements of heat flows used in the calculation and effects of corrections that are commonly made to heat flow measurements. The ratio method of determining heat capacity was applied using the results of replicate heat flow measurements made with two empty cells, a sapphire reference material, and three samples of a doped NaCl-UCl 3 salt mixture. Replicate measurements with empty cells were used to quantify the effects of system instabilities and sensitivities on the measured heat flows of sapphire and salt. The combined effects of uncertainties in individual heat flow measurements made with blank cells using this system were quantified to be 2.6 μV based on isothermal holds before and after the scan, with cell placement adding the greatest uncertainty. This value was used as the tolerance for accepting background-corrected heat flows measured with sapphire and salt to calculate the specific heat capacity. The acceptable heat flows measured for sapphire and salt over the temperature range of 540 to 725 °C resulted in calculated specific heat capacity values ranging from 0.53 to 0.91 J g −1 K −1 with an overall average value of 0.70 J g −1 K −1 and an uncertainty of 0.22 J g −1 K −1 at the 99 % confidence level. The combined uncertainty in the specific heat capacity masked detection of any effect of temperature or salt composition that occurred.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Steady-state radiochemical transport model of the molten salt reactor experiment

A radiochemical transport model of the Molten Salt Reactor Experiment (MSRE) was formulated based on the material derivative featuring system level 1-D flow providing spatial dependence accounting for multiple phenomena including radioactive decay, chemical removal, and fuel depletion across custom volume regions. Further, the steady-state version, which is based on a system of ODE's, can be solved efficiently and is useful for sensitivity studies utilizing Monte Carlo sampling. Accounting for multiple phenomena in a fully coupled model is necessary to understand the complexities of transport and distribution of radionuclides throughout the primary loop of a flowing fuel reactor.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Radioxenon signatures of molten salt reactors

Developments of molten salt reactor (MSR) technologies are making rapid progress across the globe. These reactor designs involve properly controlling radionuclides through off-gas systems. This work examines radioxenon emissions and assesses treaty detection technology to be used for monitoring. This work also includes a sensitivity study with ORIGEN on the nuclear forensic signals possible from a MSR. Multiple Isotope Ratio Comparison (MIRC) plots are used to compare the ratios of radioxenon isotopes in an MSR to a highly enriched uranium (HEU) reactor. Finally, the results show that MSR emissions may significantly overlap with signals produced in a HEU reactor and that changing reactor operations or retaining emissions in an off-gas system can significantly shift the radioxenon signature away from the HEU pulse.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
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↗