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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

A Boiling-Potassium Fluoride Reactor for an Artificial-Gravity NEP Vehicle

Several years ago a rotating manned spacecraft employing nuclear-electric propulsion was examined for Mars exploration. The reactor and its power conversion system essentially served as the counter-mass to an inflatable manned module. A solid-core boiling potassium reactor based on the MPRE concept of the 1960s was baselined in that study. This paper proposes the use of a liquid-fluoride reactor, employing direct boiling of potassium in the core, as a means to overcome some of the residual issues with the MPRE reactor concept. Several other improvements to the rotating Mars vehicle are proposed as well, such as Canfield joints to enable the electric engines to track the inertial thrust vector during rotation, and innovative "cold-ion" engine technologies to improve engine performance.

Sorensen, Kirk↗

Study of actinide chemistry in saturated potassium fluoride solution

Study concerning the chemistry of actinides in saturated KF solution included work with neptunium, uranium, and americium. Solubilities, absorption spectra, oxidation-reduction reactions, and solid compounds which can be produced in KF solution were examined. The information is used for preparation of various materials from salts of the actinides.

Cohen, D.↗

Molten salt compositions with enhanced heat transfer and reduced corrosion properties

A heat transfer (exchange) composition comprising a halide salt matrix having dispersed therein nanoparticles comprising elemental carbon in the absence of water and surfactants, wherein said halide is fluoride or chloride, wherein the halide salt may be an alkali halide salt (e.g., lithium fluoride, sodium fluoride, potassium fluoride, rubidium fluoride, sodium chloride, potassium chloride, rubidium chloride, and eutectic mixtures thereof) or an alkaline earth halide salt (e.g., fluoride or chloride salt of beryllium, magnesium, calcium, strontium, or barium), and wherein the nanoparticles comprising elemental carbon may be solid or hollow, and wherein the composition may further include nanoparticles comprising a fissile material (e.g., U, Th, or Pu) dispersed within the composition. Molten salt reactors (MSRs) containing these heat transfer compositions in coolant loops in thermal exchange with a reactor core, as well operation of such MSRs, are also described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

High Yield Exfoliation of a Sub-Micron Hexagonal Boron Nitride Using a Solvent Free Method

Sodium fluoride (NaF) has been found to be an activation agent in exfoliating a hexagonal boron nitride (hBN) platelet, using solvent free method. The procedure was carried out by enhancing the intercalation of ferric chloride (FeCl3) in an intermediate step, up to 330 °C, followed by removal of the by-product at room temperature. In this research, five additional common salts including lithium fluoride, potassium fluoride, lithium chloride, sodium chloride, and potassium chloride, were examined and compared to NaF for their effectiveness in producing the most exfoliation of a commercially available sub-micron-size hBN. It was found that, all fluoride and chloride salts significantly activated the intercalation of FeCl3 into this hBN. Analyses and characterization from X-ray diffraction, scanning electron microscopy, Fourier transform infrared spectroscopy and Raman spectroscoy revealed NaCl to be one of most effective agents without further heat treatment and the least expensive salt in the intercalating reaction and exfoliation process. The synthetic route of this solvent free method is straightforward with 89% yield and above which can be produced on a large scale.

boron nitride↗

Kinetic Parameters and Diffusivity of Uranium in FLiNaK and ClLiK

Anodic stripping voltammetry (ASV) and cyclic voltammetry (CV) measurements at 773, 823, and 873 K were made of uranium trifluoride (UF 3 ) in lithium fluoride-sodium fluoride-potassium fluoride eutectic (FLiNaK) and uranium trichloride (UCl 3 ) in lithium chloride-potassium chloride eutectic (ClLiK). ASV data were used to estimate the charge transfer coefficients, exchange current densities, and activation energies of the uranium reactions. Charge transfer coefficients of both salt systems were within the range of 0.17 to 0.38. Exchange current densities in the fluoride and chloride salts were estimated within the range of 0.060 to 0.12 A cm -2 . Activation energy of uranium exchange current was 16.0 kJ mol -1 in the fluoride salt and 28.9 kJ mol -1 in the chloride salt. Kinetics of charge transfer were found to be faster in FLiNaK. Analyses of the CV data suggest the electrochemical system was diffusion controlled and irreversible in the chloride and fluoride salt mixtures. Diffusion coefficients of uranium in the range of temperatures were on the order of 10-5 cm 2 s -1 in both systems. Greater values of diffusivity in ClLiK are attributed to its lower density compared with FLiNaK. Activation energy of uranium diffusion in the fluoride and chloride salt mixtures were 53.4 and 89.4 kJ mol -1 , respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Method for obtaining oxygen from lunar or similar soil

Recovery of oxygen from soil containing metal oxides such as alumina, silica, calcia, magnesia, and ilmenite wherein the material containing the oxides is placed in a vessel and reacted with fluorine to provide oxygen and metal fluorides. The oxygen produced from the reaction is recovered and stored, after further purifying processes, and the metal fluorides are further reacted with potassium vapor to provide potassium fluoride and free metals. The potassium fluoride is than subjected to electrolysis whereby the potassium and fluorine are separated and are recycled for further use in the system. Valuable free metals are recovered for other uses.

Downs, W. R.↗

Hydrothermal Synthesis and Solid-State Laser Refrigeration of Ytterbium-Doped Potassium-Lutetium-Fluoride (KLF) Microcrystals

Hydrothermal methods are used for the first time to synthesize distinct crystallographic stoichiometries within the potassium-lutetium-fluoride phase diagram for applications in solid-state laser refrigeration. Four crystalline phases were synthesized hydrothermally and doped with 10% Yb(III) ions, namely, orthorhombic K 2 LuF 5 (space group Pnma), trigonal KLuF 4 (space group P3 1 21), orthorhombic KLu 2 F 7 (space group Pna2 1 ), and cubic KLu 3 F 10 (space group $Fm\bar{3m}$), with each phase exhibiting unique microcrystalline morphologies. Among the four phases, the most significant cooling was observed for the KLuF 4 phase, which showed an overall refrigeration of 8.6 ± 2.1 K below room temperature. Laser refrigeration for KLuF4 was measured by observing both the eigenfrequencies of optomechanical cantilevers in vacuum and also the Brownian dynamics of optically trapped microcrystals in water. Cooling was also observed for the first time for the K 2 LuF 5 phase in vacuum based on measurements of the mean luminescence wavelength of Yb(III) ions. Finally, cooling was not observed with the other two phases.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Investigations into the Ternary NaF-KF-UF4 Salt System – Phase A

A knowledge gap exists in the data and understanding of fresh fuel salt and irradiated multicomponent fuel salt systems thermophysical properties. Quantifying these properties is necessary for the design and construction of test reactors, as well as the licensing of future commercial molten-salt reactors. To facilitate thermal property determination on a proposed fuel salt composition for Seaborg Technologies, several samples containing depleted uranium tetrafluoride (UF4), sodium fluoride (NaF), and potassium fluoride (KF) were blended, and a melt temperature analysis was performed. From the melting temperature analysis, it was determined that sample Seaborg-7, a ternary salt composition of 26.4UF4-24.7KF-48.9NaF (mol%), was very near a ternary eutectic point. Therefore, thermal properties such as melting temperature, salt stability, density, heat capacity, thermal diffusivity, and viscosity were experimentally determined on the Seaborg-7 salt. These measurements document the baseline properties of fresh fuel salt as a function of temperature, where future experiments on irradiated fuel salt will provide a holistic perspective on the change of thermophysical properties during reactor operations. Several precision instruments were used to collect property data, and instrument calibrations and data collection were performed and documented in a standardized and reproducible manner with meticulous detail. This process ensured that the measurement procedures and resulting data can readily be duplicated elsewhere. The Seaborg-7 salt was shown to be stable at temperatures up to 900°C, as no mass change was observed upon repeated heating and cooling. The peak melting temperature was determined to be 547°C (557°C endset). The enthalpy of fusion (??H?_fus^o) was determined to be 167.5 ± 2.7 J/g while the enthalpy of crystallization (??H?_c^o) was determined to be -147.8 ± 13.3 J/g. In addition to the eutectic melting peak, upon heating, several pre eutectic peaks were observed, occurring at 470°C (onset) and 499°C (peak). Specific heat capacity measurements showed a slightly increasing trend with respect to temperature in the solid phase, while the liquid-specific heat capacity showed a somewhat flat trend with an average value of 106.1 ± 1.24 J/mol·K between 600 to 800°C. Three independent trials using the Seaborg-7 salt determined the density to be ?(T) = 4.908 – 0.000363·T(°C), validated between 32 to 200°C, and ?(T) = 4.808 – 0.00113·T(°C), validated between ~575 to 850°C. Thermal diffusivity was determined for the liquid state and is represented by the linear equation y = 0.1581 + 0.000207·T(°C) between 550 to 850°C. The viscosity was determined from 600 to 800°C and is represented by the exponential fit equation, ? (mPa·s) = 736.58e^(-0.006·T(°C)). This report documents the conclusion of fuel salt thermophysical property measurements for the Seaborg SPP, Phase A project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Pilot Plant Makes Oxygen Difluoride

Pilot plant makes oxygen difluoride highly-energetic, space-storable oxidizer not made commercially. Designed to handle reactants, product, and byproduct, most of which highly reactive, corrosive, and toxic. Oxygen difluoride evolves continuously from reactor containing potassium hydroxide in water at 10 degree C. Collection tanks alternated; one filled while other drained to storage cylinder. Excess OF2 and F2 dissipated in combustion of charcoal in burn barrel. Toxic byproduct, potassium fluoride, reacted with calcium hydroxide to form nontoxic calcium fluoride and to regenerate potassium hydroxide. Equipment processes toxic, difficult-to-make substance efficiently and safely.

Humphrey, Marshall F.↗

Materials Refining for Solar Array Production on the Moon

Silicon, aluminum, and glass are the primary raw materials that will be required for production of solar arrays on the moon. A process sequence is proposed for producing these materials from lunar regolith, consisting of separating the required materials from lunar rock with fluorine. The fluorine is brought to the moon in the form of potassium fluoride, and is liberated from the salt by electrolysis in a eutectic salt melt. Tetrafluorosilane produced by this process is reduced to silicon by a plasma reduction stage; the fluorine salts are reduced to metals by reaction with metallic potassium. Fluorine is recovered from residual MgF and CaF2 by reaction with K2O.

Landis, Geoffrey A.↗

Fluoride Analysis by Ion Chromatography in Support of Fast Critical Assembly (FCA) Spent Nuclear Fuel Processing

INTRODUCTION The Savannah River Site (SRS) is currently processing Fast Critical Assembly (FCA) fuel received from the Japan Atomic Energy Agency (JAEA) for disposition. Stainless steel-clad plate and rods in stainless steel containers are dissolved using electrolysis with a solution mixture of nitric acid (HNO3), potassium fluoride (KF), and gadolinium (Gd). An ion chromatography (IC) was developed and vetted to monitor fluoride at various sampling points of the process. To finalize the method, FCA test solution was analyzed to qualify the analytical method followed by real FCA process solution analysis using two different analytical columns. This presentation summarizes the development and vetting of the IC method.

White, Thomas L. [Savannah River National Laborato↗

Effect of thermal annealing on the redistribution of alkali metals in Cu(In,Ga)Se 2 solar cells on glass substrate

The precise control of alkali-metal concentrations in Cu(In,Ga)Se 2 (CIGS) solar cells via post deposition treatment (PDT) has recently attracted attention. When PDT is performed at an elevated temperature, an accompanying annealing effect is expected. Here, we investigate how thermal annealing affects the redistribution of alkali metals in CIGS solar cells on glass substrates and the properties of the solar cells. In addition, we investigate the origin of non-homogeneous alkali-metal depth profiles that are typical of CIGS grown using a three-stage process. In particular, we use secondary-ion mass spectrometry measurements of the ion concentration as a function of distance from the CIGS surface to investigate the impact of thermal annealing on the distribution of alkali metals (Na, Ka, and Rb) and constituent elements (Ga and In) in the CIGS absorbers. We find that the depth profiles of the alkali metals strongly reflect the density of sites that tend to accommodate alkali metals, i.e., vacancies. Annealing at elevated temperature caused a redistribution of the alkali metals. The thermal-diffusion kinetics of alkali metals depends strongly on the species involved. We introduced low flux potassium fluoride (KF) to study a side effect of KF-PDT, i.e., Na removal from CIGS, separately from its predominant effects such as surface modification. When sufficient amounts of Na are supplied from the soda lime glass via annealing at an elevated temperature, the negative effect was not apparent. Conversely, when the Na supply was not sufficient, it caused a deterioration of the photovoltaic properties.

14 SOLAR ENERGY↗

Temperature and Excitation Dependence of Recombination in CIGS Thin Films with High Spatial Resolution

Alkali post-deposition treatments significantly improve the performance of CuInGaSe2 (CIGS) devices, but there is still room for improvement. Here, we investigate the effects of potassium fluoride alkali post-deposition treatment on the defect chemistry and recombination at grain boundaries and grain interiors using temperature- and injection-dependent cathodoluminescence (CL) spectrum imaging. We study CIGS thin films grown on alkali-free sapphire substrates to isolate the effects of alkali treatment from alkali metals that can diffuse from standard soda-lime glass substrates. We find that alkali treatment affects the energy and temperature dependence of the luminescence peaks, as well as the defect activation energies. CL spectrum images reveal that the luminescence transitions at grain boundaries have a distinct power dependence after alkali treatment and substantially different defect chemistry. This work shows that temperatureand injection-dependence CL spectrum images can provide unique insight into the defect chemistry and recombination behavior of CIGS thin films.

14 SOLAR ENERGY↗

Chemical Dissolution of Simulant FCA Cladding and Plates

The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO 3 -KF) flowsheets of H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Dissolution of a Can Carrier Pin in Concentrated Nitric Acid

H=Canyon will be dissolving Fast Critical Assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA). The FCA fuel will be dissolved in the electrolytic dissolver in a solution that is a 50 wt.% nitric acid (HNO 3 ), 0.5 g/l gadolinium (Gd), and 0.05 M potassium fluoride (KF). The nitric acid concentration is expected to decrease during the batch process from 10.3 to 7 M. The temperature of the nitric acid solution may be as low as 15 °C. The fuel can will be placed in the dissolver basket insert utilizing a reusable charging device. The charging device is a coated stainless-steel rod with a clevis design. The charging device employs a linchpin to secure the FCA fuel can as it is being charged into the H-canyon dissolver. Prior to beginning the electrolytic dissolution process, the pin will be dissolved, the fuel can will remain in the dissolver basket insert and the charging device will be removed. Currently the time necessary for complete dissolution of the pin is unknown and thus the timing of the removal of the charging device cannot be planned. This process is to be performed remotely and therefore complete dissolution of the pin will not be able to be visually determined. The pin will be made of a material that dissolves in the concentrated nitric acid solution in the dissolver. The facility desired to know the time that the pin would be dissolved so that the charging device for the FCA can could be removed from the dissolver. In particular, the pin dissolution time as a function of the nitric acid concentration and the temperature was desired. The facility would like to ensure that the solution environment was such that the charging device could be removed within an operational shift (8-12 hours).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

H-Canyon Flowsheet for the Neutralization of Dissolved Fast Critical Assembly (FCA) Fuel

The H-Canyon facility will be dissolving unirradiated stainless-steel clad Fast Critical Assembly (FCA) fuel in HNO3 and neutralizing the resulting solution with 50 wt% NaOH with no recovery operations prior to transfer to the Concentration, Storage, and Transfer Facility (CSTF). There are two types of FCA fuel, including a Pu-Al metal alloy and a mixed U and Pu oxide. H-Canyon anticipates dissolving 16 batches of the metal fuel and two batches of the oxide fuel. Potassium fluoride will be added to the dissolver solution at 0.05 M to promote dissolution. Gadolinium will be added after the dissolution as a thermal neutron poison for criticality control. H-Canyon requested that SRNL evaluate Gd:239Pu equivalent (239PuEq) ratios of 1:1 and 10:1. The neutralization process precipitates metals including actinides, Gd, and cladding components resulting in a slurry. Caustic neutralization is a routine H-Canyon operation, but the FCA dissolver solution will be unique relative to solutions that are typically processed due to the stainless-steel components, high HNO3 concentration of ~8.5 M, and an initial Pu concentration of up to 4 g/L. Previous neutralization studies have been performed for Pu containing solutions but at less than half the initial Pu concentration and much lower initial HNO3 concentrations. Experimental neutralizations targeted a final free hydroxide (OH-) concentration of 0.6 M as this was anticipated to be the final endpoint, but H-Canyon now expects the endpoint to be 1.2 M OH-. The purpose of this study was to characterize the distribution of the actinide, Gd, and cladding components between the precipitate and supernate, determine if the slurry will back up in the header during the transfer to the CSTF, and determine if solids will settle in the pipeline during the transfer.

Mills, Matthew S.↗

Plutonium Solubility and Supernate Concentration for Neutralized Fast Critical Assembly Discards to Savannah River Site Tank Waste

The Savannah River Site (SRS) plans to dissolve non-irradiated stainless steel (SS)-clad bundles of Fast Critical Assembly (FCA) materials in eighteen batches.1 FCA dissolution is currently underway in the 6.3D dissolver by simultaneous chemical and electrolytic dissolution, which is required to generate the harsh conditions necessary for dissolution of metal-oxide (MOX) and non-aluminum spent nuclear fuels (NASNFs).2 Nitric acid and potassium fluoride are used to promote chemical dissolution.2 Gadolinium will be added during processing as a thermal neutron poison for criticality control. There are no plans for recovering plutonium from this waste stream. After FCA dissolution, the acidic (HNO3/KF) “discards” containing the dissolved metals will be neutralized by addition of 50 wt% sodium hydroxide to a final free hydroxide concentration of 1.2 M.1 Neutralization will precipitate a slurry of insoluble solids, predominantly metal oxides/hydroxides of plutonium, uranium, and SS components. Small fractions of the SS components, Pu, U, and Gd will remain dissolved in the supernate. The neutralized slurry will be composited to existing radioactive waste storage tanks within the SRS Concentration, Storage, and Transfer Facilities (CSTF) containing other similar sludge batch (SB) materials.1 The fate of soluble plutonium and freshly-precipitated, colloidal plutonium from this process are of concern since the total Pu can challenge the waste acceptance criteria (WAC) at the downstream SRS Liquid Waste (LW) facility. Supernate decants including the neutralized FCA discards (nFCAd) within the CSTF will be composited with salt batch (StB) materials and transferred to the SRS Salt Waste Processing Facility (SWPF), where total plutonium is also of concern.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗