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Tolman, Kevin R.

Publications and source records attributed to Tolman, Kevin R..

7 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Melt Crystallization of CsF from Alkali Fluorides

Fractional melt-crystallization is a technique used to separate components in a multicomponent liquid mixture through controlled cooling. In fiscal year (FY) 2023, this technique was successfully used to separate CsCl from LiCl-KCl. This demonstrated a potential route for concentrating electrorefiner fission product waste streams in pyrochemical fuel cycles, building on previous work that developed the melt-crystallization system for fission product removal from LiCl-based electrolytes used for oxide reduction. This work investigated whether a thermally controlled process of a solid-liquid separation process could effectively remove CsF from LiF-NaF-KF (FLiNaK)-CsF salt for MSR fuel cycle applications. The designed process aimed to recover purified LiF-NaF-KF salt as solid precipitates while concentrating CsF to a remaining salt heel. This concentrated CsF can then be immobilized during a salt waste stream treatment operation, minimizing waste volume.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Synthesis and thermophysical property determination of NaCl-PuCl3 salts

Currently, a knowledge gap exists in the available data and understanding of thermophysical properties relating to fresh fuel salts, especially those containing plutonium. These data are necessary for designing, constructing, and licensing future commercial molten-salt reactors. Thermophysical properties, such as melting temperature, salt stability, density, and heat capacity were ascertained using NaCl-PuCl3 (36 mol% PuCl3) and a more sodium rich composition containing 25 mol% PuCl3. The NaCl-PuCl3 salt mixture was synthesized for this study and contained 63.4 mol% NaCl, 36.3 mol% PuCl3 and was 99.7% pure. Upon heating, the NaCl–PuCl3 eutectic was stable at temperatures up to 800 °C. The onset of melting occurred at 451 ± 3 °C, and the enthalpy of fusion was determined to be 23.0 ± 1.4 kJ/mol. Heat capacity measurements in the liquid phase ranged from 107.7 to 91.3 J/mol.K, with an average value of 104.6 ± 11.4 J/mol.K between 500 and 650 °C. Three independent trials of the molten NaCl-PuCl3 salt found the density to be ?(T) = 3.8589 – 9.5342·10-4 T(°C). In addition, ab initio molecular dynamic simulations to calculate density and heat capacity values are included.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Synthesis and characterization of uranium trichloride in alkali-metal chloride media

Given a growing interest in uranium salts for pyrochemical processing of used fuel and uranium-fueled molten salt reactors, the synthesis of uranium trichloride in alkali-metal chloride media was investigated in a series of four experiments. Specifically, uranium metal powder and uranium hydride powder were prepared and separately blended with ammonium chloride and lithium chloride – potassium chloride eutectic in two runs, while the same powders were separately blended with ammonium chloride and sodium chloride in two additional runs. Each of the lithium chloride – potassium chloride containing blends was slowly heated to 923 K, while those containing sodium chloride were heated to 1123 K. During each heat up, the ammonium chloride sublimed into gaseous ammonia and hydrogen chloride, leading to the chlorination of uranium metal or uranium hydride and the formation of molten salt solutions of the respective chlorides. Experimental conditions were incorporated in the runs to promote formation of uranium trichloride over uranium tetrachloride in the respective media. Molten samples of each run product were taken and characterized via chemical analyses, diffractometry, and microscopy. The final products from each run were dark dense ingots of the respective salt systems with uranium concentrations ranging from 44 to 51 wt%. Finally, chemical analyses and diffractometry identified the predominant presence of uranium trichloride in these systems; however, a possible minor presence of uranium tetrachloride could not be conclusively dismissed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Out-of-pile and postirradiated examination of lanthanide and lanthanide-palladium interactions for metallic fuel

Palladium is being investigated as a fuel additive to bind with and potentially immobilize lanthanide fission products. A primary cause of fuel-cladding chemical interaction (FCCI) is the lanthanide fission products migrating to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. Palladium has previously been identified as a promising additive used to prevent or decrease FCCI by reacting with the lanthanide fission products. In the current study, an alloy cast from the four highest abundant lanthanides found in irradiated metallic fuel, Nd, Ce, Pr, and La, with and without Pd, has been characterized using neutron diffraction, scanning electron microscopy, and electron probe microanalysis. In the lanthanide-Pd intermetallic compounds, all of the constituent compounds, i.e. Nd-Pd, Ce-Pd, La-Pd and Pr-Pd are known. There is very good agreement, both structurally and compositionally, between the out-of-pile lanthanide alloy and lanthanide fission products characterized in irradiated fuels. In both cases, the lanthanide elements form a solid solution in a hexagonal crystal structure. The out-of-pile lanthanide alloy follows Vegard's Law, with the measured and calculated (weighted average of constituents) lattice parameters being within 1% for both the a and c parameters. Pd bonds with the lanthanides (Ln) forming the phases LnPd and Ln7Pd3. The results indicate the properties of lanthanide compounds in irradiated metallic fuel can be reliably simulated in out-of-pile experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗