Search NASA⌕ Search

Engineering topics

Yoo, Tae-Sic

Publications and source records attributed to Yoo, Tae-Sic.

8 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 OSTI2024

Electrorefiner Speciation and Phase Model for Prediction of Operation Lifetime

The Mk IV electrorefiner (ER) at Idaho National Laboratory has recovered uranium metal from used nuclear fuels for 27 years. Thus, large quantities of fission product chlorides more electropositive than uranium have accumulated in the ER salt, increasing the complexity of phase equilibria and salt speciation. Importantly, the melt temperature of the salt increases with ER operation and will ultimately exceed design limits without intervention. Methods to remove fission products from molten chloride salt are known, but their implementation at scale is not yet demonstrated. In this work, historical ER composition and differential scanning calorimetry (DSC) are utilized to develop coupled speciation and chloride salt solution thermochemistry models, which enable accurate present-day analysis of ER salt as well as forecast its future disposition. The developed thermochemical models are applicable to chloride salts generally and have been made available in the Molten Salt Thermal Properties Database – Thermochemical.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Synthesis and characterization of super occluded LiCl-KCl in zeolite-4A as a chloride salt waste form intermediate

Here, this paper reports the hygroscopic properties of eutectic LiCl-KCl after absorption into zeolite-4A, up to salt loadings of 75 wt%. Samples of the salt occluded zeolite were hydrated in a humidity chamber at constant temperature and relative humidity for up to 100 h. At up to 45 wt% salt loading, the un-occluded phase of salt consisted primarily of NaCl, which forms when the Na + ions present in the zeolite framework exchange with Li + and K + ions from the eutectic LiCl-KCl. This results in minimal water absorption and corrosion of contacted stainless steel. At greater than 45 wt% salt loading, water absorption and corrosion progressively worsened. The mixture has a significant amount of excess LiCl-KCl, making it highly hygroscopic. This study reveals an option for the intermediate treatment of waste salt from spent nuclear fuel electrorefiners that could facilitate it to be stored in a non-inert atmosphere for extended periods of time before final conversion into a permanent waste form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Traveling Molten Zone Refining Process Development for Innovative Fuel Cycle Solutions

Considering the phase diagrams of metallic spent fuel constituents, the melting and solidifying of spent metallic fuel causes three immiscible layers (actinide-rich, lanthanide-rich, and Group II-rich) and the condensate phase (Group I) to form. We believe this anticipated immiscibility offers an untapped opportunity for innovative fuel cycle solutions. Through the proposed project, we anticipate confirming the expected phase behavior and develop a thermal treatment process to rapidly extract actinides from spent metallic fuels. The prime apparatus for both purposes is a traveling molten zone system with induction heating. We envision that one rapid pass of the molten zone from the bottom to the top of the metallic rod incorporating species of spent metallic fuels should produce the expected immiscible layer formation and provide species partitioning data effectively and cleanly. It will also demonstrate an actinide extraction process by concentrating the impurities at the top segment of the rod and leaving the actinide species behind as the bulk rod. The successful execution of the project will demonstrate proof of concept for a transformative process path for used metal fuels in terms of economics and safeguards.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Modelling and analysis of salt-convection effect on oxide reduction process for uranium oxides using smoothed particle hydrodynamics

Oxygen ion transport via molten salt continuum is a key kinetic feature of electrochemical reduction process for uranium oxides in molten salt. During the electrolysis, the evolution of oxygen ion concentration field in molten salt continuum appears from porous metallic layer to open bulk salt. Despite the important role affecting process throughput, its effect coupled with controlled salt movement was not adequately investigated. To capture the reactive oxygen ion transport phenomena, a multidimensional reduction model based on smoothed particle hydrodynamics (SPH) was combined with the advection-diffusion model. A pelletized oxide fuel bed capturing interconnected salt continuum is considered as a reactant material geometry, which allows forced convective delivery of molten salt through the reactant materials with a relatively low pressure head. Here, in this study, reduction time of a pelletized-fuel bed was quantified with respect to key design parameters of the reactant material configuration. The simulation results demonstrated the advantage of salt convection which effectively removes oxygen ion products from the reactant materials and consequently accelerates the reduction process. Moreover, the reduction-time correlation of a single pelletized fuel was developed as a function of relevant dimensionless parameters and was extended to the application for a pelletized-fuel bed. Finally, several insights and implications in the use of a pelletized oxide fuel were presented. The developed SPH-based framework enabled multi-physics analysis of governing dynamics in appreciating and screening candidate reactant material configurations in a cost-effective manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Water Sorption/Desorption Characteristics of Eutectic LiCl-KCl Salt-Occluded Zeolites

Molten salt consisting primarily of eutectic LiCl-KCl is currently being used in electrorefiners in the Fuel Conditioning Facility at Idaho National Laboratory. Options are currently being evaluated for storing this salt outside of the argon atmosphere hot cell. The hygroscopic nature of eutectic LiCl-KCl makes is susceptible to deliquescence in air followed by extreme corrosion of metallic cannisters. In this study, the effect of occluding the salt into a zeolite on water sorption/desorption was tested. Two zeolites were investigated: Na-Y and zeolite 4A. Na-Y was ineffective at occluding a high percentage of the salt at either 10 or 20wt% loading. Zeolite-4A was effective at occluding the salt with high efficiency at both loading levels. Weight gain in salt occluded zeolite-4A (SOZ) from water sorption at 20% relative humidity and 40°C was 17wt% for 10% SOZ and 10wt% for 20% SOZ. In both cases, neither deliquescence nor corrosion occurred over a period of 31 days. After hydration, most of the water could be driven off by heating the hydrated salt occluded zeolite to 530°C. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320–700°C) and ramp rates (5, 10, and 20°C min -1 ), HCl formation was no more than 0.6% of the Cl - in the original salt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗