Pyrochemical processing and molten-salt reactors have recently garnered significant attention as they are promising options for future nuclear technologies, such as those for recycling spent nuclear fuels and the next generation of nuclear reactors. Both of these technologies require the use of high-temperature molten salt. To implement these technologies, one must understand the electrochemical behavior of fission products in molten salts, lanthanides, and actinides. In this study, a rotating-disk-electrode (RDE) measurement system for high-temperature molten salts is constructed and tested by investigating the electrochemical reactions of Sm3+ in LiCl–KCl melts. The results show that the reduction of Sm3+ presents the Levich behavior in LiCl–KCl melts. Using the RDE system, not only is the diffusion-layer thickness of Sm3+ measured in high-temperature molten salts but also various electrochemical parameters for Sm3+ in LiCl–KCl melts, including the diffusion coefficient, Tafel slope, and exchange current density, are determined.

The nuclide management process for reducing the environmental burden being developed by the Korea Atomic Energy Research Institute is performed in molten salts, resulting in contaminated salt wastes containing fission products such as Cs, Sr, Ba, and rare-earth nuclides. In addition, the spent fuel of a molten salt reactor (MSR) contains a variety of fission products, and a purification process may be required for the reuse of the salt and the separation and disposal of the fission products in the spent nuclear fuel. The melt-crystallization method is a technique used for the purification and separation of chemicals or metals based on the different melting points of the different substances. In a recent study, our group developed a reactive-crystallization method using Li2CO3 precipitation agent to precipitate metal corrosion from the reactor through a chlorination reaction by HCl and Cl2, which may occur in chloride molten salt, and successfully precipitated the metal precipitate and purified and recovered LiCl salt. In this study, reactive-crystallization method has been established for removing fission products and corrosive materials. Using the reactive crystallization method, white LiCl-KCl salt that was not discolored by metal corrosion was recovered through the crystallization plates, and fission products and metal elements were shown to be suppressed to several ppm in the purified salt. Consequently, high-purity salts were recovered with high nuclide and corrosive separation efficiencies. The reactive crystallization procedure can also be applied to other salt waste systems, such as MSR nuclear fuel treatment and molten salt chemistry for the elimination of corrosive substances.

It is known that ZrCl4 can be used in the chlorination process of spent nuclear fuel. However, its solubility in high temperature molten salt is very small, making it difficult to dissolve a large amount of ZrCl4. Therefore, in this study, a flange-type sealed reactor was manufactured to observe the reaction characteristics of LiCl-KCl salt and ZrCl4. LiCl-KCl salt and ZrCl4 were placed in each alumina crucible, the reactor was sealed, and heated. The temperature at the reactor surface was above 500°C and maintained at that temperature for 48 hours. After completion of the reaction, the reactor was opened and the reaction products were recovered from each alumina crucible. The crystal structure of the reaction product was identified through XRD analysis, and the concentration of Zr was analyzed using ICP. Reaction characteristics were observed according to the molar ratio of ZrCl4 added to the number of moles of KCl in LiCl-KCl salt. The molar ratios of ZrCl4 to KCl were 0.5, 1, 2, and 3, respectively. As a result of each experiment, more than 95% of the injected ZrCl4 was vaporized and there was almost no residue in the ZrCl4 crucible. In the LiCl- KCl crucible, the weight increased in proportion to the amount of ZrCl4 added. As a result of XRD analysis, K2ZrCl6 was confirmed in all LiCl-KCl salt product. When the ZrCl4/KCl molar ratio was 2 and 3, LiCl-KCl could not be confirmed. Additionally, when the ZrCl4/KCl molar ratio was 1, LiCl was identified, but KCl was not found. Almost all of the KCl appears to have reacted with ZrCl4. ICP analysis results showed that the Zr concentration was proportional to the amount of ZrCl4 added in each LiCl-KCl salt, and exceeding the number of moles of reaction with KCl in the LiCl-KCl salt was observed. Therefore, these experimental results showed that ZrCl4 can be dissolved in LiCl-KCl salt at a maximum concentration higher than its solubility.

It has been investigated on the management of Strontium-90 in KAERI. It is needed to separate the solute from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. A vacuum distillation technology was used for the separation of strontium from the molten salt in our previous study. Strontium chloride was successfully carbonated by reactive distillation of SrCl2 – K2CO3 – LiCl – KCl system. In this study, it was tried to develop another route to recover strontium from the salt solution by a solid-solid reaction for avoiding the entrainment of product and the salt-K2CO3 reaction. Reactive distillation experiments were carried out for SrCl2 - K2CO3 – LiCl – KCl system. The carbonation temperature and pressure were 520°C and 0.8 bar. After the carbonation reaction, the temperature was elevated to 820°C to remove KCl from the reaction product. SrCO3 and KCl peaks were found in the XRD analysis of the residual product. It could be concluded that SrCl2 can be successfully carbonated after salt removal by the solid-solid reaction.

The synthesis of a novel first stage GIC containing simultaneously lithium, potassium and barium through a solid–liquid reaction by molten salts method is described. Such a route has been largely developed in our laboratory for intercalation of metals into graphite. The interplanar distance of this quaternary compound reaches 950 pm and exhibits poly-layered intercalated sheets defined by X-ray measurements. The Li0.2K0.75Ba0.6C6 chemical formula of the compound is determined by ion beam analysis and this GIC is remarkably homogeneous. This GIC is the first poly-layered one containing barium.

Measuring the concentration of corrosion products or nuclear fission products (FPs) in molten salts is crucial for pyroprocessing and molten salt reactors. Electrochemical analysis methods that can be performed in situ offer significant advantages for monitoring the concentration of corrosion products or FPs in molten salts. A microelectrode is an electrode with a length of several tens of micrometers on one side. The use of a microelectrode for electrochemical analysis has several advantages due to its small size, including rapidly reaching the limiting current regardless of the scan rate, immediate attainment of the limiting current upon applying an overpotential for instant monitoring within milliseconds, accurate measurement even in low convection systems, a small iR drop resulting from low flowing current and high signal accuracy, and high current density resulting in a high signal-tonoise ratio (SNR). Among various methods for making microelectrodes, techniques involving cutting a thin wire or using capillaries (such as the dual-bore capillary and pulled glass capillary methods) require precise manual skills and experience. Therefore, the results may vary depending on the maker’s skill level, and it can be difficult to control the electrode’s area, thickness, and surface uniformly. Recent research has focused on using semiconductor processes to fabricate microelectrodes, where CVD, metal sputtering, photolithography, and etching processes work together to deposit, refine, and shape the required material on a silicon wafer to create microelectrodes. However, the durability of microelectrodes produced this way is still low (usable for about 15-30 minutes), and there is no clear research on the degradation mechanism over time. To verify the proper operation of the fabricated microelectrodes, cyclic voltammetry (CV) is performed at various scan rates (from 10 mVs-1 to 2 Vs-1), and chronoamperometry (CA) is also examined to confirm whether the electrodes rapidly reach a steady-state current. After confirming their proper operation, CV is continuously measured until the microelectrodes are destroyed in a LiCl-KCl solution containing a small amount of FPs (Sm 340 mM) at 450°C. By observing changes in the electrical signal of the microelectrodes over time, the durability is evaluated, and the mechanism of performance degradation of the electrode is discovered. The experiment is then repeated by gradually increasing the temperature by 30°C from 450°C up to 600°C to observe the changes with temperature. This study provides basic information for future microelectrode experiments, and by diagnosing the cause of destruction, a more durable microelectrode structure can be manufactured.

In molten salt reactor (MSR), liquid fuel integrated with the molten salt coolant is used to improve the safety, resulting in the prevention of the loss-of-coolant accident (LOCA) that can occur in a pressurized water reactor (PWR). Because the structural materials used in MSRs are directly contacted with liquid fuel for a long time, they must have excellent corrosion resistance to the molten salt. Therefore, to examine the corrosion rates for Ni-based alloys in the molten salt, the corrosion experiments for alloy 600, alloy 617, and Hastelloy N were performed under LiCl molten salt at 635°C for 100 h in a glove box under Ar environment. Through a weight loss method for the three Nibased coupons before and after the corrosion tests, we evaluated their corrosion rates. Based on the results of weight loss for each alloy, we confirmed that Hastelloy N has the excellent corrosion resistance compared to the other alloys. Furthermore, the changes in the crystal structure and surface morphology with elemental distribution for the three alloys by corrosion in LiCl molten salt were analyzed, showing the variation in surface topography and the decrease in Cr element after corrosion experiments for all coupons.

As a method for chlorinating spent nuclear fuel, a method of using ZrCl4 in high-temperature molten salt is known. However, ZrCl4 has a sublimation property that vaporizes at a temperature similar to the melting temperature of molten salt. Since solubility of ZrCl4 in molten salt is very low, it is difficult to dissolve a large amount of ZrCl4 in molten salt. However, once ZrCl4 can be dissolved together with the molten salt, it remains in the molten salt without vaporizing. That is, it is known that when vaporized ZrCl4 reacts with molten salt in a sealed reactor, it dissolves into the molten salt, and ZrCl4 above the solubility remains in the molten salt in the form of M2ZrCl6. Here, M represents an alkali element. Therefore, in this study, a flange-type sealed reactor was fabricated to dissolve a large amount of ZrCl4 in LiCl-KCl salt, and LiCl-KCl salt in which ZrCl4 was dissolved as K2ZrCl6 was prepared. LiCl-KCl, KCl, and ZrCl4 salts were charged into alumina crucibles and placed in a sealed reactor. The reactor was heated to 500°C and the reaction time was about 20 hours. The temperature of the reactor surface was about 480°C. After completion of the reactions, each crucible was recovered from the inside of the reactor. All of the ZrCl4 vaporized and there was no residue in the crucible. Both KCl and LiCl-KCl salts appear to have dissolved and then cooled, with respective weight gains. XRD analysis was performed to observe the structure of the recovered salts, and ICP analysis was performed to measure the Zr element content in each salt. As a result of XRD analysis, the structure of K2ZrCl6 was found in the KCl salt, but not in the LiCl-KCl salt. As a results of ICP analysis, it was found that the LiCl-KCl salt contained about 33wt% of ZrCl4, and about 25wt% was dissolved in the KCl salt. In other words, it was shown that ZrCl4 above the solubility can be dissolved in the LiCl-KCl molten salt.

During electrorefining, fission products, such as Sr and Cs, accumulate in a eutectic LiCl-KCl molten salt and degrade the efficiency of the separation process by generating high heat and decreasing uranium capture. Thus, the removal of the fission products from the molten salt bath is essential for reusing the bath, thereby reducing the additional nuclear waste. While many studies focus on techniques for selective separation of fission products, there are few studies on processing monitoring of those techniques. In-situ monitoring can be used to evaluate separation techniques and determine the integrity of the bath. In this study, laser-induced breakdown spectroscopy (LIBS) was selected as the monitoring technique to measure concentrations of Sr and Cs in 550°C LiCl-KCl molten salt. A laser spectroscopic setup for analyzing high-temperature molten salts in an inert atmosphere was established by coupling an optical path with a glove box. An air blower was installed between the sample and lenses to avoid liquid splashes on surrounding optical products caused by laser-liquid interaction. Before LIBS measurements, experimental parameters such as laser pulse energy, delay time, and gate width were optimized for each element to get the highest signal-to-noise ratio of characteristic elemental peaks. LIBS spectra were recorded with the optimized conditions from LiCl-KCl samples, including individual elements in a wide concentration range. Then, the limit of detections (LODs) for Sr and Cs were calculated using calibration curves, which have high linearity with low errors. In addition to the univariate analysis, partial least-squares regression (PLSR) was employed on the data plots to obtain calibration models for better quantitative analysis. The developed models show high performances with the regression coefficient R2 close to one and root-mean-square error close to zero. After the individual element analysis, the same process was performed on samples where Sr and Cs were dissolved in molten salt simultaneously. The results also show low-ppm LODs and an excellent fitted regression model. This study illustrates the feasibility of applying LIBS to process monitoring in pyroprocessing to minimize nuclear waste. Furthermore, this high-sensitive spectroscopic system is expected to be used for coolant monitoring in advanced reactors such as molten salt reactors.

The effect of Li2O addition on precipitation behavior of uranium in LiCl-KCl-UCl3 has been investigated in this study. 99.99% LiCl-KCl eutectic salt is mixed with 10wt% UCl3 chips at 550°C in the Pyrex tube in argon atmosphere glove box, with 10 ppm O2 and 1 ppm H2O. Then, Li2O chunks are added in mixed LiCl-KCl-UCl3 and the system has been cooled down to room temperature for 10 hours to form enough UO2 particles in the salt. The solid salt has been taken out from the glove box, and cut into three sections (top, middle and bottom) by low-speed saw for further microscopic analysis. Three pieces of solid salt are dissolved in deionized water at room temperature and the solution is filtered by a filter paper to collect non-dissolved particles. The filter paper with particles is baked in vacuum oven at 120°C for 6 hours to evaporate remaining moisture from the filter paper. Further analysis was performed for the powder remaining on the filter paper, and periphery of the powder (cake) on the filter paper. Scanning electron microscopy (SEM), electron diffraction spectroscopy (EDS), and X-ray powder diffraction (XRD) are adopted to analysis the characteristic of the particles. From SEM analysis, the powders are consisted of small particles which have 5 to 10 m diameter, and EDS analysis shows they are likely UO2 with 23 at. % of uranium and 77 at. % oxygen. Cake is also analyzed by SEM and EDS, and needle like structures are widely observed on the particle. The length of needle is distributed from 5 to 20 m, and it has 6 to 10 at. % of chlorine, which are not fully dissolved into deionized water at room temperature. From XRD analysis, the particles show the peak position of UO2, and the result is well matched with the SEM-EDS results. We are planning to add more Li2O in the system for fully reacting uranium in UCl3, and compare the results to find the effect of Li2O concentration on UO2 precipitation.

One of the promising candidates for heat transfer fluid is molten chloride salts. They have been studied in various fields such as the electrolyte of pyroprocessing, the molten salt reactor coolant, and the energy storage system media. Main considerations for utilizing molten chloride salts are the compatibility of salts with structural materials. The corrosion behavior of structural materials in molten chloride salts must be understood to identify suitable materials against the corrosive environment. In this study, the corrosion behavior of a candidate structural material, Hastelloy N, in molten LiCl- KCl salt at 500°C were investigated by the electrochemical impedance spectroscopy (EIS) method. The sheet type of Hastelloy N was utilized as the working electrode in LiCl-KCl to measure the EIS data for 100 hours with 5 hours of time intervals. The EIS data were measured in the frequency range from 104 Hz to 10-2 Hz with the AC signal (amplitude = 20 mV) at open circuit potential. The capacitance semicircle observed in Nyquist plots for all periods indicates that charge-transfer controlled reactions occur. As the immersion time progresses, the radius of the semicircle in Nyquist plots and the impedance and phase angle in Bode plots decrease. These behaviors suggest a decreasing reaction resistance and the corrosion reactions are accelerated with the immersion time. The EIS data were fitted using the equivalent circuit to achieve quantitative results. Two capacitor-resistor components were considered due to the overlapped shape of two valleys in phase angle. The depressed shape of the semicircle in Nyquist plots led to the use of the constant phase element(Q) instead of the capacitor. Therefore, R(Q(R(QR))) circuit was selected to fit the EIS data. Fitting results show that the charge transfer resistance decreases dramatically within 1 day and then converges. The film resistance shows no clear trends, but the increase of the film admittance value indicates the decreased film thickness. Consequently, the film appears to exist like the oxide layer but it does not act as a protective layer. The real-time EIS data were measured in molten salt and provides the corrosion behavior over time. The corrosion mitigation strategy should consider that the corrosion of Hastelloy N accelerates over time and its intrinsic film cannot act as the protective layer. The next steps of this study are to evaluate other candidate structural materials and to demonstrate the presence of the film.

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℃ 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℃. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320–700℃) and ramp rates (5, 10, and 20℃ min−1), HCl formation was no more than 0.6% of the Cl− in the original salt.

To estimate the removal efficiency of TRU and rare earth elements in an oxide spent fuel, basic dissolution experiments were performed for the reaction of rare earth elements from the prepared simfuel with chlorination reagents in LiCl-KCl molten salt. Based on the literature survey, NH4Cl, UCl3, and ZrCl4 were selected as chlorination reagent. CeO2 and Gd2O3 powders were mixed with uranium oxide as a representative material of rare earth elements. Simfuel pellets were prepared through molding and sintering processes, and mechanically pulverized to a powder form. The experiments for the reaction of the simfuel powder and chlorination reagents were carried out in a LiCl-KCl molten salt at 500°C. To observe the dissolution behavior of rare earth elements, molten salt samples were collected before and after the reactions, and concentration analysis was performed using ICP. After the reaction completed, the remaining oxide was washed with water and separated from the molten salt, and XRD was used for structural analysis. As a result of salt concentration analysis, the dissolution performance of rare earth elements was confirmed in the reaction experiments of all chlorination reagents. In an experiment using NH4Cl and ZrCl4, the uranium concentration in the molten salt was also measured. In other words, it seemed that not only rare elements but also uranium oxide, which is a main component of simfuel, was dissolved. Therefore, it is thought that the dissolution of rare earth elements is also possible due to the collapse of the uranium oxide structure of the solid powder and the reaction with the oxide of rare earth elements exposed to molten salt. As a result of analyzing the concentration changes of Simfuel before and after each reaction, there was little loss of uranium and rare earth elements (Ce/Gd) in the NH4Cl experiment, but a significant amount of rare earth elements were found to be reduced in the UCl3 experiment, and a large amount of rare earth elements were reduced in the ZrCl4 reaction.

Liquid Bi pool is a candidate electrode for an electrometallurgical process in the molten LiCl-KCl eutectic to treat the spent nuclear fuels from nuclear power plants. The electrochemical behavior of Bi3+ ions and the electrode reaction on liquid Bi pool were investigated with the cyclic voltammetry in an environment with or without BiCl3 in the molten LiCl-KCl eutectic. Experimental results showed that two redox reactions of Bi3+ on inert W electrode and the shift of cathodic peak potentials of Li+ and Bi3+ on liquid Bi pool electrode in molten LiCl-KCl eutectic. It is confirmed that the redox reaction of lithium with respect to the liquid Bi pool electrode would occur in a wide range of potentials in molten LiCl-KCl eutectic. The obtained data will be used to design the electrometallurgical process for treating actinide and lanthanide from the spent nuclear fuels and to understand the electrochemical reactions of actinide and lanthanide at liquid Bi pool electrode in the molten LiCl-KCl eutectic.