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.

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.

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.

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.

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.