Low- and intermediate level waste (LILW) repository in Gyeongju, Korea is in operation and the radioactive waste should satisfy the waste acceptance criteria (WAC) of the repository. Among the WAC of the Gyeongju LILW repository, the leachability index criterion is considered to be the criterion that is directly related to the isolation of the radionuclides from biosphere. Cesium, strontium, and cobalt should satisfy the leachability index larger than six by following the ANS 16.1 leaching test method. Several research were performed for the leachability index of Cs, Sr, and Co by following the ANS 16.1 leaching test method. However, the test condition of the previous research is expected to be different to the condition of the actual waste. Due to the radioactivity of the radionuclide such as Cs, Sr and Co, most of the research applied the surrogate of those radionuclides. The concentration of those nuclides was generally measured by the inductively coupled plasma (ICP) equipment, however, high concentration compared to the disposal limit of those nuclides due to the detection limit of the ICP was applied. From the Freundlich and Langmuir adsorption isotherms, the adsorption of the nuclides differs according to the concentration of the nuclides. As the leachability index of the nuclides is affected by the adsorption of the nuclides on the binding material, the effect of nuclide concentration is expected to be not ignorable. Therefore, the leachability index difference according to the nuclide concentration should be compared to avoid over- or underestimation of the leachability index. In this study, the difference in the leachability index according to the concentration of nuclides is aimed to be checked. Cs, Sr, and Co, which should satisfy the leachability index criterion in the WAC of the Gyeongju repository, were selected as target nuclides. Three concentrations were selected to compare the leachability index: 0.1 mol, 0.001 mol and below the regulatory exemption concentration. The concentration of non-radioactive nuclides in the leachant was measured by ICPOES and ICP-MS while the concentration of radionuclides was measured by HPGe. The result of this study can be applied as background data enhancing the WAC or disposal concentration limit of the radionuclides in Gyeongju LILW repository.

Molten salt is one of the promising medium materials for molten salt reactors and energy storage systems. Molten salt is advantageous for better physical properties such as low melting point and high boiling point, high energy capacity, high thermal conductivity, and high thermal stability than other medium materials such as water or liquid metals. However, the corrosivity of the molten salt is one of the main factors that disturbs the various applications of the molten salt. On the other hand, metallic 3-D printing technologies have developed by leaps and bounds over the past 20 years and show potential for use in cutting-edge industries such as aerospace and military purposes. However, the biggest problem of 3-D printed products is that the mechanical and physical properties are very weak along the laminated plane that was generated during the manufacturing process. In particular, other research showed that corrosion is vulnerable through the laminated surface, and corrosion along the laminated plane is not completely mitigated through a general heat treatment process although the microstructure of the surface is evaluated to be partially mitigated by the heat treatment. In this study, molten salt corrosion behaviors of simple Ni-based alloy with a composition of 80Ni- 20Cr were analyzed. Ni-based alloys were fabricated by casting and 3-D printing, and some of the 3-D printed specimens were thermally treated at 1,273 K for 1 hour to examine the effects of heat treatment on corrosion behaviors. In molten eutectic NaCl-MgCl2 melts at 973 K, Ni-based alloys were corroded for 1, 3, 7, and 28 days and their microstructural changes were analyzed by SEM-EBSD-EDS and OM. The corrosion behaviors of the alloy were also evaluated by the salt composition measured with ICPOES. 3-D printed alloy with post-treatment showed more resistivity to the molten salt corrosion than as-fabricated 3-D printed alloy. However, the corrosion rate of the 3-D printed specimen after heat treatment was still higher than that made by casting.

A molten salt reactor (MSR) is a conceptual nuclear reactor that uses molten salt with liquid fuel as its primary coolant. Based on the thermophysical and neutronic properties, MSR has advantages such as high efficiency, safety, combustion of transuranic (TRU) elements, and availability of miniaturization and on-power refueling. Various research on MSR such as system development, neutronic analysis, material development, and molten salt property analysis has been conducted, but the biggest problem is the molten salt corrosion. The molten salt corrosion on structural materials can be explained by two processes; electrochemical and chemical reactions. The reduction of oxidative ions such as fuel and TRU elements is one of the major causes of molten salt corrosion. Contamination by humidity and oxygen is also known as the accelerating factor of molten salt corrosion. Also, molten salt corrosion behaviors on structural material deteriorate when dissimilar alloys are introduced in the molten salt system. Various techniques to mitigate molten salt corrosion in fluoride system has been developed, but these are not well-verified in chloride system. In this research, various methodologies to mitigate molten salt corrosion are studied. The corrosion behaviors of 80Ni-20Cr alloy in molten eutectic NaCl-MgCl2 salt at 973 K are analyzed with various applications such as salt purification, sacrificial metal injection, and salt redox potential control. Oxygen and water impurities that can accelerate molten salt corrosion have been removed by electrochemical and chemical methods; Applying the reduction potential for H+/H2 and oxidation potential for O2-/O2, introducing HCl and CCl4 gas, and introducing the metallic Cr and recovering the ionized Cr. Corrosion acceleration/deceleration effects were analyzed when introducing the reducing reagent such as Mg and Nb or oxidizing reagent such as metallic Mo and the effect of inert metallic element (W) was also investigated. The salt potential was controlled by applying the potential to the salt and adjusting the Eu3+/Eu2+ ratio.

Several tests should be performed to estimate the structural and chemical stability of the radioactive waste. Among the tests in Gyeongju LILW repository, the leaching test which follows ANS 16.1 standard test method should be conducted for Cs, Sr, and Co radionuclides and must satisfy leachability index larger than 6 which applies deionized water as a leachant. However, the expected leachant inside the silo is groundwater that contains various ions and a high pH condition is predicted due to the concrete structures inside the silo. According to the chemical environment of the leachant, the chemical form of the radionuclides varies from precipitate to ion. Cobalt precipitates when the leachant has high pH environment which is similar condition to the cement-saturated leachant. Unlike the cobalt, cesium is preferred to exist as ion in the high pH condition. Therefore, the significant effect of the chemical environment of the leachant on the leachability of the radionuclides should be considered for the waste acceptance criteria of the radioactive waste repository. From the ‘NRC, Technical position on the waste form, rev1’, the leaching test method should follow the ANS 16.1 methods by using deionized water as leachant, however, a new leachant showing more aggressive leachability can be applied instead of deionized water. In the other hand, ASTM C1308 leaching test method recommends applying actual groundwater of the repository as a leachant. FT-04-020, the leaching test method of France, suggests the ion composition of the groundwater including the pH value. Therefore, the adequacy of using deionized water as leachant for the leaching test method of Cs, Sr, and Co should be re-examined. In this study, the leaching behavior of Cs, Sr, and Co under the several leachant types is estimated. The cement solidified specimen containing single Cs, Sr, and Co were manufactured. The leaching test following ANS 16.1 was performed by applying deionized water, simulated groundater, and cement-saturated groundwater. As a result, a leachability index difference according to the leachant type was discussed. The result of this study is expected to be a background data that helps understanding the actual leaching behavior of the Cs, Sr, and Co in the Gyeongju LILW repository.

Attempts to use the molten salt system in various aspects such as MSR or energy storage systems are increasing. However, there are limitations in the molten salt-assisted technique due to the harsh corrosiveness of the molten salt, and a more detailed study on salt-induced corrosion is needed to solve this problem. In this study, corrosion behaviors of 80Ni-20Cr alloy in various salt environments such as eutectic NaCl-MgCl2 with NiCl2, CrCl2, and EuCl3 additives were investigated. Meanwhile, the corrosion acceleration effects of 80Ni-20Cr specimens were analyzed for various ceramic materials such as SiC, Al2O3, SiO2, graphite, and BN, and metallic materials such as Ni-based alloy, Fe-based alloy, and pure metals in a molten salt environment. The experiments were conducted at 973 K for up to 28 days, and after the experiment, the microstructural change of the specimen was analyzed through SEM-EDS, and salt condition was analyzed by ICP-OES.

When disposing of spent nuclear fuel, there is a risk of exposure that could exceed the annual allowable dose due to human intrusion after the institutional control period. Therefore, it can be treated with the pyroprocess, but the decontamination factor is not sufficient, and an additional actinide recovery is required because molten waste salt-containing actinide is generated. In the case of reducing the element in the spent molten salt through an electrochemical method using a liquid Bi electrode, it is difficult to separate only the actinide element because the two-element groups are reduced together due to the large concentration difference between the actinide and the rare earth element. Therefore, a process of forming a Bi intermetallic compound using a liquid Bi electrode, which has higher element separation efficiency than a liquid Cd electrode, and physically separating the Bi intermetallic compound using the difference in density of the produced compound has been proposed. For this, it is necessary to understand the properties and density separation of the intermetallic compound to be produced, and experiments were planned and conducted for this purpose. Various metals were added to the molten Bi to form an intermetallic compound, and an analysis device such as SEM was used to determine the intermetallics distribution, composition, and internal structure. As the added metal, Ce is a representative element for lanthanide, and Hf with the most similar intermetallic density, decomposition temperature, and standard reduction potential to U, and U as a substitute element for actinide was adopted. As a result of SEM and EDS analysis, it was confirmed that the separation was made in Bi due to the density difference between the produced intermetallic compounds. A Ce-Bi intermetallic compound was observed in the upper part, Hf at a concentration smaller than the error range was detected, and a Hf-Bi intermetallic compound which containing high concentration of Ce was observed in the lower part. Separation of high-purity Ce seems to be possible in the upper part, and it seems difficult to separate high-purity Hf in the lower part. Therefore, to separate highpurity Hf, an additional process suitable for it seems to be necessary.