Wasteform is the first barrier to prevent radionuclide release from repositories into the biosphere. Since leaching rates of nuclides in wasteform significantly impact on safety assessment of the repository, clarifying the leaching behavior is critical for accurate safety assessment. However, the current waste acceptance criteria (WAC) of the Gyeongju repository only evaluates leachability indexes for Cs, Sr, and Co, which are tracers for nuclear power plant waste streams. Furthermore, ANS 16.1, the current leaching test method used in WAC, applies deionized water (DI) as leachant. However, the interactions between wasteform and groundwater environment in the repository may not be reflected. Therefore, it is necessary to review the current leaching test method and nuclides that may require the extra evaluation of leachability beyond the Cs, Sr, and Co. Tc and I are key nuclides contributing to high radioactive dose in safety assessment due to their high mobility and low retardation factor. The groundwater conditions within the repository, such as pH and Eh significantly affect the chemical form of Tc and I. For example, Tc in H2O system tends to form hydroxide precipitates in neutral pH condition and TcO4 - in strong alkaline environments according to the Pourbaix diagram. In case of I, it generally exists in the form of I-, while it exists as IO3 - as Eh increases. Although the current leaching test at the Gyeongju repository applies DI as a leachant, the actual repository is expected to have a highly alkaline environment with a substantial amount of various ions in the groundwater. Consequently, the leaching behavior in the ANS 16.1 test and the actual disposal condition is different. Thus, it is necessary to analyze the leaching behavior of Tc and I with reflecting the actual disposal environment. In this study, the leaching behavior of Tc and I is investigated by following ANS 16.1 leaching test method. The solidified waste specimens containing 10 mmol of Re and I were manufactured with cement, which is widely used as a solidification material. Re was applied instead of Tc, which has similar chemical behavior to Tc, and NH4ReO4 and NaI were used as surrogates for Re and I. As a leachant, deionized water and cement-saturated groundwater were prepared and the concentration of nuclides in the leachant is analyzed by ICP-OES. As the result of this study, experimental data can be applied to improve the WAC and disposal concentration standards in the future.

The solid-state chemistry of uranium is essential to the nuclear fuel cycle. Uranyl nitrate is a key compound that is produced at various stages of the nuclear fuel cycle, both in front-end and backend cycles. It is typically formed by dissolving spent nuclear fuel in nitric acid or through a wet conversion process for the preparation of UF6. Additionally, uranium oxides are a primary consideration in the nuclear fuel cycle because they are the most commonly used nuclear fuel in commercial nuclear reactors. Therefore, it is crucial to understand the oxidation and thermal behavior of uranium oxides and uranyl nitrates. Under the ‘2023 Nuclear Global Researcher Training Program for the Back-end Nuclear Fuel Cycle,’ supported by KONICOF, several experiments were conducted at IMRAM (Institute of Multidisciplinary Research for Advanced Materials) at Tohoku University. First, the recovery ratio of uranium was analyzed during the synthesis of uranyl nitrate by dissolving the actual radioisotope, U3O8, in a nitric acid solution. Second, thermogravimetric-differential thermal analysis (TG-DTA) of uranyl nitrate (UO2(NO3)2) and hyper-stoichiometric uranium dioxide (UO2+X) was performed. The enthalpy change was discussed to confirm the mechanism of thermal decomposition of uranyl nitrate under heating conditions and to determine the chemical hydrate form of uranyl nitrate. In the case of UO2+X, the value of ‘x’ was determined through the calculation of weight change data, and the initial form was verified using the phase diagram for the U-O system. Finally, the formation of a few UO2+X compounds was observed with heat treatment of uranyl nitrate and uranium dioxide at different temperature intervals (450°C-600°C). As a result of these studies, a deeper understanding of the thermal and chemical behavior of uranium compounds was achieved. This knowledge is vital for improving the efficiency and safety of nuclear fuel cycle processes and contributes to advancements in nuclear science and technology.

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.

본 논문은 패닝을 활용한 사운드 연출의 새로운 접근법에 대해 연구하기 위해 패닝의 다양한 활용을 통해 개성 있는 사운드와 폭넓은 음악적 파노라마를 연출하는 미국 일렉트로닉 듀오 오데자(ODESZA)의‘A Moment Apart’를 분석하였다. 분석을 통해 일반적인 팝 음악이 중앙 중심의 대칭으로 진행되는 반면 연주곡의 경우 곡 전반에 걸쳐 불규칙하고 비대칭적인 요소들을 포함하는 양상을 보이고 있으며 스테레오 채널의 좌, 우를 분리·교차·이동 하는 등의 비대칭적인 요소를 사용해 스테레오 이미지를 형성하고, 나아가 곡 전반에 걸쳐 역동적인 스테레오 이미지 변화를 연출하는 것을 확인하였다. 이처럼 패닝은 악기의 음상 배치를 위한 용도 외에도 사운드 연출에 있어 폭넓게 활용될 수 있으며, 나아가 독특한 음색을 연출하고 공간을 재창조하는 도구로 응용될 수 있다는 것을 제시한다.

Membrane separation processes have been widely used in water purification or gas separation applications because of their process simplicity, low cost operation and small footprint. Nowadays, applications of ceramic membranes with high thermal, chemical and mechanical resistance has rapidly grown in new separation applications where the polymeric membranes cannot be used (e.g., high temperature, strong acidic/basic or solvent-contained corrosive feed solution). In this study, robust ceramic hollow fiber membranes were prepared by extrusion-phase inversion followed by sintering. The effects of preparation conditions on membrane characteristics were studied to improve the separation performance of ceramic membranes. In addition, a variety of modification methods and applications based on ceramic hollow fiber membranes will be discussed.