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.

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.

The stabilization techniques are highly required for damaged nuclear fuel to strengthen safety in terms of transportation, storage, and disposal. This technique includes recovering fuel materials from spent fuel, fabrication of stabilized pellets, and fabrication of fuel rods. Thus, it is important to identify the leaching behavior of the stabilized pellets to verify their stability in humid environments which are similar to storage conditions. In this study, we introduce various leaching experiment methods to evaluate the leaching behavior of the stabilized pellets, and determine the most suitable leaching test methods for the pellets. Also, we establish the leaching test conditions with various factors that can affect the dissolution and leaching behavior of the stabilized pellets. Accordingly, we prepare the simulated high- (55 GWd/tU) and low- (35 GWd/tU) burnup nuclear fuel (SIMFUEL) and pure UO2 pellets sintered at 1,550°C and 1,700°C, respectively. Each pellet is placed in a vessel and filled with DI water and perform the leaching test at three different temperature to verify the leaching mechanism at different temperature range. Based on the standard leaching test method (ASTM C1308-21), the test solution is removed from the pellet after specific time intervals and replaced in the fresh water, and the vessel is placed back into the controlled-temperature ovens. The test solutions are analyzed by using ICP-MS.

Uranium-235, used in nuclear power generation, produces a lot of radioactive waste. Among radioactive waste nuclides, I-129 is problematic due to its long half-life (1.57×107 y) with high mobility in the environment. It should be captured and immobilized into a geological disposal environment through a stable waste form. In this study, various additives including Al, Bi, Pb, V, Mo and W were added to silver tellurite glass to prepare a matrix for immobilizing iodine, and its thermal and leaching properties were evaluated. To prepare glass, the glass precursor mixture was placed in alumina crucibles and heated at 800°C for 1 h. Except for aluminum, there was no significant loss of constituent elements. The loading of iodine in the matrix was approximately 11-15% by weigh, excluding oxygen. The normalized releases of all the elements obtained by PCT-A were below the order of 10-1 g/m2, which satisfies US regulation (2 g/m2). Differential scanning calorimetry was performed to evaluate the thermal properties of the glass samples. The glass transition temperature (Tg) increased by adding such as V2O5, MoO3, or WO3. The similar relative electrostatic field values of V2O5, MoO3, and WO3 could provide sufficient electro static field to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. The addition of MoO3 or WO3 in the silver tellurite glass system increased glass transition temperature (Tg) and crystallization temperature (Tc) while maintaining the glass stability.

Irradiated uranium dioxide in damaged used fuel could oxidize during transportation, interim storage or disposal, resulting that the fuel pellet fragments are reduced to a grain-sized powder that can easily escaped from the damaged rod. It has been reported that oxidized spent fuel (i.e. U4O9+x) that was in contact with water could increase the dissolution rate by making the grain boundaries more accessible to the water. Therefore, the damaged used fuel requires stabilization technology including nuclear material recovery, pellet manufacturing process, and stabilization fuel rod manufacturing that can secure safety in terms of permanent disposal. In this study, we prepared pure UO2 and SIMFUEL pellets that are a mixture of UO2 and surrogated metallic oxides for fission products equivalent to a burn-up of 35 GWd/tU and 55 GWd/tU as the stabilized spent fuel. The UO2 and fission products powders were milled and pressed into pellets at 250 MPa and sintered at 1,550°C and 1,700°C for 6 hours in an atmosphere of 4%H2-Ar. The prepared UO2 and SIMFUEL pellets were placed in PTFE Teflon vessels and filled with deionized water to identify the leaching behavior by a long-term leaching experiment under the similar condition to a repository for the safe disposal.

In nuclear power plants and nuclear facilities, radioactive waste containing hazardous substances (Mixed waste) is continuously generated due to research such as radiochemical study and nuclide analysis. In addition, radioactive waste including heavy metals and asbestos is generated during the dismantling process of nuclear power plants. Mixed wastes have both radiation hazards and chemical hazards, and there’s a possibility of synergistic effects generation. However, in most countries except the United States, there are no regulatory standards for the chemical hazards of mixed waste. The regulations applicable to mixed waste in Korea include the Nuclear Safety Act and the Waste Management Act. The Nuclear Safety Act prohibits the acceptance of hazardous radioactive waste in disposal facilities, but there is no definition or characteristic identification procedure for “hazardous.” The Waste Management Act also does not state the regulation for radioactive waste. In the Gyeongju disposal facility in Korea, the leachate in the disposal facility is expected to be a groundwater saturated with concrete and is expected to irradiated by radioactive waste. On the other hands, most of the non-radioactive waste landfill facilities are built on the surface, and the leachate is expected to be rainwater that reacts with the soil. Due to the differences in leaching environments, there’s a potential to overestimate or underestimate the leaching properties of hazardous substances if the standard leaching test is applied. To show for this, a leaching test simulating disposal facility’s environment were applied to sample waste containing heavy metals. The leaching solution was groundwater collected from the area near the Gyeongju disposal facility, which is then saturated with concrete and adjusted to pH 12.5. In addition, gamma-ray irradiation was conducted during the leaching test to observe changes in the leaching behavior of heavy metals in the actual radioactive waste disposal environment. As a result, lead showed significantly increased leaching compared to the standard test method, and cadmium was not detected in all experimental conditions except heavy irradiation. This study suggested that regulations on the hazardous of mixed waste should be settled, which should be established in sufficient consideration of the types and characteristics of substances contained in the waste.

Leaching of MOCVD dust in the LED industry is an essential stage for hydro-metallurgical recovery of pure Ga and In. To recover Ga and In, the leaching behavior of MOCVD scrap of an LED, which contains significant amounts of Ga, In, Al and Fe in various phases, has been investigated. The leaching process must be performed effectively to maximize recovery of Ga and In metals using the most efficient lixiviant. Crystalline structure and metallic composition of the raw MOCVD dust were analyzed prior to digestion. Subsequently, various mineral acids were tested to comprehensively study and optimize the leaching parameters such as acidity, pulp density, temperature and time. The most effective leaching of Ga and In was observed for a boiling 4 M HCl solution vigorously stirred at 400 rpm. Phase transformation of GaN into gallium oxide by heat treatment also improved the leaching efficiency of Ga. Subsequently high purity Ga and In can be recovered by series of hydro processes.

콘크리트는 장기수명이 요구되는 구조물에 적합한 건설재료로 내구성이 우수하지만 장기간 지하수에 노출되어 발생하는 칼슘용출 현상에 대한 이해 및 이에 따른 RC 부재의 휨 거동 특성을 평가할 필요가 있다. 실험결과에 따르면, 광물질 혼화재는 RC 부재의 장기강도 개선에 효과적이지만, 칼슘용출이 발생하면 RC 부재의 항복하중 및 휨 강성이 감소하고 중립축깊이와 처짐량이 증가하는 것으로 나타났다. 따라서 칼슘용출에 의한 열화는 RC 부재의 성능 저하를 유발하므로 칼슘용출 환경에 노출되는 지하구조물에 광물질 혼화재가 적용될 경우에는, 광물질 혼화재 종류에 따른 최적 혼입 비율이 마련되어야할 것으로 사료된다.

In this paper, the flexural test of concrete member damaged due to Calcium leaching were performed to evaluate the characteristics of structural behavior. From the results, yield load, maximum load and flexural rigidity of RC member damaged by calcium leaching were decreased and the depth of the neutral axis was increased.

다양한 열화 인자에 의해 저감된 콘크리트의 내구성은 구조물의 구조적 성능과 사용 수명에 부정적인 영향을 미치게 되며 콘크리트 연구 분야 중에서도 매우 중요하고 매력적인 주제라고 할 수 있다. 이러한 이유로 콘크리트의 내구성과 관련된 많은 연구들이 발표되었으나 물리-화학적 열화에 기인하는 콘크리트의 본질적인 물성 변화에 주된 초점이 맞춰져 왔으며, 콘크리트 내구성과 구조물의 구조적 성능 사이의 관계 정립은 아직 미흡한 실정이다. 본 연구에서는 콘크리트의 강도 감소에 원인이 되는 칼슘 용출 열화를 적용하였으며, 열화 정도에 따른 구조적 거동을 평가하기 위하여 열화 손상을 입은 콘크리트 부재의 압축 및 휨 거동 실험을 수행하고 그 결과를 비선형 유한요소해석 결과와 비교 분석하였다. 연구 결과에 따르면 칼슘 용출 열화는 콘크리트의 압축 강도를 저하시키며, 열화가 진행됨에 따라 취성 거동에서 연성 거동으로 변화되는 경향을 나타냈다. 또한 열화에 의한 압축 영역의 손상 정도가 심화될수록 RC 부재의 내하력과 강성은 저하되었으며, 이러한 구조적 거동은 ABAQUS의 CDP 모델을 사용한 비선형 유한요소해석의 결과와도 비교적 잘 일치하였다.

Waste dusts generated from various metallurgical processes were investigated for effective management of the designated waste in South Korea. Source of the study were selected based on the industries registered on the Allbarosystem. Raw material, products and manufacturing process were investigated on-site and leaching behavior of six heavy metals (Cr6+, Cd, Cu, Pb, Hg, As) along with CN was studied in 41 waste dust samples. We identified 14 samples exceeded the Korean regulation standard out of which 5 samples were designated waste. We suggest that regular monitoring is essential for the proper management.