In this study, an aerosol process was introduced to produce CaCO3. The possibility of producing CaCO3 by the aerosol process was evaluated. The characteristics of CaCO3 prepared by the aerosol process were also evaluated. In the CaCO3 prepared in this study, as the heat treatment proceeded, the calcite phase disappeared. The portlandite phase and the lime phase were formed by the heat treatment. Even if the CO2 component is removed from the calcite phase, there is a possibility that the converted CO2 component could be adsorbed into the Ca component to form a calcite phase again. Therefore, in order to remove the calcite phase, carbon components should be removed first. The lime phase was formed when CO2 was removed from the calcite phase, while the portlandite phase was formed by the introducing of H2O to the lime phase. Therefore, the order in which each phase formed could be in the order of calcite, lime, and portlandite. The reason for the simultaneous presence of the portlandite phase and the lime phase is that the hydroxyl group (OH−) introduced by H2O was not removed completely due to low temperature and/or insufficient heating time. When the sufficient temperature (900°C) and heating time (60 min) were applied, the hydroxyl group (OH−) was removed to transform into lime phase. Since the precursor contained the hydrogen component, it could be possible that the moisture (H2O) and/or the hydroxyl group (OH−) were introduced during the heat treatment process.

Uranium-235, used for nuclear power generation, has brought radioactive waste. It could be released into the environment during reprocessing or recycling of the spent nuclear fuel. Among the radioactive waste nuclides, I-129 occurs problems due to its long half-life (1.57×107 y) with high mobility in the environment. Therefore, it should be captured and immobilized into a geological disposal system through a stable waste form. One of the methods to capture iodine in the off-gas treatment process is to use silver loaded zeolite filter. It converts radioactive iodine into AgI, one of the most stable iodine forms in the solid state. However, it is difficult to directly dispose of AgI itself in an underground repository because of its aqueous dissolution under reducing condition with Fe2+. It must be immobilized in the matrix materials to prevent release of iodine as a result of chemical reaction. Among the matrix glasses, silver tellurite glass has been proposed. In this study, additives including Al, Bi, Pb, V, Mo, and W were added into the silver tellurite glass. The thermal properties of each matrix for radioactive iodine immobilization were evaluated. The glasses were prepared by the melt-quenching method at 800°C for 1 h. Differential scanning calorimetry (DSC) was performed to evaluate the thermal properties of the glass samples. From the study, the glass transition temperature (Tg) was increased by adding additives such as V2O5, MoO3, or WO3 in the silver tellurite glass. The relative electro-static field (REF) values of V2O5, MoO3, and WO3 are about three times higher than that of the glass network former, TeO2. It could provide sufficient electro-static field (EF) to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. Therefore, the addition of V2O5, MoO3, or WO3 reinforced the glass network cohesion to increase the Tg of the glass. The addition of MoO3or WO3 in the silver tellurite glass increased Tg and crystallization temperature (Tc) with remaining the glass stability.

During the treatment of spent nuclear fuel, radioactive iodine is generated in a liquefied or gaseous form in a specific process. In the case of iodine 129, it is a long-lived nuclide with a very long halflife and has high groundwater mobility under repository conditions. Despite showing a low radioactivity value, research on the management of radioactive iodine from a long-term perspective is continuously being performed. Although research has been conducted using borosilicate glass as a medium for solidifying iodine, compatibility of I in borosilicate glass is very small and the volatility is high in the solidification process. So it is not suitable as a solidified substance of iodine. Therefore, studies on other solidification media to replace them are continuously being conducted. Our research team tried to develop a new medium that can contain iodine in a solidified body stably through a simple heat treatment process and can improve problems such as volatility and waste loading. Iodine is captured as AgI in the Ag ion-exchanged zeolite. So, TeO2, Ag2O, and Bi2O3 having a high AgI loading rate were used as main components. It was named TAB after taking the first letter of each element. In previous studies, the physical properties, structure, and chemical stability of TAB materials were confirmed. PCT (Product Consistent test) was performed to confirm chemical stability. It is mainly used to compare the chemical stability of glass materials with other glass materials, but there are limitations in evaluating the long-term chemical stability of materials. In this experiment, we tried to evaluate the long-term stability of TAB and compare it with borosilicate, which is conventionally used to treat radioactive waste. In addition, we tried to understand the leaching behavior inside the TAB medium. For this purpose, ASTM C1308 test was performed for 365 days, and distilled water and KURT groundwater were used as leachates to examine the effect of ions in the groundwater on the solidified body. To analyze the leaching behavior, ICP-MS and ICP-OES analyses were performed, and the cross-section of the sample after leaching was observed through SEM.

To reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the repository facility, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute (KEARI). The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. In addition, for efficient storage facility utilization, the short-term decay heat generated by spent nuclear fuel must be removed from the waste stream. To minimize the short-term thermal load on the repository facility, it is necessary to separate heat generating nuclides such as Cs-137 and Sr-90 from the spent fuel. In particular, Sr-90 must be separated because it generates high heat during the decay process. KAERI has developed a technology for separating Sr nuclides from Group II nuclides separated through the nuclide management process. In this study, we prepared Sr ceramic waste form, SrTiO3, by using the solid-state reaction method for long-term storage for the decay of separated Sr nuclides and evaluated the physicochemical properties of the waste form. Also, the radiological and thermal characteristics of the Sr waste form were evaluated by predicting the composition of Sr nuclides separated through the nuclide management process, and the estimation of centerline temperature was carried out using the experimental thermal data and steady state conduction equation in a long and solid cylinder type waste form. These results provided fundamental data for long-term storage and management of Sr waste.

This study deals with replacement analysis of deteriorated equipment for improving productivity of production system. Frequent breakdown of the deteriorated equipment causes a situation that reduces productivity such as low product quality, process delay, and repair cost. However, the replacement of new equipment will be required a high initial investment cost, so it is important to analysis the economic feasibility. Therefore, we analyze the effect of the production system due to the aging effect of the equipment and the feasibility of equipment replacement based on the economic analysis. The process flow, working time, logistics movement, etc. are analyzed in order to build the simulation modeling for a ship and land switchboard production system. Using numerical examples, the economic feasibility analysis of equipment replacement through replacement of existing deteriorated equipment and additional arrangement of new facilities is performed.

위치 기반 AR 게임에서 모바일 기기의 위치를 정확하게 추정하는 것은 중요한 요소 중 하나이다. 모바일 기 기에 내장되어 있는 위치 추정 시스템이 얼마나 정확하게 위치정보를 추정하는 지가 증강되는 AR 콘텐츠의 정확도를 결정한다. 하지만 도시 환경에서는 건물, 건축물, 광고판, 표지판 등 지형지물 및 장애물에 의해서 위치 추정을 위해 필요한 신호가 반사, 굴절, 회절, 차단 등이 발생하게 되고 그로 인하여 모바일 기기의 위 치 추정 시스템으로부터 추정되는 위치에 오차가 생기게 된다. 본 논문에서는 도시 환경에서 모바일 기기의 위치 오차가 증가하는 현상에 대해서 상용 스트리트 뷰와 문자 태그를 활용하여 심플하면서도 신뢰성 있는 위치 보정 방안을 제안한다. 제안하는 위치 보정 방안은 쿼리 이미지와 스트리트 뷰 파노라마 이미지로부터 생성된 문자 태그를 대조하는 것을 통해서 매칭 스코어를 계산한고 매칭 스코어에 따라 쿼리 이미지를 촬영 한 실제 위치와 가까운 스트리트 뷰를 검색하는 것을 통해서 모바일 기기의 위치를 보정한다. 제안하는 위치 보정 방안은 위치 오차가 43.71m인 위치를 위치 오차가 4.09m인 위치로 보정하였으며 낮과 밤에 관계없이 위치를 보정할 수 있다는 장점이 있다.

Graphene fiber is considered as a potential material for wearable applications owing to its lightness, flexibility, and high electrical conductivity. After the graphene oxide (GO) solution in the liquid crystal state is assembled into GO fiber through wet spinning, the reduced graphene oxide (rGO) fiber is obtained through a reduction process. In order to further improve the electrical conductivity, herein, we report N, P, and S doped rGO fibers through a facile vacuum diffusion process. The precursors of heteroatoms such as melamine, red phosphorus, and sulfur powders were used through a vacuum diffusion process. The resulting N, P, and S doped rGO fibers with atomic% of 6.52, 4.43 and 2.06% achieved the higher electrical conductivities compared to that of rGO fiber while preserving the fibrious morphology. In particular, N doped rGO fiber achieved the highest conductivity of 1.11 × 104 S m−1, which is 2.44 times greater than that of pristine rGO fiber. The heteroatom doping of rGO fiber through a vacuum diffusion process is facile to improve the electrical conductivity while maintaining the original structure.