In an automotive plant, an automated storage and retrieval system (ASRS) synchronizes material handling flows from a part production line to an auto-assembly line. The part production line transfers parts on small-/large-sized pallets. The products on pallets are temporarily stored on the ASRS, and the ASRS retrieves the products upon request from the auto-assembly line. Each ASRS aisle is equipped with narrow-/wide-width racks for two pallet sizes. An ASRS aisle with narrow-/wide-width racks improves both storage space utilization and crane utilization while requiring delicate ASRS aisle design, i.e., the locations of the narrow-/wide-width racks in an ASRS aisle, and proper operation policies affect the ASRS performance over demand fluctuations. We focus on operation policies involving a common storage zone using wide-width racks for two pallet sizes and a storage-retrieval job-change for a crane based on assembly-line batch size. We model a discrete-event simulation model and conduct extensive experiments to evaluate operation policies. The simulation results address the best ASRS aisle design and suggest the most effective operation policies for the aisle design.

본 연구는 코팅 방법을 활용한 단결정 양극 소재 연구로서 Ni-rich계 다결정 양극 소재로 부터 단결정 양극 소재를 합 성하여 사이클 구동 시 양극 소재의 안정성을 향상시키고자 한다. 양극 소재에 LixCoO2와 LixSnO3 를 각각 코팅하여 이차입자 내부 혹은 외부에 코팅층이 형성된 양극 소재를 합성한 후 이를 소결하여 단결정 형성에 대한 영향을 비교 하였다. 입자 외부에 LixSnO3가 코팅되어 열처리 된 Ni0.8Co0.1Mn0.1O2(NCM811)의 경우 코팅 처리 없이 열처리된 양극 소재 보다 개선된 수명특성을 보였으나, 단결정화가 이뤄지지 않았다. 입자 내부에 LixCoO2 코팅층이 형성된 NCM811 을 열처리 한 결과 이차입자 내부에 형성된 Co 코팅층이 결정화되어 50회 사이클 후 기준 단결정 양극 소재의 방전용 량인 117.34 mAh·g-1 대비 129.11 mAh·g-1의 높은 방전용량을 나타내었고, 형상제어를 통해 이성적인 단결정화가 이뤄 졌다. 본 연구는 다결정체인 Ni-rich 양극소재의 단결정화에 대한 유요한 통찰력을 제공할 것으로 예상한다.

In additive manufacturing, the flowability of feedstock particles determines the quality of the parts that are affected by different parameters, including the chemistry and morphology of the powders and particle size distribution. In this study, the microstructures and flowabilities of gas-atomized heat-resistant alloys for additive manufacturing applications are investigated. A KHR45A alloy powder with a composition of Fe-30Cr-40Mn-1.8Nb (wt.%) is fabricated using gas atomization process. The microstructure and effect of powder chemistry and morphology on the flow behavior are investigated by scanning electron microscopy (SEM), X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), and revolution powder analysis. The results reveal the formation of spherical particles composed of single-phase FCC dendritic structures after gas atomization. SEM observations show variations in the microstructures of the powder particles with different size distributions. Elemental distribution maps, line scans, and high-resolution XPS results indicate the presence of a Si-rich oxide accompanied by Fe, Cr, and Nb metal oxides in the outer layer of the powders. The flowability behavior is found to be induced by the particle size distribution, which can be attributed to the interparticle interactions and friction of particles with different sizes.

본 연구의 목적은 어머니의 주중 미디어 놀이 횟수가 유아의 놀이상호작용에 미치는 영향에서 자기통제와 주의집중문제의 매개역할을 살펴보는 것이다. 이를 위해 육아정책연구소의 한국아동패널 7차년도 데이터를 사용했고, 자료는 R 4.1.3과 PROCESS macro v4.1 for R로 분석했다. 본 연구의 결과는 다음과 같다. 첫째, 어 머니의 주중 미디어 놀이 횟수는 유아의 자기통제에 부적 영향, 주의집중문제에 정적 영향을 미쳤다. 또래 놀이상호작용에 대해 자기통제는 정적 영향, 주의집중문제는 부적 영향을 미쳤다. 자기통제는 주의집중문제 에 부적 영향을 미쳤다. 둘째, 어머니의 주중 미디어 놀이 횟수가 유아의 또래놀이상호작용에 미치는 영향 에서 자기통제와 주의집중문제는 각각 부적 매개효과가 나타났다. 마지막으로 어머니의 주중 미디어 놀이 횟수가 유아의 또래놀이상호작용에 미치는 영향에서 자기통제와 주의집중문제는 부적 직렬다중매개효과가 있었다. 본 연구의 결과는 유아의 자기통제와 주의집중문제가 어머니의 주중 미디어 놀이 횟수와 또래놀이 상호작용을 매개하는 주요한 요인임을 시사한다.

Untreated waste is temporarily stored on the site of the nuclear power plant. In some nuclear power plants, saturation period of temporary storage waste is less than 10 years away. As untreated waste continues to be generated in nuclear power plants, it could also affect management of operations. Accordingly, CRI is developing the 3.5 generation plasma torch melting facility for waste treatment. The 3.5th generation plasma torch melting facility consists of melter, plasma torch, waste supply device, exhaust gas treatment facility, power supply, etc. Melter is composed of melting chamber for melting control and pyrolysis chamber for waste pretreatment, and dam-type discharge device is adopted to overflow the melt. Plasma torch is hollow type with reversed discharge, has a rating of megawatt class, and has two gas supply lines. It can be used in transfer mode, non-transfer mode and mixed mode. There are three types of device for waste supply. The first is a drum pusher for injecting 200 L drums, the second is a screw-type waste supply and hopper for injecting solid waste, and the third is a nozzle-type waste supply device for injecting liquid waste. Exhaust gas treatment facility was equipped with post combustion chamber, off-gas cooler, high-temperature filter, HEPA filter, reheater, scrubber, ID fan and etc. Power supply of plasma torch operation is designed with a capacity of 1.5 megawatt (Maximum) and consists of channels A and B. Transfer mode, non-transfer mode and mixing mode of plasma torch may be selected through the control of PLC. This paper introduces the composition and function of the 3.5th generation plasma torch melting facility of CRI. In order to solve the problems arising through the operation of the 3rd generation plasma torch melting facility, an optimization plan is applied.

The depth of geological disposal of high-level radioactive waste (HLW) varies from country to country, but it is generally considered below 300 m underground. As one of the reliable methods to understand the geological characteristics of these deep areas, the site investigation through drilling is recommended. This paper deals with multidisciplinary research that evaluates the geological characteristics of the site using deep drilling. The deep drilling is 750 m, which is higher than the planned disposal depth. Prior to drilling, literature and surface geological surveys of the target area were conducted, and during drilling, real-time measurement of excavated information for obtaining drilling information, circulating water management and chemical composition through a closed system were monitored. After drilling, field tests such as geophysical borehole logging, deep groundwater sampling, constant pressure injection test, and hydraulic fracturing test were performed. Analysis of the recovered drilling core from a geological point of view such as age dating, rock formation and structural geological analysis, and from geochemical perspectives such as concentration of major/ minor cationic elements, major anions, and trace elements along with the water quality parameters pH, DO, Ec, Eh, etc., from geothermal perspective such as thermal conductivity and coefficient of thermal expansion, from rock mechanical aspects such as physical and mechanical properties of intact rocks and joints, joint distribution, etc. Deep drilling has been completed with 2 holes for granite and 2 holes for sedimentary rocks, and further drilling for gneiss and sedimentary rocks is in progress.

Considering the domestic condition with small land area and high population density, it is necessary to develop technology that can reduce the disposal area than the deep geological disposal method. For this, KAERI is developing a nuclide management process that can reduce the environmental burden of spent fuel, and establishing an evaluation model that can evaluate the performance of various process options. It is expected that an optimal option of the nuclide management process can be derived from disposal perspective by applying the evaluation model. The mass flow between processing steps of the radionuclide management process is the basic quantity required to quantify the evaluation criteria. Therefore, we built a generalized block model on GoldSim, which can simulate mass flow of various radionuclide management process options. In addition to the mass flow, this model was established to derive the amount of wastes generated by each processing step, the composition of nuclides, and radiological properties (decay heat, radioactivity, etc.). The mass flow and waste property derived from the models are closely related to the factors that determine the area of disposal concepts. Based on this, a disposal area calculation model was established as a model to evaluate the effectiveness of the radionuclide management process on environmental burden reduction. For verification, three process options, which can manage radionuclides having high decay heat (Cs, Sr) or large volume (U), were selected and evaluated as reference processes. And two disposal options, deep geological disposal and deep borehole disposal concepts were considered to be linked with the processes. As a result, it was confirmed that the disposal area could be reduced in the process separating radionuclides having high decay heat. In the future, other evaluation models for economic viability and safety will be added in the GoldSim model.

An automated material handling system (AMHS) has been emerging as an important factor in the semiconductor wafer manufacturing industry. In general, an automated guided vehicle (AGV) in the Fab’s AMHS travels hundreds of miles on guided paths to transport a lot through hundreds of operations. The AMHS aims to transfer wafers while ensuring a short delivery time and high operational reliability. Many linear and analytic approaches have evaluated and improved the performance of the AMHS under a deterministic environment. However, the analytic approaches cannot consider a non-linear, non-convex, and black-box performance measurement of the AMHS owing to the AMHS’s complexity and uncertainty. Unexpected vehicle congestion increases the delivery time and deteriorates the Fab’s production efficiency. In this study, we propose a Q-Learning based dynamic routing algorithm considering vehicle congestion to reduce the delivery time. The proposed algorithm captures time-variant vehicle traffic and decreases vehicle congestion. Through simulation experiments, we confirm that the proposed algorithm finds an efficient path for the vehicles compared to benchmark algorithms with a reduced mean and decreased standard deviation of the delivery time in the Fab’s AMHS.

The amount of radioactive waste generated during decommissioning directly affects the disposal cost of waste. Most of the radioactive waste generated is a concrete waste. Therefore reducing the amount of concrete waste can ensure the economic feasibility of the decommissioning project. The activated concrete in a concrete waste can reduce waste only by physical cutting. Therefore it is most important to accurately identify and categorize radionuclides, radioactivity levels, and radioactivity distribution. In the case of radioactive concrete, radiological characteristics are generally evaluated by laboratory analysis after sampling. However it is difficult to apply to all facilities (accelerator & NPP, etc.) because it is a destructive method. Therefore it is necessary to secure verified in-situ measurement technology that can be applied to operational monitoring or decommissioning plans. In this study, the applicability of cyclotron facilities was evaluated based on the evaluation algorithm derived from the Peak to Compton (PTC) method of in-situ measurement technology. And the reliability of the PTC method was verified through qualitative analysis and quantitative analysis. In the case of qualitative analysis, the analysis results of KAERI which has core technology are compared. To this end SAEAN and KAERI conducted field application tests on the front concrete shielding wall of the cyclotron facility at the same time. After removing the background spectrum from the measured spectrum the PTC method was applied to calculate the Q-value for the counting rate in the peak area per counting rate in the Compton continuum area was calculated. As a result the Q-values of SAE-AN and KAERI were 0.52 and 0.24 respectively, and the result of deriving activation distribution(β) by substituting this for the β-Q correlation equation was found that 14.78 and 12.94. As a result of evaluating the activation by the thickness of the shielding wall it was found that 89.1% (SAE-AN) and 91.9% (KAERI) of the total radioactivity were exist at a depth of 5 cm. And it was found that 97.7% and 99.05% of the total radioactivity exists at a depth of 10 cm. The relative error between SAE-AN and KAERI is 1.35%, indicating that the analysis results of the two institutions are highly consistent. A core drill was performed on the concrete shielding wall in the cyclotron facility for the technical verification of the quantitative analysis method. A core sample (6 cm in diameter, 10 cm in depth) was cut to a depth of 2 cm and analyzed in the laboratory. The activation distribution(β) was calculated based on the radioactivity level of each depth sample, and it was found to be 16.99. The relative error between the quantitative analysis and the on-site measurement results was 14.95% confirming that the accuracy is relatively high.