이 연구는 해외의 복수국적 제도를 살펴보고 한국의 제도와 비교해 보 며, 한국의 복수국적제도의 방향을 제시하고자 하였다. 지난 20년 사이, 전 세계에서 복수국적의 취득이 점점 더 많이 허용되고 있음에도 불구하 고 한국은 여전히 다중국적 취득을 부정적으로 생각한다. 이는 하나의 국적을 가진 일반인들에게 상실감과 역차별 논의를 제공할 수 있기 때문 이다. 그러나 이제 전통적인 국가의 개념이 약화된 만큼, 편리한 삶을 위 해 다국적 제도의 선택적 기능을 활용할 필요가 있다. 복수국적 취득 제 한은 정체성의 형성 과정 및 표현과 생활 영역 선택 등을 제한하여 물리 적인 거주·이전의 자유와 행복추구권에 대한 제한이 되며, 국적 보유의 자유에 대한 제한이 된다. 재외동포단체들은 국내 체류 동포를 국적으로 구분하지 말고 한국 사회의 주요 구성원으로 수용해 주기를 희망하였다. 이에 본 연구자는 현재 허용되고 있는 복수국적자의 연령을 하향하여 경 제활동 인구 증가 등 국가에 이익이 되는 정책을 제안하였다. 또한 설문 을 통해 차세대의 국민됨의 인식이 ‘국적’ 자체보다는 사회·정서적 교감 을 더 중요시한다는 것을 알 수 있었다. 이 결과는 미래세대의 인식 변 화에 맞게 국적 제도도 변화가 필요하다는 것을 의미하는 것이다.

In the decommissioning process of nuclear power plants, Ni-59, Ni-63 and Fe-55 present in radioactive waste are crucial radionuclides used as fundamental indicators in determining waste treatment methods. However, due to their low-energy emissions, the chemical separation of these two radionuclides is essential compared to others. Therefore, this study aims to evaluate the suitability of various pre-treatment methods for decommissioning waste materials by conducting characteristic assessments at each chemical separation stage. The goal is to find the most optimized pre-treatment method for the analysis of Ni-59, Ni-63 and Fe-55 in decommissioning waste. The comparative evaluation results confirm that the chemical separation procedures for Fe and Ni are very stable in terms of stepwise recovery rates and the removal of interfering radionuclides. However, decommissioning waste materials, which mainly consist of concrete, metals, etc., possess unique properties, and a significant portion may be low-radioactivity waste suitable for on-site disposal. Considering that the chemical behavior and reaction characteristics may vary at each chemical separation stage depending on the matrix properties of the materials, it is considered necessary to apply cascading chemical separation or develop and apply individual chemical separation methods. This should be done by verifying and validating their effectiveness on actual decommissioning waste materials.

In the decommissioning site of Korean Research Reactor 1&2 (KRR-1&2), according to Low and Intermediate-level Radioactive Waste Disposal Acceptance Criteria of the Korea Radioactive Waste Agency (WAC-SIL-2022-1), characteristics of radioactive waste was conducted on approximately 550 drums of concrete and soil waste for a year starting from 2021. Among them, 50 drums of concrete waste transported and disposed to Gyeongju LILW disposal facility at the end of 2022. For the remaining approximately 500 drums of concrete and soil waste stored on-site, they were reclassified into two categories: permanent disposal grade and clearance grade. This classification was based on calculating the sum of fractions (SOF) per drum for each radionuclides. The plan is to dispose of around 200 drums in the permanent disposal grade and about 300 drums in the clearance grade by the end of 2023. Since concrete and soil decommissioning wastes are generated in large quantities over a short period with similar origins, they were grouped within five drums as suggested by the acceptance criteria. Mixed samples were collected from each group and used for radionuclide analysis. When utilizing mixed samples, three distinct samples are collected and analyzed for each group. The maximum value among these three radionuclide analysis results is then uniformly applied as the radionuclide concentration value for all drums within that group. Radioactive nuclides contained in similar types of radioactive waste with similar origins can be expected to have some statistical distribution. However, There has been no verification as to whether the maximum value among the three mixed samples exists within the statistical distribution or if it deviates from this distribution to represent a different value. In this study, we confirmed characteristics of radionuclide concentration distribution by examining and comparing radionuclide concentration distributions for radioactive wastes drum grouped for nuclear characteristic among 50 concrete wastes drum disposed in year 2022 and 500 concretes & soils drum scheduled for disposal (clearance or permanent disposal) in year 2023. In particular, when comparing tritium to other nuclides, it was observed that the standard deviation for the distribution of maximum values was approximately 318 times larger.

Plasma torch melting technology has been considered as a promising technology for treating or reducing the radioactive waste generated by nuclear power plants. In 2006, IAEA announced that the technology is able to treated regardless of the type of target wastes. Because of the advantage, many countries have been funding, researching and developing the treatment technology. In this study, oversea plasma torch melting facilities for radioactive wastes treatment are reviewed. Also, plasma torch melting facility developed by KHNP CRI is briefly introduced.

Buffer materials are one of the engineering barrier components in high-level radioactive waste disposal facilities. Compacted bentonite has been known as the most suitable buffer material so far, and research is being conducted worldwide to determine the characteristics of suitable bentonite blocks in each country. Therefore, this study aims to compare and analyze various properties of different buffer material candidates, including thermal, hydraulic, and mechanical properties. Buffer material candidates for Korea disposal system, Kyungju Bentonite (KJ-I, KJ-II), and Bentonil- WRK were compared. The properties were compared and analyzed based on experimental and literature data. The data obtained from this report can be used to understand the behavior of buffer materials and assess whether they meet various criteria, such as temperature and saturation, and ultimately serve as an essential input variable database for safety evaluations of disposal systems.

A comparison and validation between the analysis and vibration test data of a nuclear fuel assembly were conducted. During the comparison and validation process, various parameters that govern the vibration behavior of the fuel assembly were determined, including nuclear fuel rod’s stiffness, spring constants of the dimple and spring of support structures, and damping coefficients. The calibration of the vibration analysis model aimed to find analysis parameters that can accurately simulate the vibration behavior of the test data. For calibration, power spectral density (PSD) diagrams were generated for both the measured signals from the test and the calculated signals from the analysis. The correlation coefficient between these two PSD plots was calculated. To find the analysis parameters, each parameter was defined as a variable with an appropriate range. Latin hypercube sampling was used to generate multiple sample points in the variable space. Analysis was performed for the generated sample points, and PSD plot correlation coefficients were calculated. Using the generated sample points and their corresponding results, a Gaussian Process Regression model was implemented for PSD plot correlation coefficients and the maximum PSD value. Based on the constructed surrogate model, the optimal analysis parameters were easily found without additional computations. Through this method, it was confirmed that the analysis model using the optimal parametes appropriately simulates the vibration behavior of the test.

Ring Tensile Test (RTT) is mainly performed for comparing tensile strength and total strain between nuclear fuel cladding specimens under various initial conditions. Through RTT, the loaddisplacement (F-D) curve obtained from the uniaxial tensile test can also be obtained. However, the Young’s modulus estimated from the gradient of the straight portion is much lower than general value of materials. The reasons include tensile machine compliance, slack in the fixtures, or elastic deformation of the fixtures and the tooling. Another reason is that the bending of the test part in the ring is stretched with two pieces of tools. Although the absolute value of the Young’s modulus is smaller than the actual value, it is applicable to calculate the ratio of the Young’s moduli of different materials, that is, the relative value. The Young’s modulus, or slope of the linear section, varies slightly depending on which location data is used and how much data is included. In order to obtain a more accurate ratio of Young’s moduli between materials using the RTT results, a post-processing method for the ring tensile test results that can prevent such human errors is proposed as follows. First, the slope of the linear section is obtained using the displacement and load when the load increase is the largest and the displacement and load of the position that is 95% of the maximum load increase. To replace the section where the ring-shaped specimen is stretched at the beginning of the F-D curve, a straight line equal to the slope of the linear section is drawn to the displacement axis from the position of maximum load increase and moved to the origin to obtain the final F-D curve for a RTT. Lastly, the yield stress uses the stress at the point where the 0.2% offset straight line and the F-D curve meet as suggested in the ASTM E8/E8M-11 “Standard test methods for tensile testing of metallic materials”. RTT results post-processing method was coded using FORTRAN language so that it could be performed automatically. In addition, sensitivity analysis of the included data range on the Young’s modulus was performed by using the included data range as 90%, 85%, and 80% of the maximum load increase.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. Thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. COBRA-SFS is a legacy code based on subchannel analysis code, and its fidelity is verified for evaluating the thermal analysis for licensing a dry cask system. Herein, thermal analysis result based on CFD and COBRA-SFS codes is compared and the Dry Cask Simulator (DCS) is assessed as a benchmark experiment in this study. Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to address the degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT was used to simulate DCS. Simulations were conducted in various decay heat and helium pressure inside the canister. In realistic conditions of decay heat and helium pressure of actual dry cask system, CFD and COBRA-SFS analysis results gave good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also showed good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K. In high decay heat and high pressure condition, however, CFD and COBRA-SFS underestimated peak cladding temperature than experimental results.

After the decision of the Wolsong unit 1 permanent shutdown (2019), spent fuel stored in the spent fuel bay (hereafter, SFB) should be transported to a dry storage facility (MACSTOR or Canister) in order to decommission Wolsong unit 1. Accordingly, KHNP has established a shipment schedule for damaged fuel of Wolsong Unit 1 and is trying to complete the shipment according to the schedule. Wolsong is equipped with transportation casks and dry storage facilities, but baskets need to be manufactured separately. In addition, license approval is required for baskets, transport cask, and dry storage facilities for legal grounds to contain, transport, and store damaged fuels. In this paper, the initial model, upgrade model, and automation model of encapsulation equipment planned to be introduced in Canada to handle PHWR’s damaged fuel were compared, and the optimal model was selected in consideration of KHNP’s planned spent fuel shipment schedule. The PHWR’s damaged fuel encapsulation system is a system developed the PHWR’s damaged spent fuel to be handled in the same way as the existing PHWR when storing it in the dry storage facility and loading a basket for capsulation into transport cask. At the Gentilly-2 nuclear power plant in Canada, a manually operated encapsulation system was used due to the low quantity of damaged fuel, which can be encapsulated two bundles a day, and this model is an initial model. In the case of Wolsong Unit 1, it has about 300 damaged fuels, so it takes about nine months to work when using the initial model. The upgrade model developed to improve work efficiency and reliability has increased work efficiency through some automation, but it would take about eight months to process the damaged spent fuels of Wolsong Unit 1, and this model has not yet been manufactured and applied. Lastly, the automation model changed the work location outside the SFB and automated drainage/drying operations. It is easy to maintain and replace consumables because the work is carried out by lifting the damaged fuel to a shuttle outside the SFB surrounded by a shielding chimney. Considering the reduction of drainage/drying time, it is possible to save about four times as much time as the initial model. That is, if the automation model is used, it is judged that the supply of Wolsong Unit 1 can be processed in about two months. However, in terms of license, initial model and upgrade model are expected to be easier and the period is expected to be shortened. However, if licensing is carried out as soon as equipment design is completed, it is believed that the period can be shortened by parallel equipment manufacturing and licensing. It is judged that the best way to comply with the target schedule is to select an automation model with excellent work performance, develop equipment, and proceed with licensing at the same time. Accordingly, KHNP is in the process of designing equipment with the aim of using the automation model to take out damaged fuel for Wolsong Unit 1.

This study aimed to suggest a suitable collar pattern by visually evaluating the appearance of the amount of collar drape by the starting position of the lapel line of a double-breasted tailored jacket using a 3d virtual fitting program. It created an avatar based on the mean size of women in their 20s (the 8th Size Korea) using clo network (double fastening: 10cm, collar width: 4.5cm, collar stand: 3cm, and lapel width: 8.5cm). The starting of the lapel twist line was waistline level, the 1/2 level of bustline and waistline, or bustline level, and collar laying amount was 4.5, 5.5, 6.5, or 7.5cm. It was evaluated by garment construction experts using 5, 6, and 4 items on the front, sides, and back, respectively. Descriptive statistics, F-test, Duncantest, and reliability analysis were conducted using SPSS 22. When collar laying amount was 6.5cm, it was best rated regardless of the starting point. Under waist line, when collar laying amount was 6.5cm, it was best rated regardless of the starting point. When collar laying amount was large, the collar’s outline length increased, resulting in unnecessary wrinkles from the neckline to the lapel, affecting the overall collar appearance. When collar laying amount was the smallest, the collar was lifted and the width was narrowed, exposing the seam connecting the collar and neckline. The length of the collar’s outline varied depending on collar laying amount, which was important to make the outline sit comfortably on the body.

Determining the size or area of a plant's leaves is an important factor in predicting plant growth and improving the productivity of indoor farms. In this study, we developed a convolutional neural network (CNN)-based model to accurately predict the length and width of lettuce leaves using photographs of the leaves. A callback function was applied to overcome data limitations and overfitting problems, and K-fold cross-validation was used to improve the generalization ability of the model. In addition, ImageDataGenerator function was used to increase the diversity of training data through data augmentation. To compare model performance, we evaluated pre-trained models such as VGG16, Resnet152, and NASNetMobile. As a result, NASNetMobile showed the highest performance, especially in width prediction, with an R_squared value of 0.9436, and RMSE of 0.5659. In length prediction, the R_squared value was 0.9537, and RMSE of 0.8713. The optimized model adopted the NASNetMobile architecture, the RMSprop optimization tool, the MSE loss functions, and the ELU activation functions. The training time of the model averaged 73 minutes per Epoch, and it took the model an average of 0.29 seconds to process a single lettuce leaf photo. In this study, we developed a CNN-based model to predict the leaf length and leaf width of plants in indoor farms, which is expected to enable rapid and accurate assessment of plant growth status by simply taking images. It is also expected to contribute to increasing the productivity and resource efficiency of farms by taking appropriate agricultural measures such as adjusting nutrient solution in real time.

본 연구에서는 강한 내열성을 가지는 비병원성 Bacillus atrophaeus, Bacillus subtilis, Geobacillus stearothermophilus 포자의 열 저항성을 분석하여 레토르트 식품 제조 시 직 접적인 멸균 여부 판정에 사용 가능성을 평가하고자 하였 다. B. subtilis 포자의 D121-value는 2.9±0.1분이었으며, Zvalue는 43.0±1.4oC로 나타났다. G. stearothermophilus 포 자의 D121-value는 4.3±0.1분이었으며, Z-value는 25.0±1.6oC 로 나타났다. B. atrophaeus 포자의 D121-value는 3.7±0.1분 이었으며, Z-value는 35.8±1.4oC로 나타났다. B. subtilis, G stearothermophilus와 B. atrophaeus 포자의 D121-value는 모 두 레토르트 식품 멸균 확인에 사용되는 C. botulinum 포 자의 D121-value 보다 높은 값을 나타내었다. 이러한 결과 를 종합하여 볼 때 레토르트 식품 멸균 시 병원성 포자형 성균인 C. botulinum 대신 B. subtilis, G. stearothermophilus, B. atrophaeus 포자를 사용할 수 있을 것으로 판단된다. 또 한 기존 세균발육 실험에 소요되는 13일보다 단시간인 2- 3일에 멸균 여부를 확인할 수 있을 것으로 판단된다.

본 연구는 야생조류 생물음향 탐지와 현장 조사의 결과 비교를 통해 효과적인 야생조류 조사 방법을 제시하는데 목적이 있다. 연구대상지는 제주 동백동산과 1100고지 습지였다. 생물음향 탐지 기간은 2020년 12개월간이었다. 생물음향 탐지는 시간당 1분씩, .wav, 44,100Hz 포맷으로 생물음향 측정장비(Song meter SM4)를 각 연구대상지에 설치하여 녹음하였다. 현장 조사는 Banjade et al. (2019)의 ｢제주 동백동산과 1100고지 습지의 조류군집 장기간 동향｣ 결과를 인용하였다. 연구결 과는 다음과 같다. 첫 번째, 생물음향 탐지 기법을 통해 확인된 조류상은 동백동산에서 29과 46종, 1100고지 습지에서 16과 25종이었다. 두 번째, 울음 빈도를 통해 나타난 종 구성은 동백동산에서 직박구리(33.62%), 섬휘파람새(12.13%), 동박새(9.77%) 순으로 나타났고, 1100고지 습지에서 큰부리까마귀(27.34%), 섬휘파람새(19.43%), 직박구리(16.56%) 순으 로 나타났다. 세 번째, 동백동산에서 실시한 현장 조사의 경우 2009년 39종, 2012년 51종, 2015년 35종, 2018년 45종이 조사되었고, 생물음향 탐지를 통해 46종이 확인되었다. 1100고지 습지에서 실시한 현장 조사의 경우 2009년 37종, 2012년 42종, 2015년 34종, 2018년 38종이 조사되었고, 생물음향 탐지를 통해 25종이 확인되었다. 종합적으로 현장 조사를 통해 동백동산에서 확인된 78종 중 43.6%인 34종, 1100고지 습지에서 확인된 47종 중 38.3%인 18종이 생물음향 탐지로 확인되었 다. 네 번째, 현장 조사에서 확인되지 않았지만, 생물음향 탐지를 통해 확인된 종은 동백동산에서 9과 12종, 1100고지 습지에서 3과 7종으로 대표적으로 소쩍새, 솔부엉이와 같은 야행성 조류, 쇠유리새, 울새와 같은 나그네새, 도래하는 개체수가 상대적으 로 적은 황여새, 솔잣새와 같은 겨울철새가 있다. 다섯 번째, 현장 조사에서 확인된 종 중에서 생물음향 탐지를 통해 확인되지 않은 종은 동백동산에서 18과 48종, 1100고지 습지에서 14과 27종으로 대표적으로 백로과, 수리과, 솔딱새과가 있다. 본 연구는 국내에서 생물음향 장비 운용을 통한 조류 연구가 활성화되고 있는 가운데 현장 조사와 생물음향 탐지를 병행했을 때 효과적인 조사를 실시할 수 있다는 기초 자료를 작성했다는 점에서 의미가 있다.