본 논문에서는 컨테이너선의 선형 최적 설계 자동화와 관련하여 연구한 내용과 결과를 정리하였다. 컨테이너선은 일반적으로 프루우드 수 0.26 근처에서 운항하는 선박으로 이 속도에서 운항하는 선박 전용 선형 최적 설계 자동화를 구현하기 위하여 최적화 알고리 즘, 선형 변경 알고리즘, 선박 성능 예측 알고리즘, 자동화 알고리즘 그리고 반복적 계산 기법을 적용하여 컨테이너선의 선형 최적 설계 자동화가 가능한 수치해석 컴퓨터 프로그램을 개발하였으며, HOTCONTAINER라고 명명하였다. 본 연구에서는 선형 최적 설계를 위한 설 계 변수의 적절한 선정을 위하여 민감도 분석 알고리즘을 개발하여 적용하였다. 개발된 선형 최적 설계 자동화 알고리즘의 신뢰성과 실 선 적용성을 파악하기 위하여 세계적으로 다양한 연구가 진행된 컨테이너 선박인 KCS 선박을 대상 선박으로 하여 선형 최적 설계 자동 화 수치해석을 수행하여 그 결과물로써 최적 선박을 도출하고, 대상 선박과 최적 선박의 조파저항과 파계 그리고 파고를 비교하였다. 결 론적으로 최적 선박이 대상 선박과 비교하여 조파저항이 47.63% 감소한 것을 볼 수 있었으며, 배수량과 접수 표면적은 각각 0.50%, 0.39% 감소한 것을 볼 수 있었다.

본 연구의 목적은 컨테이너 육묘 시스템을 활용한 참외 접목 묘의 안정적인 생산 가능성을 평가하는 것이었다. 이를 위해, 컨테이너 육묘 시스템과 고온 조건의 플라스틱 온실에서 육묘 한 접수와 대목, 접목묘의 생육을 비교 분석하였다. 접목활착 후 육묘 환경에 따른 참외 접목묘의 생육과 묘소질을 0일, 4일, 7일, 11일, 14일째에 비교하였다. 컨테이너 육묘 시스템에서 는 주야간 온도를 25/20°C, 상대습도를 70%로 설정하여 재 배기간 동안 안정적으로 유지하였으며, 플라스틱 온실 내의 주야간 평균온도는 28.1/15.4°C로 주야간 온도차(DIF)가 더 크게 나타났다. 조사기간 동안 참외 접목묘의 초장은 플라스 틱 온실 육묘 처리구에서 컨테이너 육묘 시스템 처리구보다 더 길게 나타났다. 참외 접목묘 조직의 충실도는 지상부 건물 중을 초장으로 나누어 계산하였다. 육묘장에서 접목한 묘는 접목 후 7－10일 경과하여 활착이 완료되고 초장이 10cm 내 외일 때 출하하여 정식에 이용되게 된다. 본 연구에서 접목활 착 후 7일째에 컨테이너 육묘 시스템에서 재배된 묘의 충실도 는 44.9±2.64mg/cm으로 나타났으며, 플라스틱 온실 육묘 처 리구에서는 24.4±1.56mg/cm로 나타났다. SPAD 평균은 플 라스틱 온실 육묘에서 30.5, 컨테이너 육묘 시스템에서 41.1 로 측정되었다. 이러한 결과는 컨테이너 육묘 시스템의 활용 이 고온기 또는 저일조 시기와 같은 육묘 환경에서도 고품질 모종을 안정적으로 생산할 수 있는 것을 확인하였고, 인공광 을 이용한 육묘 시스템의 활용 범위가 앞으로 더 확대될 것으 로 기대된다.

Truck no-show behavior has posed significant disruptions to the planning and execution of port operations. By delving into the key factors that contribute to truck appointment no-shows and proactively predicting such behavior, it becomes possible to make preemptive adjustments to port operation plans, thereby enhancing overall operational efficiency. Considering the data imbalance and the impact of accuracy for each decision tree on the performance of the random forest model, a model based on the Borderline Synthetic Minority Over-Sampling Technique and Weighted Random Forest (BSMOTE-WRF) is proposed to predict truck appointment no-shows and explore the relationship between truck appointment no-shows and factors such as weather conditions, appointment time slot, the number of truck appointments, and traffic conditions. In order to illustrate the effectiveness of the proposed model, the experiments were conducted with the available dataset from the Tianjin Port Second Container Terminal. It is demonstrated that the prediction accuracy of BSMOTE-WRF model is improved by 4%-5% compared with logistic regression, random forest, and support vector machines. Importance ranking of factors affecting truck no-show indicate that (1) The number of truck appointments during specific time slots have the highest impact on truck no-show behavior, and the congestion coefficient has the secondhighest impact on truck no-show behavior and its influence is also significant; (2) Compared to the number of truck appointments and congestion coefficient, the impact of severe weather on truck no-show behavior is relatively low, but it still has some influence; (3) Although the impact of appointment time slots is lower than other influencing factors, the influence of specific time slots on truck no-show behavior should not be overlooked. The BSMOTE-WRF model effectively analyzes the influencing factors and predicts truck no-show behavior in appointment-based systems.

Port operational efficiency is considered as one of the most important competitive factors and plays a critical role in the port development all over the world, especially container ports. Haiphong Port, which is in the northern of Vietnam, is planned to become one of the national and regional ports. To do this objective, it is important to analyse the operational efficiency of its container terminals. The paper aims to comparatively analyse the operational efficiency of 16 container terminals in Haiphong Port from 2016 to 2022 by basic and Malmquist DEA models. With 112 observations collected and calculated in R software, DEA models have five inputs (container yard area, number of quay crane, berth draft, berth length, labour force) and one output (annual cargo throughput). Consequently, Hai An, Tan Vu, and Vip Greenport are more efficient terminals over the 7-year period, whereas Transvina and MIPEC have lower efficiency. Paper contributions are the literature review about port operational efficiency and references to propose resolutions in next author’s research as well as masterplans to develop Vietnam seaport’s system. Besides, the limitations are discussed as the number of observations and environmental factors in ports.

To safely dispose of highly radioactive spent resin and concentrate waste generated through nuclear power plant operations, it is essential to meet the physicochemical properties requirements of the packages and ensure the accuracy and reliability of radiological characteristics determination. Both spent resin and concentrate are packaged in high-integrity containers (HICs) after drying and are homogeneous waste products generated in the primary system and liquid radioactive waste treatment system. Meeting the physicochemical properties requirements does not appear to be difficult. However, to achieve reliable radiological characterization of high-integrity container packages, it is necessary to take a representative sample and perform accurate radiological analysis. Therefore, this paper discusses the methodology for evaluating the radionuclide inventory of high radioactive resin and concentrate packages, as well as the essential element technology and considerations. For relatively high radioactive resin and concentrate packages, the radionuclide inventory for each package should be evaluated with high reliability through direct radiological analysis of the representative samples collected for each package. This can contribute to the efficient operation of radioactive waste disposal facilities. Radionuclide-specific concentrations directly analyzed for each package will be managed in a database. As analytical data accumulates and direct measurements of high-integrity container package such as the radwaste drum assay system (RAS) become feasible, statistical techniques such as correlation analysis between easy-tomeasure (ETM) nuclides and difficult-to-measure (DTM) nuclides can lead to the development of efficient and reasonable indirect evaluation methods, such as scaling factor and the mean activity concentration method. As for the element technology, a remote representative sampling technique should be developed to safely and effectively take representative samples of highly radioactive materials, including granulated or hardened concentrate waste. Considerations should also be given to determining the sample quantity representing each package, as well as establishing radiation calibration and measurement methods appropriate to the radiation levels of the representative samples.

At domestic nuclear power plant, concrete containers are stored to store waste generated before waste acceptance criteria (WAC) was established. Concrete container store concentrated waste liquid and waste resin. In order to disposal radioactive waste to a disposal site, it is necessary to conduct a characteristic evaluation inside the waste to check whether it satisfies the WAC. Two types of concrete containers are stored: round and square. The round type is filled with one 200-liter drum, and the square type is filled with four 200-liter drums. In the case of a round shape, the top lid is fastened with bolts, so it is possible to collect samples after opening the top lid without the need for additional equipment. However, in the case of a square shape, there is no top lid, and concrete is poured to cure the lid, so the separate equipment for characteristic evaluation is required. It is necessary to install a workstation for sample collection on the top of the concrete container, equipment for coring the top of the concrete container, and a device to prevent concrete dust scattering. Currently, the design of equipment for evaluating the characteristics of concrete containers has been completed, and equipment optimization through mock-up test will be performed in the future.

Concentrated effluent and spent ion exchange resins (IERs) from nuclear power plants (NPPs) were generated prior to the establishment of a disposal facility site and waste acceptance criteria have been temporarily stored at the NPPs because their suitability for disposal has not been confirmed. In particular, at the Kori Unit 1, which was the first to start the commercial operation in South Korea, the initially generated concentrated effluent and IERs are repackaged in large size of concrete containers and stored without provided regulation standard. The concentrated effluent is package as cementitious form in 200 L drums and repackaged in concrete containers, case of the IERs were solidified or dehydrated and repackaged in round concrete container. In this study, we review and propose a disposal plan for concentrated effluent and IERs repackaging drums that have not been confirmed to be suitable for disposal from the first operating nuclear power plant, Kori Unit 1, 2. First, the concentrated effluent was stored in four 200 L drums respectively, and then, it was again stored in concrete container and which was poured on top using grouted concrete. Therefore, the process was required by cutting concrete container for extracting the internal drums at first. Internal radioactive waste should be crushed to the suitable waste criteria and solidified, finally disposal in to the polymer concrete high integrity container (PC-HIC). IER was repackaged and disposal in square type of 200 L concrete drums respectively covered the cap. So, extracting the internal drums should be extracted after removing the cap of external concrete container. Cement solidification drums can be crushed and re-solidified or disposed in the PC-HIC. Stored IER after dehydrated can be disposal in PC-HIC. In conclusion, the container was used as a package that repackaging the concentrated effluent and IER was separated into two different types of waste depending on the level of contamination of radioactivity, the polluted area is disposed of as radioactivity contamination or the unspoiled area will be treated as self-disposal waste.

This paper described a method for analyzing the structural performance of a metal container used for disposing radioactive waste generated during the decommissioning of a nuclear power plant, and numerical analysis results of a method for reinforcing the container. The containers to be analyzed were those that can be used in near-surface and landfill disposal facilities scheduled to be operated at the Gyeongju radioactive waste disposal facility. Structural reinforcement of the container was performed by lattice reinforcement, column reinforcement, and bottom plate reinforcement. Accordingly, a total of 14 reinforcement cases were modeled. The external force causing damage to the container was set equivalent to the impact of a 9-m fall, accounting for the height of the vault at the near-surface disposal facility. The reinforcement methods with a high contribution to the structural performance of the container were concluded to be lattice and column reinforcements.

Licensing for the application of the Polymer Concrete High Integrity Container (PC-HIC) to nuclear power plants has been completed or is in progress. Approval for the expanded application to all domestic nuclear power plants has been completed to utilize the 860 L PC-HICs for the 2nd stage surface repository, and the regulatory body is reviewing the license application to use the 510 L PCHICs for the 1st stage underground repository in the representative nuclear power plants. The 860 L PC-HICs, which have been licensed for all domestic nuclear power plants, will be used for safe storage management and disposal of low-dose dried concentrate waste and spent resin, and a total of 100 units is expected to be supplied to representative nuclear power plants that have been licensed first. The 510 L PC-HICs are planned to be used for underground disposal of high-dose spent resin and dried concentrate waste. Prior to the application of PC-HICs to nuclear power plants and disposal to the repository, it is necessary to establish realistic and reasonable requirements through close consultations between waste generator and disposal operators to ensure the suitability for disposal of PC-HIC packages and to carry out disposal delivery and acceptance work. Since the Polymer Concrete High Integrity Container (PC-HIC) has long-term integrity of more than 300 years and the barrier does not temporarily collapse, spent resin and dried concentrate waste, which are radioactive wastes to be solidified, can be disposed of much more safely in PC-HIC packages than solidified types. Acceptance criteria for the PC-HIC packages should be prepared fully reflecting the advantages of PC-HIC, and quality assurance methods for physical/chemical/radiological characterization results based on the Waste Certification Program (WCP) should be supported. In addition, infrastructure should be secured for safe transportation, handling, and storage of the PC-HIC packages. In this paper, we have tried to find a reasonable acceptance criteria, quality assurance method, and infrastructure level according to the dose and disposal conditions of PC-HIC packages.

Currently, KHNP-CRI has developed 100 kW plasma torch melting facility to reduce the amount of radioactive waste in nuclear power plant. Plasma torch melting technology uses electric arc phenomena like lightning to melt the target material at a high temperature of about 1,600°C. The technology is applicable to treatment for various types of waste such as combustible, non-combustible and mixed wastes. The volume reduction ratio by the technology is respectively expected to be about 1/60 of combustible wastes and about 1/5 for non-combustible wastes. It is important to discharge the melt without problems in the melting technology. In general, molten slag has properties such as high viscosity and quick solidification. Because of the properties, when discharging into slag container, the final product is accumulated like a mountain. To improve this problem, there is three suggestions; 1) rotation of the slag container, 2) vibration of the slag container, and 3) heating of the slag container.

In case of damaged spent fuels, it would require additional treatment for their transportation and storage to capture the radioactive fission products in a defined space. The canning container for the damaged spent fuels is one way to seal the radioactive fission products inside the container. In the Post Irradiation Examination Facility (PIEF) of KAERI, the Quiver container has been introduced for canning damaged spent fuels from Westinghouse Sweden. The main container body has been manufactured for particle-tightness of spent fuel. In addition, drying equipment is being prepared for gas-tightness of spent fuel. The drying equipment can remove water and fill the inert gas inside the container. Before drying inside the container, we evaluated the volatile fission products inventory because volatile fission products could be released during the drying process. Despite assuming highly conservative hypotheses for the inventory remaining in damaged fuel rods, the amount that could be released during the drying process was less and dose rate levels around the evacuation piping system were low.

코로나 19 팬데믹 및 기후 변화 등으로 전 세계적으로 필수적인 생필품과 자원의 품귀 이슈가 지속해서 발생하고 있다. 이러한 현상을 극복하고자 교역량의 수요가 갑자기 증가하였으며 이 결과 컨테이너선의 운임이 대폭 상승하였다. 컨테이너선의 크기 변화는 1960 년대 1,500TEU(twenty-foot equivalent unit)를 시작으로 2021년에는 24,400TEU로 대형화가 진행되고 있다. 컨테이너 적재 능력의 향상은 라싱브 릿지 구조의 대형화와 긴밀하게 연관되어 있고, 안전한 컨테이너 고박 및 항해 시 발생하는 다양한 외력 하중에 안전한 구조설계를 해야 한다. 현재 주요 선급에서는 라싱브릿지 구조 안전성을 평가할 수 있는 구조해석 기반의 지침서를 배포하고 있으나, 허용기준 및 평가 방 법이 달라서 설계 시 엔지니어들에게 혼선을 주고 있다. 본 연구에서는 결과에 영향을 줄 가능성이 큰 주요 변수들(모델링 범위, 오프닝 고 려 여부, 메쉬 크기) 변화에 따른 강도 변화 특성을 정리하였다. 이 결과를 바탕으로 저자들은 합리적인 구조해석 기반 평가에 대한 검토사 항을 제안하였고, 추후 선급 기준 개정 시 참고가 될 수 있을 것으로 기대한다.

본 연구는 후박나무 용기묘를 대상으로 전광, 35%, 55%, 75% 차광수준에 따른 생장 특성 및 생리적 반응을 알아보고자 실시하였다. 후박나무 묘목은 차광 시 상대적으로 지상부의 생장특성을 보여주는 지표들이 높은 경향을 보였으며, 전광에서는 뿌리와 관련된 지표인 뿌리건중비, 뿌리생장에 서 유의적으로 높은 경향을 보였다. 또한, 전광에서의 낮은 광보상점과 높은 순양자수율은 이미 후박나무가 내음성을 가지고 있고, 광이 부족한 환경에 노출된 차광처리구들에서 암호흡량의 감소, PIABS 및 SFIABS의 증가로 광합성 기구의 활력을 효율적으로 증가시키는 것을 알 수 있다. 한편, 전광은 지상부 비율의 증가, 낮은 T/R율과 가장 높은 순광합성속도를 보여 광합성으로 합성되는 탄수화물이 지상부 뿐만 아니라 지하부의 생장 역시 증가시킨 것을 알 수 있는 결과였다. 따라서 후박나무 용기묘는 전광조건에서 생장 및 생리적 특성이 가장 양호한 경향을 보였다.

본 연구는 헛개나무 용기묘를 생산하기 위한 적정 차광수준을 밝히고자 수행되었다. 헛개나무 1년생 용기묘를 대상으로 4가지 차광처리(0%, 35%, 55%, 75%)에 따른 용기묘의 간장, 근원경, 건중량, 묘목품질지수, 엽록소 함량 등의 반응 특성을 조사하였다. 연구 결과, 헛개나무 용기묘의 근원경, 잎 건중량, 뿌리 건중량은 차광처리구 간에 통계적인 유의성이 나타나지 않았으나(p≥0.05), 간장, 줄기 건중량, 엽록소 함량은 차광처리에 의해 55%와 75%의 차광처리에서 유의적인 차이를 나타내며 증가하였다(p<0.001). 간장과 줄기 건중량의 증가로 헛개나무 용기묘의 H/D율과 T/R율은 증가하였으며 묘목품질지수는 낮게 나타났다. 따라서, 헛개나무 1년생 용기묘 생산을 위한 적절한 차광수준은 묘목품질지수가 가장 높게 나타난 35%로 판단된다. 본 연구결과는 우량한 헛개나무 용기묘 생산을 위한 기초자료로 활용될 수 있을 것으로 기대된다.

The recent increase in earthquake activities has highlighted the importance of seismic performance evaluation for civil infrastructures. In particular, the container crane essential to maintaining the national logistics system with port operation requires an exact evaluation of its seismic response. Thus, this study aims to assess the seismic vulnerability of container cranes considering their seismic characteristics. The seismic response of the container crane should account for the structural members’ yielding and buckling, as well as the crane wheel’s uplifting derailment in operation. The crane’s yielding and buckling limit states were defined using the stress of crane members based on the load and displacement curve obtained from nonlinear static analysis. The derailment limit state was based on the height of the rail, and nonlinear dynamic analysis was performed to obtain the seismic fragility curves considering defined limit states and seismic characteristics. The yield and derailment probabilities of the crane in the near-fault ground motion were approximately 1.5 to 4.7 and 2.8 to 6.8 times higher, respectively, than those in the far-fault ground motion.