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.

This paper introduces a container loading problem and proposes a theoretical approach that efficiently solves it. The problem is to determine a proper weight of products loaded on a container that is delivered by third party logistics (3PL) providers. When the company pre-loads products into a container, typically one or two days in advance of its delivery date, various truck weights of 3PL providers and unpredictability of the randomness make it difficult for the company to meet the total weight regulation. Such a randomness is mainly due to physical difference of trucks, fuel level, and personalized equipment/belongings, etc. This paper provides a theoretical methodology that uses historical shipping data to deal with the randomness. The problem is formulated as a stochastic optimization where the truck randomness is reflected by a theoretical distribution. The data analytics solution of the problem is derived, which can be easily applied in practice. Experiments using practical data reveal that the suggested approach results in a significant cost reduction, compared to a simple average heuristic method. This study provides new aspects of the container loading problem and the efficient solving approach, which can be widely applied in diverse industries using 3PL providers.

본 연구는 컨테이너 재배용 중간목의 노지 생산기술 개발을 위하여 왕벚나무 3년생을 대상으로 전정, 시비종류·시비량 처리에 미치는 영향을 구명하기 위해 수행되었다. 노지에 정식 후 전정 유무(1.8m 이하 전정(P) 및 무전정(N.P))와 시비종류 및 시비량(유박비료 21.7, 43.3g/㎡/yr, 퇴비 1.0, 2.0 ㎏/tree/yr, 고형비료 40.5, 81.0g/㎡/yr)을 대조구를 포함하여 처리하였다. 묘목의 생장에 미치는 영향을 평가하기 위해 정식 2년 후 생장 및 상대생장량, 건물생산량, 품질지수 및 양분특성을 비교‧분석하였다. 묘고 및 근원경 상대생장량은 무전정 처리보다 전정처리에서 유의적(p≤0.000)으로 높았으며, 전정을 실시한 고형비료 81.0g/㎡/yr 처리에서 가장 높은 생장량을 보였다. 본 연구결과 왕벚나무 접목 3년생을 노지에 정식한 후, 1.8m 이하 전정을 실시하고 고형비료 81.0g/㎡/yr을 처리하여 2년 재배 시, 고품질의 컨테이너 재배용 왕벚나무 중간목이 생산되는 것으로 판단된다.

본 연구에서는 전장이 400m인 24,000TEU급 컨테이너선박을 대상으로 선박 내 빌지펌핑의 성능에 대한 케이스 스터디를 수행하 였다. 본 연구의 대상인 24,000TEU급 컨테이너선박의 빌지시스템의 경우 선급의 규칙에 맞게 설계되었지만, 선박 내 설치되어 있는 빌지펌 프의 정격유량 및 최대유량 조건에서도 SOLAS Reg.II-1/35-1의 2 m/s 요건을 만족시키지 못하였다. 특히 1번 ~ 4번 화물창에 대해, 해수로 가득차 있다고 가정한 상태에서 해수를 모두 배출하는 동안에 빌지 주관에서의 평균유속을 계산할 결과, 2번 화물창, 3번 화물창 및 4번 화물창은 평균유속이 2 m/s 미만으로 기준에 적합하지 않은 것으로 나타났다. 본 연구에서는 이를 해결하기 위해 2번, 3번 및 4번 화물창의 150A 빌지 지관을 200A 배관으로 교체하여 계산을 수행하였으며, 그 결과 화물창 내의 해수를 모두 배출하는 동안의 빌지 주관에서의 해 수평균유속이 각각 2.479m/s , 2.476,m/s 및 2.459m/s 로 기준을 만족시키는 것을 알 수 있었다.

Test of the operating characteristics and energy saving performance of a container cooling system that reduces the operating energy of a refrigeration system using a loop thermosyphon heat exchanger that removes heat by temperature difference between outdoor and indoor was performed. As a result of the experiments, when the loop thermosyphon and the refrigeration system were operated simultaneously, the refrigeration system operated intermittently by reducing the heat load. As the temperature difference between indoor and outdoor increased, the operating time of the refrigeration system decreased and the energy efficiency rate increased. Energy efficiency rate showed a tendency to increase with increasing temperature difference, and the predicted correlation of energy efficiency rate using the performance of the loop thermosyphon heat exchanger and the refrigeration system was relatively consistent with the experimental value.

High Integrity Container (HIC) made of polymer concrete was developed for the efficient treatment and safe disposal of radioactive spent resin and concentrate waste in consideration of the disposal requirements of domestic disposal sites. Permission for application of Polymer Concrete High Integrity Container (PC-HIC) to the domestic nuclear power plants (NPPs) has been completed or is under examination by the regulatory agency. In the case of 860 L PC-HIC for very-low-level-waste (VLLW) or low low-level-waste (LLW), the application of four representative NPPs has been approved, and the license for extended application to the rest NPPs is also almost completed. A licensing review is also underway to apply 510 L PC-HIC for intermediate and low-level-waste (ILLW) to representative nuclear power plants. In order to handle and efficiently store and manage PC-HICs and high-dose PCHIC packages, a gripper device that can be remotely operated and has excellent safety is essential, and the introduction of NPPs is urgent. The conventional gripper device developed by the PC-HIC manufacturer for lifting test to evaluate the structural integrity of PC-HIC requires a rather wide storage interval due to its design features, and does not have a passive safety design to handle heavy materials safely. In addition, work convenience needs to be reinforced for safety management of high radiation work. Therefore, we developed a conceptual design for a gripper device with a new concept to minimize the work space by reflecting on-site opinions on the handling and storage management conditions of radioactive waste in NPPs, and to enhances work safety with the passive safety design by the weight of the package and the function of checking the normal seating of the device and the normal operation of the grip by the detector/indicator, and to greatly improves the work efficiency and convenience with the wireless power supply function by rechargeable battery and the remote control function by camera and wireless monitoring & control system. Through design review by experts in mechanical system, power supply and instrumentation & control fields and further investigations on the usage conditions of PC-HICs, it is planned to facilitate preparations for the application of PC-HIC to domestic NPPs by securing the technical basis for a gripper device that can be used safely and efficiently and seeking ways to introduce it in a timely manner.

The Deep Borehole Disposal (DBD) method has various advantages, such as minimizing the use of site area and corrosion of the disposal container and improving long-term structural safety. However, it is necessary to review the problems that may occur in various technologies related to the emplacement and retrieval of the disposal container and the sealing of the borehole. Therefore, the purpose of this study is to evaluate the structural integrity of an emplacement and retrieval device (hereinafter, the disposal container connecting device) of a DBD container. The disposal connecting device was evaluated according to ANSI 14.6 and NUREG-0612 standards. The allowable stress should be less than the yield strength under the load condition of 3g. The length of the disposal container connecting device was about 2,900 mm, the diameter was 406 mm, and the weight was about 1.2 tons. In addition, 10 disposal containers weighing up to 2.2 tons were handled. The disposal container connecting device was made of stainless steel, and the maximum operating temperature was about 300°C. For structural evaluation, ABAQUS finite element analysis program was used. The analysis model was modeled only 1/2 part considering symmetry condition. The analysis model was modeled using 410,431 nodes and 344,119 solid elements. Three times load was applied to the weight of the disposal container. Axisymmetric conditions were applied to the symmetrical surface of the disposal container, and vertical restraints were applied to the upper lifting lugs. A surface-to-surface contact condition was applied to the part where the contact occurred. As a result of the analysis, the greatest stress was generated at the part supported by the clamp at the disposal container connector at 168.9 MPa. In the lugs and pins connecting the guide and the connecting device, a stress of 530.1 MPa was generated by shearing. In the bolts of the disposal container connecting device, a stress of 498MPa was generated and the safety margin was 1.73. A stress of 486.1 MPa was generated in the disposal container connecting device, and the safety margin was the smallest 1.16. As a result of the analysis, all components of the disposal container connecting device showed a safety margin of 1.16 or more at the maximum operating temperature and satisfied the allowable stress.

Currently, the most widely accepted disposal concept for long-term isolation of high level radioactive waste including spent nuclear fuels is to disposal in a deep geological repository designed and constructed with multiple barriers composed of engineered and natural barriers so that the waste can be completely isolated in a stable deep geological environment. In this concept, an important consideration is the heat generated from the waste due to the large amount of fission products present in the high level waste loaded in the disposal container. For safe and complete isolation of high level radioactive waste in the deep geology, the disposal concepts that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. In this paper, the deposition hole configuration and the decay heat dissipation area (surface area) of disposal container were considered for the efficient thermal management in the deep geological disposal concept. Heat transfer through the waste form, its container and surrounding components and the rock will be mainly by conduction. Heat transfer by radiation and convection can be negligible after backfilling. When considering heat conduction, according to Fourier’s law, if the thermal conductivity of the repository components is the same, the greater the heat dissipation area and the adjacent temperature gradient, the greater the conduction effect. Therefore, rather than the conventional concept of loading 4 PWR spent fuel assemblies per disposal container and placing one disposal container in a deposition hole, it is better to load one assembly per disposal container and place 4 disposal containers in a deposition hole. In this case, it was found that the disposal area could be reduced through efficient thermal management. Considering this thermal management method as an alternative to the concept of deep geological disposal, additional research is needed.

To reduce the air pollution from maritime activities, which is proven to have severe impacts on the worldwide environment and human health, many international regulations have been established. Therefore, an effective political strategy and a complete inventory of emissions are needed to control atmospheric ship pollution and comply with these international standards. The purpose of this study is to calculate the amount of emission in three operating modes (cruising, maneuvering, and hoteling) for some main pollutants emitted from container ships and trucks operating in Daesan port in Korea based on bottom-up methodology. The results showed that the volume of air pollution of about 6,500 tons from container ships and 1,455 tons from container trucks were emitted in Daesan port area. Also, a total of 4 billion won (about 3.6 billion won from container ships, and about 400 million won from container trucks) was estimated.

As the decommissioning of nuclear power plants increases, there is an increasing interest in the amounts of radioactive waste. Especially, the radiation dose limit for packaging of radioactive wastes shall not exceed 2 mSv·h−1 and 0.1 mSv·h−1 on contact and at 2 m, respectively in South Korea. The DEMplus provides various environmental geometry and all properties such as materials, absorptions, and reflections and the estimation of the radiation dose rates is based on the radiation interactions of the designed 3D geometry model. With the consideration of the radiation dose rate by using DEMplus and its strategy of packaging plan, the radiation shielding was optimized and estimated in this paper. The modular shielded containers (MSC) with shielding inserted were used for radioactive wastes that require shielded packaging. In order to verify the accuracy of the estimated radiation dose rate by using DEMplus, the estimated results were compared with those obtained using MicroShield. The trends of the estimated radiation dose rates using DEMplus and the estimation of MicroShield were similar to each other. The results of this study demonstrated the feasibility of using DEMplus as a means of estimating the radiation dose limit in packaging plan of the radioactive waste.

During the normal operation boron concentrates and spent resins are generated. The boron concentrates are treated by concentrated waste drying system (CWDS) and results in fine powder form. The solidification or application of high integrity container (HIC) is required for the disposal of the dried boron concentrates. The spent resin is stored in storage tank after the water treatment. The spent resin also requires solidification or application of HIC to satisfy the waste acceptance criteria (WAC) in Korea. The solidification process requires periodic validation. The repeated validation and complicated process hesitates the practical application. The application of HIC offers various advantages, including flexible free standing water requirement, higher waste loading compared to solidification, and simple process. The polymer concrete (PC), which is a primary component for PC-HIC exhibits good material stability. The expected transportation mechanism of nuclide in the PC-HIC are 1) diffusion by concentration, 2) permeation by pressure, and 3) capillary suction when considering the disposal condition. Since the PC-HIC effectively prevents the intrusion of neighboring water and volume of free standing water is lower than 1%, it seems that diffusion by concentration is the primary transportation mechanism. In this study, the property of PC is investigated based on Cl ion diffusion test to evaluate the material reliability. The results indicate that PC exhibits superior stability compared to ordinary portland cement. In addition, the reliable life time of PC is estimated base on the element transportation phenomena.

For transport containers for radioactive wastes, a drop test should be performed at a height of 0.3– 1.2 m on a rigid target depending on the weight as a normal condition in the regulation. In the drop test, a strain gauge is commonly used to measure the local strain, and the position of the strain gauges is determined by the experiences of the engineer in advance of the test. For this reason, the strains can be measured at only predetermined points. The DIC (Digital Image Correlation) method using highspeed cameras can be used to measure the change in strain over the region of interest. In addition, it is possible to measure effectively even in areas with high strain gradients that are difficult to measure with strain gauges. Therefore, the DIC method can measure the strain change according to time over the entire load path. When the drop test of the transport container is performed, the impact load is delivered through the lower corner fittings-corner posts-upper corner fittings-lids. In this study, white spray was sprayed on these main load path, and black speckles were created on the spayed surface to trace the rigid motion of speckles. The images taken during the drop test can be used to create a strain field over region of interest.

In this study, a drop analysis of metallic disposal containers for radioactive wastes is performed according to accident scenarios at the disposal site. The weight of the disposal container is about 8 tons, and the ingot-type wastes are loaded in the disposal container. To simulate the floor of the disposal site as the impact target, the reinforced concrete pad is modeled. High impact energy of the disposal container due to their heavy weight and high drop height causes excessive deformation and failure of the concrete target having relatively weak strength. Dynamic growth of cracks due to such failures causes penetration and delamination of concrete. Since the impact force delivered to the container strongly depends on the failure of the concrete pad, it is important to properly simulate the failure of the concrete in the drop analysis. A material erosion method can be used to simulate the concrete failure. In the case of applying erosion based on the finite element method (FEM), the element is deleted when the element exceeds a certain criterion, which causes material and energy loss problem. To solve this problem, mesh-free methods such as smoothed particle hydrodynamics (SPH) can be commonly used, but the mesh-free method has the disadvantage of incurring high numerical cost. Therefore, an adaptive method combining SPH and FEM-based SOLID elements is used for concrete target modeling to simulate excessive deformation and failure of the concrete target. In the adaptive coupling method of SPH and SOLID, the concrete target is first modeled as a solid element. When the damage of concrete exceeds the failure criterion, the solid element is eroded and the SPH element replacing the solid element is activated. Since the activated SPH element continues to participate in the impact, the problem of loss of materials and energy can be effectively solved. In this way, analysis results consistent with actual physical phenomena can be obtained.

The mechanical safety of the container designed according to the IP-2 type technology standard was analyzed for the temporary storage and transportation of Very-Low-Level-Waste (VLLW) for liquid occurring at the nuclear facilities decommissioning site. The container was designed and manufactured as a composite shielding container with the effect of storing and shielding liquid radioactive waste using High Density Polyethylene (HDPE) and eco-friendly shielding material (BaSO4) with corrosion and chemical resistance. The main material of the composite shielding container is HDPE and BaSO4, the material of the cover, cage and pallet is SUS304, and the angle guard is elastic rubber. The test and analysis requirements were analyzed for structural analysis of container drop and lamination test. As test requirements for IP-2 type transport containers should be verified by performing drop and lamination tests. There should be no loss or dispersion of contents through the 1.2 m high free-fall drop and lamination test for a load five times the amount of transported material. ABAQUS/Explicit, a commercial finite element analysis program, was used for structural analysis of the drop and lamination test of the transport and storage container. (Drop test) It was confirmed that the container was most affected when it falls from a 45-degree slope. Although plastic deformation was observed at the edge axis of the cover, it was evaluated that the range of plastic deformation was limited to the cover and cage, and stress within the elastic limit occurred in the inner container. In the analysis results for other falling direction conditions, it was evaluated that stress within the elastic limit was generated in the inner container except for minor plastic deformation. In the case of on-site simulation evaluation, deformation of the inner container and frame due to the drop impact occurred, but leakage and loss of contents, which are major evaluation indicators, did not occur. (Lamination test) The maximum stress was calculated to be 19.9 MPa under the lamination condition for a load 5 times the container weight, and the maximum stress point appeared at the corner axis of the pallet. The calculated value for the maximum stress is about 10%, assuming the conservative yield strength of SUS304 is 200 MPa. It was evaluated that stress within the limit occurred. In the case of on-site simulation evaluation, it was confirmed that there was no container deformation or loss of contents due to the load.

Radioactive waste disposal facility in Korea, radioactive waste packaged in 200 L drums is placed in a concrete disposal container and disposed of at an underground silo type (cave) disposal facility. At this time, the disposal container cover is seated on the top of the disposal container, and if the disposal container and the cover are not completely combined, the container cover is raised up from the top of the disposal container, so safety problems may occur when stacking the disposal container. Therefore, various methods exist to secure a margin for the pure height inside the disposal container. The disposal container cover only covers the upper surface of the container to shield radiation, and structural performance is not required. Therefore, the method of processing the cover, such as a method of making the cover of the disposal container thin, is the easiest method to apply. In this study, a method to reduce the thickness of the cover of a concrete disposal container was devised, and structural performance under usability conditions such as lifting and seating was analyzed. In addition, the disposal container cover has a reinforced concrete form in which dissimilar materials (concrete and steel) are combined, an integrated analysis was performed to secure the reliability of the analysis results for this, and the analysis results were described. It was found that the proposed disposal container cover structure can improve usability by reducing the stress concentration phenomenon.

The structural safety of prototype transport and storage containers for very low-level radioactive liquid waste was experimentally estimated for its localization development. Transport containers for radioactive liquid waste have been researched and developed, however, there are no standardized commercial containers for very low-level radioactive waste in Korea. In this study, the structural safety of the designated IP-2 type container capable of transporting and temporarily storing large amounts of very low-level liquid waste, which is generated during the operation and decommissioning of nuclear power plants, was demonstrated. The stacking and drop tests, which were conducted to determine the structural integrity of the container, verified that there was no external leakage of the contents in spite of its structural deformation due to the drop impact. This study shows the effort required for the localization of the technology used in manufacturing transport and storage containers for very low-level radioactive liquid waste, and the additional structural reinforcement of the container in which the commercial intermediate bulk container (IBC) external frame was coupled.