Uranium (U) is a hazardous material that can lead to both chemical and radiological toxicity, including kidney damage and health issues associated with radiation exposure. In South Korea. In Korea, where shallow weathered granitic aquifers are widespread, several previous studies have reported high levels of radioactivity in shallow groundwater. This ultimately led to the closure of 60 out of 4,140 groundwater production wells in South Korea. In this study, we examined aquifers currently dedicated to drinking water supply and investigated a dataset of 11,225 records encompassing 103 environmental parameters, based on the random forest classifier. This dataset comprises 80 physical parameters associated with the hydraulic system and 23 chemical parameters linked to water-rock interactions. Among the hydraulic parameters, the presence of a coarse loamy texture in the subsoil displayed a notable positive relationship with the concentration of uranium, implying that it plays a significant role in forming redox conditions for the leaching of uranium from host rocks. Fluorine (F), a major product of water-rock interaction in granitic aquifers, exhibited a positive correlation with the distribution of uranium concentrations. The positive relationship between F concentration and uranium levels suggests that the dissolved uranium originates from groundwater interacting with granites. In conclusion, our findings indicate that two key factors, namely the infiltration capacity of soil layers and the aqueous speciation in groundwater resulting from interactions with local solids, play important roles in determining uranium concentrations in granitic aquifers.

For safe and economical spent fuel management, assessing the integrity of the cladding, which is the first barrier to the escape of radioactive material, is very important. For the sake of risk assessment, it is essential to calculate the probability of failure of the spent fuel rods loaded inside the cask during the transportation or storage. However, due to the large amounts of calculations required, it is not practical to analyze every detail of the spent fuel rods and assemblies. This study presents a methodology to perform a cask-level analysis by sequentially simplifying the fuel rods and spent fuel assemblies for the calculation of fuel rod failure probability. A simplified single fuel rod model was generated by considering the material properties of a high burnup fuel rod stored in dry storage for approximately 5 years and the interfacial bonding conditions of the cladding tube. The simplified model produces the same deflection as the detailed model at the critical moment that produces a fracture plastic strain of 1%. The developed single fuel rod simplified model is assembled in a CE 16×16 configuration, and a methodology is presented in which the CE 16×16 assembly model is once again replaced by a simplified model with a cuboidal shape. Compression analyses were performed on each part of the CE 16×16 model to obtain isotropic property data, and a simplified model was created based on those data and the cross-sectional second moment values of the parts. A cask drop analysis was performed to validate the similarity of the CE 16×16 model and the simplified model by comparing important structural responses such as impact acceleration. The 20 simplified fuel assembly models and one detailed model were loaded into a cask to perform the drop analysis. For the detailed model, the impact acceleration was extracted for different loading positions and the corresponding impact load and pinch load were derived. The spring force and contact force corresponding to the pinch load were extracted by applying a Python script technique to extract the maximum value of them exerted on each fuel rod. The vulnerability of spent fuel rods to bending loads and the failure criteria were considered during the simplification process of a single fuel rod. From the extracted impact and pinch loads, the probability of failure of the spent fuel rods as a function of impact acceleration can be calculated.

Considering the domestic situation where all nuclear power plants are located on seaside, the interim storage site is also likely to be located on coastal site. Maritime transportation is inevitable and the its risk assessment is very important for safety. Currently, there is no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the immersion of transport cask. Previous studies show that the release rate of radionuclides contained in a submerged transport cask is significantly affected by the area of flow path generated at the breached containment boundary. Due to the robustness of a cask, the breach is the most likely generated between the lid and body of cask. CRIEPI investigated the effect of cask containment on the release rate of radioactive contents into the ocean and proposed a procedure to calculate the release rate considering the socalled barrier effect. However, the contribution of O-ring on the release rate was not considered in the work. In this study, test and analysis is performed to determine the equivalent flow path gap considering the influence of O-rings. These results will be implemented in the computational model to assess sea water flow through a breached containment boundary using CFD techniques to assess radionuclide release rates. To evaluate the release rate as a function of lid displacement, a small containment vessel is engineered and a metal O-ring of the Helicoflex HN type is installed, which is the most commonly used one in transport and storage casks. The lid of containment vessel is displaced in vertical and horizontal direction and the release rate of the vessel was quantified using the helium leak test and the pressure drop test. Through this work, the relationship between the vertical opening displacement and horizontal sliding displacement of the cask lid and the actual flow path area created is established. This will be implemented in the CFD model for flow rate calculation from a submerged transport cask in the deep sea. In addition, the compression of the O-ring causes very small gaps, such as capillaries. In these cases, Poiseuille’s law is used to calculate the capillary flow rate.

The Spent Nuclear Fuel (SNF) cladding serves as the first barrier that prevents the release of radioactive materials. It is very important to maintain cladding integrity in SNF management. It is known that the pinch load applied to the cladding can lead to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, a numerical analysis process was proposed to scientifically and systematically evaluate the fracture resistance of cladding with reoriented hydrides under pinch load. The mechanical behavior and fracture of the irradiated cladding under pinch load can be evaluated by Ring Compression Test (RCT). Under the stress field generated by RCT, the cracks propagate more easily through radial hydrides than circumferential hydrides. The δ-hydride which form within the α-zirconium matrix causes a large expansion strain due to the volume difference and voids form at the interface between the hydride and the zirconium matrix. Chan demonstrated that the load needed to form voids and separate the hard hydride precipitates from the Zr matrix is considerably lower than that which initiates brittle fracture of hydrides using a micro-cantilever test. Therefore, we propose a microstructure crack propagation analysis method based on Continuum Damage Mechanics (CDM) that can simulate fracture of hydride, zirconium matrix, and Zr/hydride interface. CDM is possible to simulate the hydride, zirconium matrix, and interface cracking in a continuum model based on cladding deformation. The RCT simulation model was constructed from the microscopic images of irradiated cladding. A pixel-based finite element model was created by separating the hydride, zirconium matrix, and interface using the image segmentation method on a morphology operation basis. The appropriate element size was selected for the efficiency of the analysis and crack propagation using CDM. The force-displacement curves and strain energy from RCT were compared and analyzed with the simulation results of different element sizes. The finalized RCT simulation model can be used to evaluate the fracture resistance of the irradiated cladding under the quantified pinch load and to establish the failure criterion of fuel rods under pinch load. The advantages and limitations of the proposed process are discussed.

Given the situation in the Republic of Korea that all nuclear power plants are located at the seaside, the interim storage facility is also likely to be located at seaside and the maritime transportation of Spent Nuclear Fuel is considered inevitable. The Republic of Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radionuclides from a submerged transportation cask in the sea. Therefore, there is a need to develop a technology that can assess the impact of immersion accidents and establish a regulatory framework for maritime transportation accidents. The release rate of radionuclides should be calculated from the flow rate through a flow path in the breached containment boundary. According to the cask design criteria, it is anticipated that even under severe accident conditions, the flow path size will be very small. Previous studies have evaluated fluid flow passing through micro-scale channel by integrating internal and external flows within and around a transport cask. As part of the evaluation, a comprehensive “Full-Field Model” incorporating external flow fields and a localized “Local-Field Model” with micro-scale flow paths were constructed. Sub-modeling techniques were employed to couple the flow field calculated by the two models. The aforementioned approach is utilized to conduct the evaluation of fluid flow passing through micro-scale flow paths. This study aims to evaluate fluid flow passing through micro-scale flow paths using the aforementioned CFD (Computational Fluid Dynamics) method and aims to code the findings. The Gaussian Process Regression technique, a machine learning model, is utilized for developing a mathematical metamodel. The selected input parameters for coding are organized and their respective impacts are analyzed. The range of these selected parameters is tailored to suit domestic environments, and computational experiments are planned through Design of Experiments. The flow path size is included as an input parameter in the coded model. In cases where the flow path size becomes extremely small, making it impractical to use CFD techniques for calculations, Poiseuille’s law is employed to calculate the release rate. In this study, a model is developed to evaluate the release rate of radionuclides using CFD and mathematical equations covering the whole possible range of flow path size in a lost cask in the deep sea. The model will be used in the development of a maritime transportation risk assessment code suitable for the situation and environment in Korea.

본 연구는 워터프론트에 조성된 공원을 “워터프론트 공원”이라는 유형으로 특정하고 워터프론트 공원 이 다른 공원과 구분되는 특징 및 가치를 고찰한다. 그리고 뉴욕시에서 2010년부터 2020년까지 공사 완료 개장한 워터프론트 공원들의 분석 연구를 통해 최신 워터프론트 공원에서 두드러지는 설계적 특 징 및 경향을 종합한다. 최신 워터프론트 공원의 설계경향 및 특징은 다음과 같다. 첫째, 다양한 경험이 가능한 수직적 구조의 제안 둘째, 배후도심과 공원구조의 유기적 연결을 통한 접근성 향상 셋째, 생태적 다양성의 증진과 원생경관의 회복 넷째, 기후 리질리언스 설계 도입 다섯째, 역사적 장소성 강조를 위한 역사 유산의 물리적 활용 마지막으로 자전거 전용도로와 공공 수상 교통을 위한 접안시설의 제공이다. 이를 바탕으로 본 연구는 앞으로 조성될 워터프론트 파크의 설계 및 계획을 위해 다음의 시사점을 제시 한다. 먼저 워터프론트 공원의 설계 및 조성 기법의 지속적 연구 둘째, 배후도심과의 워터프론트 공원의 강한 연계를 위한 행정적 지원 및 계획적 기획 세 번째, 워터프론트 공원을 통해 도시의 리질리언스를 증진할 수 있는 설계 연구의 필요성 마지막으로 개별적 워터프론트 공원의 조성에만 집중하는 것이 아 닌 도시/지역적 스케일을 아우르는 상위계획의 확립이다.

일인가구의 증가 SNS와 온라인을 통한 소통의 증대 등 점차 파편화 되는 현대사회에서 개인들이 직접 대면하고 소통할 수 있는 물리적 커뮤니티공간의 필요성은 점차 증가하고 있지만 이의 확충은 경제 및 개발 논리에 밀려 쉽게 이루어지기 어렵다. 본 연구는 커뮤니티공간의 부족함을 완화시키고 공평하게 향유할 수 있는 커뮤니티공간을 제공할 수 있는 대안으로 도시의 가로공간에 집중하고 도로가 가진 높 은 접근성을 활용하여 도로를 커뮤니티 활성화를 위한 커뮤니티공간으로 전환시키는 기법을 고찰하 고 시사점을 제공하고자 한다. 먼저 선행연구와 관련문헌을 통해 커뮤니티 활성화를 위한 주요 요소들 을 도출하여 분석의 틀을 만든다. 다음으로 뉴욕시에서 진행하고 있는 세 가지 프로그램인 뉴욕시 오픈 레스토랑, 뉴욕시 오픈 스트리트, 뉴욕시 플라자 프로그램의 사례연구 및 분석을 진행한다. 그리고 커 뮤니티공간에 커뮤니티를 위한 프로그램을 제공하는 방식도 함께 살펴본다. 분석 결과 세 개 프로그램 을 통해 가로공간을 전용하여 확보한 커뮤니티공간들이 커뮤니티의 활성화를 위한 다음의 요소들 안 전, 참여, 접근성과 보행성, 가로의 활력, 경제적 이득, 유연성과 적응력, 다양성과 포용성을 증진 시킬 수 있음을 확인하였다. 이를 통해 본 연구가 제시하는 시사점은 다음과 같다. 먼저, 지역 커뮤니티가 계 획 단계에서부터 커뮤니티공간의 조성, 운영, 관리의 주체가 되어야 한다. 두 번째, 시정부는 커뮤니티 들이 활용할 수 있는 프로그램을 개발, 운영하고 이를 제공하여야 한다. 세 번째, 커뮤니티와 그 주변 여건의 변화에 쉽게 적응하고 변화할 수 있는 가변적이면서도 능동적인 조성이 필요하다. 마지막으로 는 모든 커뮤니티공간에 적용될 수 있는 동일한 규격, 기준, 가이드라인의 제공이 필요하다.

Iron oxide (Fe2O3) nanoclusters exhibit significant potential in the biomedical and pharmaceutical fields due to their strong magnetic properties, stability in solutions, and compatibility with living systems. They excel in magnetic separation processes, displaying high responsiveness to external magnetic fields. In contrast to conventional Fe2O3 nanoparticles that can aggregate in aqueous solutions due to their ferrimagnetic properties, these nanoclusters, composed of multiple nanoparticles, maintain their magnetic traits even when scaled to hundreds of nanometers. In this study, we develop a simple method using solvothermal synthesis to precisely control the size of nanoclusters. By adjusting precursor materials and reducing agents, we successfully control the particle sizes within the range of 90 to 420 nm. Our study not only enhances the understanding of nanocluster creation but also offers ways to improve their properties for applications such as magnetic separation. This is supported by our experimental results highlighting their size-dependent magnetic response in water. This study has the potential to advance both the knowledge and practical utilization of Fe2O3 nanoclusters in various applications.

Digital fashion represents key technologies for the online environment, and it has been used as a new marketing strategy for the fashion industry. As consumer digital experience has been diversified, research on the effect on consumer attitudes and behavior toward digital fashion needs to be investigated. This study examines the effect of visual tactile and presence of VR fashion stores on consumer attitude via consumer surveys. Visual tactile and presence positively influence consumer experience which includes fantasy, feeling, and fun. In addition, these three experiences positively affect consumer attitude. The moderating effect of cyber motion sickness is discussed. Academic and practical implications for digital fashion are provided.

The damage ratio of Spent Nuclear Fuel (SNF) is a very important intermediate variable for dry storage risk assessment which require an interdisciplinary and comprehensive investigation. It is known that the pinch load applied to the cladding can lead to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, a sensitivity analysis was performed to evaluate the importance of the damage parameters that need to be calibrated for the simulation of zircaloy-4 cladding failure using computational mechanics. The simulation model was generated from a microscopic image of the cladding with hydride. The image segmentation method was used to separate the Zircaloy-4, hydride, and hydride- Zircaloy matrix interfaces to create a pixel-based finite element model. The ring compression test (RCT) was simulated because the resistance of the cladding under pinch load can be evaluated by this test. It was assumed that the damage starts with the formation and growth of voids or small cracks in the material, which grow and combine to form larger cracks, eventually leading to the complete fracture of the material. Therefore, the ductile damage criterion was applied to all materials to simulate crack formation and propagation. The sensitivity analysis was performed based on the design of experiments using L8 orthogonal array. The effects of five factors on the fracture resistance of hydrided cladding were quantified, and they are the fracture strains describing the damage initiation in zircaloy-4 matrix, hydride, and hydride-zirconium matrix, and yield stress and Young’s modulus for hydride-zirconium matrix. Information on those parameters are hardly available in literature and experimental data which enable the estimation of those are also very rare. It is planned to build a computational model which can accurately simulate the fracture behavior of hydrided cladding by calibrating significant fracture parameters using reverse engineering. The results of this study will help to figure out those significant parameters.

Considering the domestic situation where all nuclear power plants are located on seaside, the interim storage site is also likely to be located on coastal site. Maritime transportation is inevitable and the its risk assessment is very important for safety. Currently, there is no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the immersion of transport cask. Previous studies show that the release rate of radionuclides contained in a submerged transport cask is significantly affected by the area of flow path generated at the breached containment boundary. Due to the robustness of a cask, the breach is the most likely generated between the lid and body of cask. CRIEPI investigated the effect of cask containment on the release rate of radioactive contents into the ocean and proposed a procedure to calculate the release rate considering the so-called barrier effect. However, the contribution of O-ring on the release rate was not considered in the work. In this study, test and analysis is performed to determine the equivalent flow path gap considering the influence of O-rings. These results will be implemented in the computational model to assess sea water flow through a breached containment boundary using CFD techniques to assess radionuclide release rates. The evaluation of release rate due to container lid gaps has been performed by CRIEPI and BAM. In CRIEPI, the gap of the flow path was calculated from the roughness of the container surface without a quantitative assessment of the severity of the accident. In this work, to evaluate the release rate as a function of lid displacement, a small containment vessel is engineered and a metal Oring of the Helicoflex HN type is installed, which is the most commonly used one in transport and storage casks. The lid of containment vessel is displaced in vertical and horizontal direction and the release rate of the vessel was quantified using the helium leak test and the pressure drop test. Through this work, the relationship between the vertical opening displacement and horizontal sliding displacement of the cask lid and the actual flow path area created is established. This will be implemented in the CFD model for flow rate calculation from a submerged transport cask in the deep sea.

In the event of a loss of a SNF (spent nuclear fuel) transport cask during maritime transportation, it is essential to evaluate the critical depth at which the integrity of the cask can be maintained under high water pressure. SNF transport casks are classified as Type B containers and the integrity of of the containment boundary must be maintained up to a depth of 200 meters unless the containment boundary was breached under beyond-design basis accidents. However, if an intact SNF cask is lost at a depth deeper than 200-meter, release of radioactive material may occur due to breach of containment boundary with over-pressure. In this study, we developed a code for the evaluation of the pressure limit of SNF transport cask, which can be evaluated by inputting the main dimensions and loading conditions of cask. The evaluation model was coded as a computer module for ease of use. In the previous study, models with three different fidelities were developed to ensure the reliability of the calculation and maintain sufficient flexibility to deal with various input conditions. Those three models consisted of a high-fidelity model that provided the most realistic response, a low-fidelity model with parameterized simplified geometry, and a mathematical model based on the shell theory. The maximum stress evaluation of the three models confirmed that the mathematical model provides the most conservative results than the other two models. The previous results demonstrate that mathematical models can be used in the code of computer modules. In this study, additional models of transport cask were created using parametric modeling techniques to improve the accuracy of the pressure limit assessment code for different cask and situations. The same boundary conditions and loading conditions were imposed as in the previous simplified model, and the maximum stress results considering the change in the shape of the transport container were derived and compared with the mathematical model. The comparison results showed that the mathematical model had more conservative values than the simplified model even under various input conditions. Accordingly, we applied the mathematical model to develop a transportation container pressure limit evaluation code that can be simulated in various situations such as shape change and various situations.

Given the domestic situation, all nuclear power plants are located at the seaside, where interim storage sites are also likely to be located and maritime transportation is considered inevitable. Currently, Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radioactive waste from a submerged transportation cask in the sea. Therefore, secure technology is necessary to assess the impact of immersion accidents and establish a regulatory framework to assess, mitigate, and prevent maritime transportation accidents causing serious radiological consequences. The flow rate through a gap in a containment boundary should be calculated to determine the accurate release rate of radionuclides. The fluid flow through the micro-scale gap can be evaluated by combining the flow inside and outside the transportation cask. In this study, detailed computational fluid dynamic and simplified models are constructed to evaluate the internal flow in a transportation cask and to capture the flow and heat transfer around the transportation cask in the sea, respectively. In the future, fluid flow through the gap will be evaluated by coupling the models developed in this study.

In case a spent nuclear fuel transport cask is lost in the sea due to an accident during maritime transport, it is necessary to evaluate the critical depth by which the pressure resistance of the cask is maintained. A licensed type B package should maintain the integrity of containment boundary under water up to 200 m of depth. However, if the cask is damaged during accidents of severity excessing those of design basis accidents, or it is submerged in a sea deeper than 200 m, detailed analyses should be performed to evaluated the condition of the cask and possible scenarios for the release of radioactive contents contained in the cask. In this work, models to evaluate pressure resistance of an undamaged cask in the deep sea are developed and coded into a computer module. To ensure the reliability of the models and to maintain enough flexibility to account for a variety of input conditions, models in three different fidelities are utilized. A very sophisticated finite element analysis model is constructed to provide accurate response of containment boundary against external pressure. A simplified finite element model which can be easily generated with parameters derived from the dimensions and material properties of the cask. Lastly, mathematical formulas based on the shell theory are utilized to evaluate the stress and strain of cask body, lid and the bolts. The models in mathematical formula will be coded into computer model once they show good agreement with the other two model with much higher fidelity. The evaluation of the cask was largely divided into the lid, body, and bottom, bolts of the cask. It was confirmed that the internal stress of the cask was increased in accordance with the hydrostatic pressure. In particular, the lid and bottom have a circular plate shape and showed a similar deformation pattern with deflection at the center. The maximum stress occurred where the lid was in the center and the bottom was in contact with the body. Because the body was simplified and evaluated as a cylinder, only simple compression without torsion and bending was observed. The maximum stress occurred in the tangential direction from the inner side of the cylinder. The bolt connecting the lid and the body was subjected to both bending and tension at the same time, and the maximum stress was evaluated considering both tension and bending loads. In general, the results calculated by the formulas were evaluated to have higher maximum stresses than the analysis results of the simplified model. The results of the maximum stress evaluation in this study confirms that the mathematical models provide conservative results than the finite element models and can be used in the computer module.

As the saturation rate of temporary storage facilities for spent nuclear fuel increases, regulatory demands such as interim storage and permanent disposal of spent nuclear fuel are expected to begin in earnest. Considering the domestic situation where all nuclear power plants are located on the waterfront site, the interim storage site is also likely to be located on the waterfront site, and maritime transportation is one of the essential management stages. Currently, there are no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the sinking of transport container. Therefore, it is necessary to secure technology to properly reflect the domestic maritime transportation environment and to assess the impact of the sinking accident and to carry out safety regulations. To accurately calculate the releaser rate of radionuclides contained in a cask with breached containment boundary, the flow rate through the gap generated in the containment boundary should be calculated. The fluid flow through this gap which is probably in micro scale in most situations should be evaluated combining the fluid flow inside and outside the cask. In this study, a detailed computational fluid dynamics model to evaluate the internal fluid flow in the cask and a simplified model to capture the fluid flow and the heat transfer around the cask in the sea are constructed. The results for the large scale model are compared with the analytic formula for verification of heat transfer coefficient and they showed good agreements. The heat transfer coefficient thus found can be used in the detailed model to provide more realistic data than those obtained from assumed heat transfer coefficient around the surface of the cask. In the future, fluid flow through the gap between the lid and the body of the cask will be evaluated coupling the models developed in this work.

Barrier effect model developed by CRIEPI is used for the estimation of rate of radioactive material release from a transport cask submerged in the ocean. If the containment boundary of cask is broken in an accident during maritime transportation, the sea water comes into the cask cavity and the leaching of radioactive material occurs. If the release of radioactive material thorough the opening of the containment boundary of cask is less than the leaching rate of the radioactive material inside the cask, then the release rate is controlled by the saturation limit of the sea water inside the cask cavity. In this study, the release rate estimation using the barrier effect model is compared with the model used in other codes, such as MARINRAD. And by parameter study, important factors that affect the releaser rate are identified and prioritized. It is shown that the gap generated in the containment boundary is the key parameter that determine the release rate of the radioactive material and the leaching rate is the dominant parameter to determine the saturation time of the cavity sea water.

For economic and safe management of Spent Nuclear Fuel (SNF), it is very important to maintain the structural integrity of SNF and to keep the fuel undamaged and handleable. The cladding surrounding nuclear fuel must be protected from physical and mechanical deterioration. The structural evaluation of SNF is very complicated and numerically demanding and it is essential to develop a simplified model for the fuel rod. In this study, a simplified model was developed using a new cladding failure criterion. The simplified model was developed considering only the horizontal or lateral static load utilizing the cladding material properties of irradiated Zirclaoy-4, and applicability in horizontal and vertical drop impacts was investigated. When a fuel rod is subject to bending, a very complicated 3D stress state is generated within the vicinity of the pellet–pellet interface. A very localized stress concentration is observed in the area where the edges of the pellets contact the cladding. If the failure strain criteria obtained from the uniaxial tension test or biaxial tube test is applied, failure is predicted at the beginning stage of loading with premature through-thickness stress or strain development. The localized contact stress or strain is self-limiting and is not a good candidate for the cladding failure criteria. In this work, a new cladding failure criterion is proposed, which can account for the localized stress concentration and the through-thickness stress development. The failure of the cladding is determined by the membrane plus bending stress generated through the thickness of the cladding, which can be calculated by a process called stress linearization along the stress classification line. The failure criterion for SNF was selected as the membrane plus bending stress through stress linearization in the cross-sections through the thickness of the cladding. Because the stress concentration in the cladding around the vicinity of the pellet–pellet interface cannot be simulated in a simplified beam model, a stress correction factor is derived through a comparison of the simplified model and detailed model. The applicability of the developed simplified model is checked through horizontal and vertical drop impact simulations. It is shown that the stress correction factor derived considering static bending loading can be effectively applied to the dynamic impact analyses in both horizontal and vertical orientations.

The evaluation of the damage ratio of spent nuclear fuel is a very important intermediate variable for dry storage risk assessment, which requires an interdisciplinary and comprehensive investigation. It is known that the pinch load applied to the cladding can leaded to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, the failure resistance of Zircaloy-4 cladding against the pinch load is investigated using numerical simulations assuming the existence of radial hydrides. The simulation model is based on the microscopic images of cladding. A pixel-based finite element model was created by separating the Zircaloy-4 and hydride using the image segmentation method. The image segmentation method uses a morphology operation basis, which is a preprocessing method through erosion operation after image expansion to enable normal segmentation by emphasizing pixels corresponding to hydrides. The segmented images are converted into a finite element model by assigning node and element numbers together with corresponding material properties. Using the generated hydride cladding finite element model, several numerical methods are investigated to simulate crack propagation and cladding failure under pinch load. Using extended finite element (XFEM) models the initiation and propagation of a discrete crack along an arbitrary, solution-dependent path can be simulated without the requirement of remeshing. The applicability of fracture mechanical parameters such as stress intensity, J-integral was also investigated.

In this study, the removal efficiency of PFCs(perfluorinated compounds) in the GAC(granule activated carbon) process based on the superheated steam automatic regeneration system was investigated in laboratory scale and pilot-scale reactor. Among PFCs, PFHxS(perfluorohexyl sulfonate) was most effectively removed. The removal efficiency of PFCs was found to be closely related to the EBCT, and the removal efficiencies of PFOA(perfluorooctanoic acid), PFOS(perfluorooctyl sulfonate), and PFHxS were 43.7, 75, and 100%, respectively, under the condition of EBCT of 6 min. Afterward, PFOA, PFOS, and PFHxS exhibited the earlier breakthrough time in the order. After that, GAC was regenerated, and the removal efficiency of the PFCs before and after regeneration was compared. As a result, it was shown that the PFCs removal efficiency in the regenerated GAC process were higher, and that of PFOA was improved to 75%. The findings of this study indicate the feasibility of the superheated steam automatic regeneration system for the stable removal of the PFCs, and it was verified that this technology can be applied stably enough even in field conditions.

Extracellular vesicles (EVs) are nanovesicles that carry bioactive cargoes of proteins, lipids, mRNAs, and miRNAs between living cells. Their role in cellular communication has gained the attention of several research reports globally in the last decade. EVs are critically involved in sperm functions, oocyte functions, fertilization, embryonic development, and pregnancy. The review summarizes the state-of-theart of EVs research in the diagnostic and therapeutic (theranostic) potentials of the EVs during the pregnancy that might provide a solution for gestational disturbances such as implantation failure, maternal health problems, gestational diabetes, and preeclampsia. EVs can be found in all biological fluids of the fetus and the mother and would provide a non-invasive and excellent tool for diagnostic purposes. Moreover, we provide the current efforts in manufacturing and designing targeted therapeutics using synthetic and semi-synthetic nanovesicles mimicking the natural EVs for efficient drug delivery during pregnancy.