상선에 비해 잦은 변침을 하고 어획물로 인한 중량 및 무게중심의 변화와 같은 다양한 운항조건을 가진 어선의 경우, 조종성능 은 선박 운항 시에 매우 중요한 역할을 한다. 소형 어선의 사고는 2022년 기준 전체 해상 사고의 약 60%를 차지하며, 이는 부족한 조종성 능으로 인한 충돌과 좌초 사고가 주요 원인이다. 특히 10톤 미만의 소형선박에서 발생한 사고는 전체 사고의 약 65%를 차지하는데, 소형 어선의 조종성능 관련 기준이 부재하여 이를 정확히 평가하기엔 어려움이 있다. 이에 본 연구에서는 4.99톤급 소형 어선을 대상선으로 선 정하여 3D-CAD로 모델링 한 후, 상용 수치해석 프로그램인 STAR-CCM+를 활용하여 선박의 조종운동 시뮬레이션 환경을 구축하였다. 이 를 바탕으로 다양한 표준재화상태와 무게중심을 고려하여 10° / 10° 및 20° / 20° zigzag test와 35° turning test를 수행하였고, 선체 중량이 증가 함에 따라 변침성능이 감소하고 선회성능이 향상되는 경향을 분석하였다. 그 중, 만재출항과 부분만재입항 상태에서는 상대적으로 선회 성능이 부족한 결과를 확인하였다. 이를 바탕으로 소형선박의 안전한 운항을 위한 표준재화상태와 무게중심을 고려한 조종성능의 평가 및 그에 상응하는 표준화된 조종성능 평가 기준의 필요성을 제시하였다. 또한, 본 연구의 조종성능 평가 결과가 소형선박의 조종성능 평 가 기준 선정을 위한 기초자료로 활용될 수 있을 것으로 기대된다.

Research has been conducted on a wide variety of 3D printer circular fin heads. In this study, we proposed a sequence and method for a more efficient mesh study in the CFD model to calculate the Nusselt number of the circular fin head of an FDM 3D printer using the Taguchi method, sensitivity, and ANOVA. As a result, the CFD model to calculate the Nusselt number of the circular fin head of an FDM 3D printer has high sensitivity and contribution in the order of Base target mesh size, Prism layer number, and Prism layer thickness. We propose to increase work efficiency by performing mesh optimization in the order of factors with high sensitivity to level changes.

본 연구에서는 너클 라인이 다수 존재하면서 안팎 형상이 비대칭으로 설계된 특이점을 갖는 쌍동선의 자항성능을 예측하기 위 해 CFD 해석을 수행하였고, 해석 기법에 따른 차이를 파악하기 위해 MRF(Moving Reference Frame) 기법과 SDM(Sliding Mesh) 기법을 적용하 였다. MRF 기법을 적용한 경우에는 time step당 프로펠러를 1˚ 회전시켰고, SDM 기법의 경우 10˚, 5˚, 1˚씩 회전시키며 각 기법별 예측된 자 항성능을 비교하였다. 자항점 추정을 위한 몇 가지 프로펠러 회전수에서의 해석 결과 중 프로펠러의 토크는 기법에 따른 차이가 거의 없었 지만 추력 및 선체가 받는 저항은 MRF 기법보다는 SDM 기법을 적용했을 때 더 낮게, SDM 기법의 time step당 프로펠러 회전각이 작을수 록 높게 계산되었다. 선형 내삽을 통해 추정된 자항점의 프로펠러 회전수, 추력, 토크와 실선 확장법을 사용해 추정된 실선의 전달동력, 반 류 계수, 추력 감소 계수 및 프로펠러 회전수도 동일한 경향을 보였으며, 대부분의 자항효율은 반대의 경향을 보였다. 프로펠러 후류의 경 우 MRF 기법을 적용했을 때 정확도가 떨어졌고, SDM 기법의 time step당 프로펠러 회전각에 따라 표현되는 후류의 차이는 거의 없었다.

We analyzed the performance of hubless rim propellers based on the number of blades, maintaining a fixed pitch ratio and expanded area ratio, using computational fluid dynamics (CFD). Thrust coefficient, torque coefficient and efficiency according to the number of blades were analyzed. In addition, the pressure distribution on the discharge and suction sides of the blade was analyzed. As the advance ratio increases, the thrust coefficient decreases. The highest thrust was shown when the advance ratio was lowest. For the three, four, five and six-blades, the torque coefficient tended to decrease as the advance ratio increased. In the case of seven and eight-blades, the torque coefficient tended to increase as the advance ratio increased. The maximum efficiency was found when the advance ratio was 0.8. When the three-blade, it showed high efficiency at all advance ratios. A high pressure distribution was observed at the leading edge of the discharge blade, and a low pressure distribution was observed at the trailing edge. Applying a hubless rim-driven thruster with the three-blade can generate higher thrust and increase work efficiency.

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.

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.

Korea Hydro & Nuclear Power (KHNP) is currently developing a vertical concrete dry storage module for the dry storage of used nuclear fuel within nuclear power plants. This module is designed with a structure consisting of cylinders, which can block the ingress of external air, thereby preventing Chloride-Induced Stress Corrosion Cracking (CISCC). However, due to the presence of these cylinder structures, unlike conventional dry storage systems, it cannot directly dissipate heat to the external atmosphere, making thermal evaluation an important issue. The SF dry storage module being developed by KHNP is a massive concrete structure of approximately 20 m × 10 m × 7 m in size, employing a vertical storage system. To demonstrate the safety of such a large structure, there is no alternative to conducting experiments with scaled-down models. Furthermore, according to NUREG-2215 Section 5.5.4, it is explicitly mentioned that design-verification testing can be performed using scaled-down models. In this paper, a 1/4 scaled-down model was constructed to perform thermal performance verification experiments, and the effectiveness of this model was analyzed using Computational Fluid Dynamics (CFD) methods. The analysis results indicated that there was not a significant difference in terms of maximum concrete temperature and air outlet temperature. However, a considerable difference was observed in the canister surface temperature. Therefore, it is concluded that careful consideration of natural convection heat transfer is necessary for the full application of the scaled-down model.

Small hydropower systems have emerged as an attractive solution for areas with low head and flow rates, offering versatility for implementation in diverse locations such as rivers and wastewater treatment plants. This research specifically focuses on exploring the potential of small hydropower generation within wastewater treatment plants. Through the utilization of computational fluid dynamics (CFD) analysis, the study successfully predicted the torque and power generation capacity of the installed turbine. The analysis underscored the effective control of fluid flow achieved through careful turbine design, including considerations of blade shape and quantity. For instance, in the case of the Tancheon wastewater treatment plant, the study revealed the ability to generate a torque of approximately 7000 Nm, translating to an estimated power production of around 48.3 kW per hour. Ultimately, this research significantly contributes to evaluating the feasibility and viability of small hydropower generation within wastewater treatment plants.

A deep geological repository for disposal of high-level radioactive waste (HLW) consists of the canister, buffer material, and natural rock. If radionuclides leak from a disposal container, it can pass through buffer materials and rock, and move into the biosphere. Transport and migration of radionuclides in the rock differently were affected by the fracture type, filling minerals in the fracture, and the chemical and hydraulic properties of the groundwater. In this study, aperture distribution in fractured granite block was investigated by hydraulic test and CFD analysis. The fractured rock block (1 m × 0.6 m × 0.6 m), which is simulated as natural barrier, was prepared from Iksan, Jeollabuk-do. 9 test holes were drilled and packer system was installed to perform hydraulic test at the surface of fracture. 3D model simulated for aperture distribution of rock block was made using results of hydraulic test. And then, CFD analysis was performed to evaluate the co-relation between experiment results and analysis results using FLUENT code.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. During long-term storage, there are several mechanisms that could result in the degradation of spent nuclear fuels, and the temperature is the most important parameter to predict and estimate the degradation behaviors. Therefore, 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. Herein, Explicit CFD analysis model is introduced and validated by estimating the thermal hydraulic response of the dry storage system that is Dry Cask Simulator (DCS). Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to assess 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 is used to simulate DCS, and the test condition of 0.5 kW decay heat and 100 kPa helium pressure was investigated in this study. In case of peak cladding temperature (PCT), PCT from the experiment was 376 K while that of CFD was 374 K. It implies CFD simulation gives good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also show good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K.

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.

수많은 함정용 채프들은 폭발에 의해 확산되어 채프운을 형성하며, 채프운은 허위 레이더 반사 단면적을 생성하여 적의 레이더를 기만한다. 본 논문에서는 전산유체역학-이산요소법 단방향 연동 기법을 기반으로 공기 중에 분포하는 함정용 채프운의 시공간 분포 를 해석하는 수치적 프레임워크를 구축하고 바람의 방향과 속도, 채프 카트리지의 초기 각도와 폭발 압력이 채프운 분포에 미치는 영 향을 분석하였다. 채프운의 확산은 폭발에 의한 방사형 확산, 난류와 충돌에 의한 전 방향 확산, 낙하 속도 차이에 의한 중력 방향 확산 과 같이 세 단계로 구분되는 것을 확인하였다. 바람은 채프운의 평균 위치를 이동시켰으며, 항력에 의한 확산 효과는 나타나지 않았다. 카트리지 초기 각도에 따라 폭발에 의한 방사형 확산 방향이 달라졌으며, 각도가 지면과 수직에 가까울수록 더 넓게 확산되었다. 폭발 압력이 증가할수록 채프운은 더 넓게 확산되었으나 중력 방향으로는 분포 차이가 작았다.