• International Journal of Concrete Structures and Materials
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• 제3권1호
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• pp.63-69
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• 2009

This paper provides the engineer with a simple design method dealing with situations arise where in-situ reinforced concrete joints are cast between hollow core units. Using finite element method, hollow core slabs with wide in-situ RC joints under point load and line loads are analysed. In addition, some important behavioural characteristics of the floor slab subjected to line and point loads are investigated. In-situ reinforced concrete joint causes reduction of load distribution for remote units because distance to the remote units from the point of load is increased, while the portion of load distribution carried by loaded unit increases. Also, it was turned out load distribution factors for point load and line loads are almost same. Finally, we suggest a simple analytical method, which can determine load distribution factors using normalized deflections by regression analysis for design purposes.

• Steel and Composite Structures
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• 제27권2호
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• pp.135-147
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• 2018

Quasi-static three-point bending tests on sandwich beams with expanded metal sheets as core were conducted. Relationships between the force and displacement at the mid-span of the sandwich beams were obtained from the experiments. Numerical simulations were carried out using ABAQUS/EXPLCIT and the results were thoroughly compared with the experimental results. A parametric analysis was performed using a Box-Behnken design (BBD) for the design of experiments (DOE) techniques and a finite element modeling. Then, the influence of the core layers number, size of the cell and, thickness of the substrates was investigated. The results showed that the increase in the size of the expanded metal cell in a reasonable range was required to improve the performance of the structure under bending collapse. It was found that core layers number and size of the cell was key factors governing the quasi-static response of the sandwich beams with lattice cores.

• 한국시뮬레이션학회:학술대회논문집
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• 한국시뮬레이션학회 2005년도 춘계학술대회 논문집
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• pp.89-92
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• 2005

Each 3D visualization program has its own different structure as for the purpose. This paper describes the design and development of an on-line 3D core data visualization program, $RocDis^{TM}$, for the nuclear simulator. It is possible to analyze the inside of the core status including neutron flux, relative power, moderator and fuel temperature in 3D distribution. Some of other essential information, axial flux distribution etc. could also display in 2D graphs. This program would be design, tuning and training for the simulator core model.

• Nuclear Engineering and Technology
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• 제34권1호
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Nuclear Engineering and Technology
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• 제17권1호
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• pp.53-67
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• 1985

중대 노심사고는 아직 Design Basis Accident에 포함되지 않고 있으나, 극히 적은 사고 확률을 가지는 반면 사고 후 영향이 큼으로해서 원자력발전소의 전반적 위험 평가에 중요한 요인중의 하나가 되고 있다. 중대 노심사고시 격납용기 손상에 관련된 물리현상들은 Steam Explosion, Debris Bed Coolability, Hydrogen Burning, Steam Spike와 Core-Concrete Interaction 등이며, 각각의 현상에 대한 좀 더 나은 이해를 위해 현재 이루어지고 있는 연구들에 대한 개략적 설명을 시도 하였다.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• 조명전기설비학회논문지
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• 제28권11호
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• pp.58-67
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• 2014

A grid-connected inverter used for renewable energy resources produces harmonic components in the switching frequency. To effectively reduce switching harmonic components, several types of filter are generally used in the output stage of the grid-connected inverter. Many research works on LCL filter design have been done to maintain the performance with low cost. However, it is not easy to make the filter design be economical and optimal due to the varying characteristic of magnetic core and redundancy design by experience. In this paper, a design method for a LCL filter is presented through the inductor optimization scheme in view of the size and cost when the inductor is manufactured using the magnetic core. The effectiveness is verified through tests using a 3kW grid-connected inverter by simulations and experiments.

• 한국실내디자인학회논문집
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• 제26권6호
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• pp.24-32
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• 2017

STEAM education is an effective teaching method to develop self-problem-solving skills through creative thinking. In order to revitalize STEAM education, various program models are being developed recently. The purpose of this study is to develop a STEAM education program based on the project-based learing method that includes the process of solving global environmental problems. The STEAM element was extracted by linking the zero energy house design with the middle school curriculum, and the STEAM education program was developed considering career activities. It was analyzed whether the developed program can improve STEAM core competence and job preparation ability. The education program was conducted for middle school students and the program was evaluated through questionnaires. In order to strengthen the STEAM competency, project-based learning method was applied and it was able to enhance the active problem solving ability of learners. In addition, opportunities for career experience could be provided through career exploration programs and various activities. Through this STEAM education program, it is expected to contribute to cultivating human resources with convergence knowledge and core competency.

• 한국안전학회지
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• 제34권3호
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• pp.102-110
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• 2019

This paper describes the work and results of the preliminary Probabilistic Safety Assessment (PSA) for a research reactor in the design phase. This preliminary PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, eight typical initiating events are selected regarding internal events during the normal operation of the reactor. Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. A total of 32 core damage accident sequences for an internal event analysis were identified and quantified using the AIMS-PSA. LOCA-I has a dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of a research reactor is estimated to be 7.38E-07/year. The CDF for the representative initiating events is less than 1.0E-6/year even though conservative assumptions are used in reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.