• Nuclear Engineering and Technology
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• 제39권3호
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3635-3642
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• 2021

Natural circulation systems (NCSs) are extensively applied in nuclear power plants because of their simplicity and inherent safety features. For some passive natural circulation systems in floating nuclear power plants (FNPPs), the ocean is commonly used as the heat sink. Condensation induced water hammer (CIWH) events may appear as the steam directly contacts the subcooled seawater, which seriously threatens the safe operation and integrity of the NCSs. Nevertheless, the research on the formation mechanisms of CIWH is insufficient, especially in NCSs. In this paper, the characteristics of flow rate and fluid temperature are emphatically analyzed. Then the formation types of CIWH are identified by visualization method. The experimental results reveal that due to the different size and formation periods of steam slugs, the flow rate presents continuous and irregular oscillation. The fluid in the horizontal hot pipe section near the water tank is always subcooled due to the reverse flow phenomenon. Moreover, the transition from stratified flow to slug flow can cause CIWH and enhance flow instability. Three types of formation mechanisms of CIWH, including the Kelvin-Helmholtz instability, the interaction of solitary wave and interface wave, and the pressure wave induced by CIWH, are obtained by identifying 67 CIWH events.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1651-1664
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• 2023

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.

• 해양환경안전학회지
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• 제16권4호
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• pp.387-391
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• 2010

유엔해양법협약과 의제21의 17장에서 해양의 통합관리를 권고한 이후 미국, 캐나다, 호주 등 선진 해양국가들은 해양법 제정과 해양정책의 수립을 통해 통합적인 해양관리를 지향하고 있다. 그러나 이들 국가는 연방정부의 체제로서 해양관련 조직의 통합과 정책의 통합에 있어 일정한 한계를 보이고 있다. 즉, 이들 국가에서는 해양관련 주관부서가 없으며, 지방정부 관할해역을 연방정부의 해양정책에 포함하지 못하고, 민간부문 관련 산업의 육성을 직접적으로 언급하지 않고 있다. 이들 연방국가의 해양정책은 범부처 및 산업계를 포함한 위원회 등을 설립하여 관련 부처 및 지방정부의 다양한 이해관계를 조정하고 있다.

• KSII Transactions on Internet and Information Systems (TIIS)
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• 제12권3호
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• pp.1031-1046
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• 2018

Vehicular ad hoc network (VANET) is a dedicated network to connect vehicles without any centralized administration or infrastructure. The wireless access in vehicular environments (WAVE) protocol leveraging IEEE 1609/802.11p is widely implemented for VANETs. However, in congested traffic situation, the performance of the WAVE system degrades significantly due to serious collision, especially for safety related broadcast services on the control channel (CCH) interval due to the inherent drawback of its collision avoidance mechanisms called carrier sense multiple access with collision avoidance (CSMA/CA). In this paper, we propose a method that can decrease the number of frame collisions in CCH with a few modifications to the IEEE 802.11p protocol. In the paper, vehicles still employ CSMA/CA to compete for the channel access opportunity. However, by taking advantage of periodicity of synchronization interval, a two-state switching scheme introducing two new inter frame space (IFS) is proposed to reduce the number of competing vehicles substantially and as a result, the collision probability is significantly decreased. The simulation results demonstrate the superiority of the proposed method in packet collision rate.

• Nuclear Engineering and Technology
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• 제33권2호
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• pp.209-224
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• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• 대한안전경영과학회지
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• 제4권2호
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• pp.189-197
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• 2002

Composites have wide applications in aerospace vehicles and automobiles because of the inherent flexibility in their design lot improved material properties. Composite tubes in particular, are potential candidates for their use as energy absorbing elements in crashworthiness applications due to their high specific energy absorbing capacity and the stroke efficiency. Their failure mechanism however is highly complicated and rather difficult to analyze. This includes fracture in fibers, in the matrix and in the fiber-matrix interface in tension, compression and shear. The purpose of this study is to investigate the energy absorption characteristics of Gr/E(Graphite/Epoxy) tubes on static and impact tests. The collapse characteristics and energy absorption of a variety of tubes have been examined. Changes in the lay-up which increased the modulus increased the energy absorption of the tubes. Based on the test results, the following remarks can be made: Among CA15, CA00 and CA90 curves the CA90 tube exhibits the highest crush load throughout the whole crush process, and max load increases as interlaminar number increase. Among all the tubes type CC90 has the largest specific crushing stress of 52.60 kJ/kg which is much larger than other tubes.

• 한국지반공학회지:지반
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• 제6권2호
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• pp.47-54
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• 1990

사면안정해석의 정확도는 흙의 강도특성과 사면의 기하학적 조건의 파악정도와 해석방법 본말 의 정확도에 지배된다. 그리고 대부분의 사면안정해절방법은 컴퓨터 프로그랙으로작성되어 있다. 국내에서 실용되고 있는 프로그램의 정확성을 비교한 것이다. 비교적 간단한 조건의 각 프로그램으로 얻은 안전률의 차리는 5% 이내이다. 요면에 관한 정확한 타보입수가 가능한 파괴사례 연구에서도 유사한 결과를 얻었다. 그러나 컴퓨터가 찾아낸 임계파괴면과 실제 파괴면은 약간 상리한 경우도 있었다. 이 때에도 두 활동면을 해석하여 구한 안전률 값은 대차 없이 거의 동일하였다. 결국 국내에서 사용되고 있는 사면안정해석 컴퓨터 프로그램은 그 실용성이 인정된다고 할 수 있다.

• 한국항행학회논문지
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• 제10권3호
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• pp.226-234
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• 2006

항로표지는 해상에서의 인명과 재산의 안전을 확보하기 위하여 매우 중요한 항행지원시설이다. 하지만 이 시설을 설계하고 설치 및 관리 그리고 이용하는 과정에서 매체별 또는 관리주체별로 서로 상이한 위치정보를 포함하고 있다. 항로표지의 위치신뢰도에 가장 영향을 미치는 요인으로서는 조류의 방향과 수심의 함수가 생성하는 선회반경이 있고, 위치측정의 주요 수단인 수평육분의 및 dGPS로 인한 시점별 변수가 있다. 본 논문은 이러한 잠재오차를 수치적으로 해석하고, 요인별 영향력과 감소방안을 제시하였다.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.889-896
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• 2020

The vertically inverted U-tube steam generator (UTSG) is widely used in the pressurized water reactor (PWR). The reverse flow behavior generally exists in some U-tubes of a steam generator (SG) under both single- and two-phase natural circulations (NCs). The behavior increases the flow resistance in the primary loop and reduces the heat transfer in the SG. As a consequence, the NC ability as well as the inherent safety of nuclear reactors is faced with severe challenges. The theoretical models for calculating single- and two-phase flow pressure drops in U-tubes are developed and validated in this paper. The two-phase reverse flow characteristics in two types of SGs are investigated base on the theoretical models, and the effects of the U-tube height, bending radius, inlet steam quality and primary side pressure on the behavior are analyzed. The conclusions may provide some promising references for SG optimization to reduce the disadvantageous behavior. It is also of significance to improve the NC ability and ensure the PWR safety during some accidents.