• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.11
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• pp.2219-2227
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• 2002

In a nuclear power plant, reactor pressure vessel (RPV) is the primary pressure boundary component that must be protected against failure. The neutron irradiation on RPV in the beltline region, however, tends to cause localized damage accumulation, leading to crack initiation and propagation which raises RPV integrity issues. The objective of this paper is to estimate the integrity of RPV under hot leg leaking accident by applying the finite element analysis. In this paper, a parametric study was performed for various crack configurations based on 3-dimensional finite element models. The crack configuration, the crack orientation, the crack aspect ratio and the clad thickness were considered in the parametric study. The effect of these parameters on the maximum allowable nil-ductility transition reference temperature ($(RT_{NDT})$) was investigated on the basis of finite element analyses.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.832-845
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• 2024

Reactor cavity cooling systems (RCCSs) comprising passive safety features use the atmosphere as a coolant, which cannot be lost. However, their drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of the two types of passive RCCSs, namely RCCSs based on atmospheric radiation and atmospheric natural circulation, their safety must be evaluated, that is, they must be able to remove heat from the reactor pressure vessel (RPV) surface at all times and under any condition other than under normal operating conditions. These include both expected and unexpected natural phenomena and accidents. Moreover, they must be able to eliminate the heat leakage emitted from the RPV surface during normal operation. However, utilizing all of the heat emitted from the RPV surface increases the degree of waste heat utilization. This study aims to understand the characteristics and degree of passive safety features for heat removal by comparing RCCSs based on atmospheric radiation and atmospheric natural circulation under the same conditions. It was concluded that the proposed RCCS based on atmospheric radiation has an advantage in that the temperature of the RPV could be stably maintained against disturbances in the ambient air.

• Nuclear Engineering and Technology
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• v.55 no.5
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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.

• Proceedings of the KIPE Conference
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• 2007.11a
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• pp.202-205
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• 2007

현재 태양광, 풍력 및 연료전지 즉, 신재생에너지를 많이 사용하게 됨으로써 이를 계통과 병렬로 연결하여 쓰고 있다. 그러나 계통에 연결할 경우 여러 가지 문제가 발생할 수가 있는데 그중에서 대표적으로 단독운전을 들 수가 있다. 이를 방지하기 위한 검출기능은 시스템의 안전성 및 전기적 안전성 측면에서 매우 중요하다. 본 연구에서는 단독운전 검출을 위한 RPV 방식 및 제어알고리즘을 제안하였으며, 계통의 전압 위상을 기준으로 전류위상각도에 따른 주파수 변동 특성을 해석하였다. 해석 및 제안된 방법의 타당성은 EMTP 및 Matlab 을 이용하여 검증하였다.

• Journal of the KSME
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• v.34 no.7
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• pp.546-557
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• 1994

원격 조정식 비익체(이하 RPV : Remoteloted Vehicle)의 안정성을 향상시키고 조정을 간략화하기 위하여 사용되고있는 관성센서에 관해서 알아본다. 관성센서의 기본 원리는 뉴톤의 운동 제3 법칙인 관성의 법칙이고, 특징은 외부 측정기준을 필요로 하지 않은 점에 있으므로, 관성센서를 탐재한 RPV는 공중에서 운동 상태를 외부의 정보 없이 검토할 수 있다. 실제적으로 기계용으로 실용화되고 있는 센서는 관성항법장치(INS:Inertial Navigation System)라고 불리워지는 매우 고급자립형 장치로부터 자이로 컴파스로 불리워지는 방위 자이로와 자기 방위 센서를 조합한 방법까지 여러 가지가 있지만, 여기에서는 산업용 소형 RPV의 크기, 가격 및 입수성에서 이용이 가능하다고 생각되는 센서를 중심으로 원리, 종류 및 응용예를 설명한다.

• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.408-422
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• 1999

To reduce the fast neutron fluence which deteriorates the RPV integrity, additional shields were assumed to be installed at the outer core structures of the Kori Unit 1 reactor, and its reduction effects were examined. Full scope Monte Carlo simulation with MCNP4A code was made to estimate the fast neutron fluence at the RPV. An optimized design option was found from various choices in geometry and material for shield structure. It was expected that magnitude of fast neutron fluence would be reduced by 39% at the circumferential weld of the RPV, resulting in extension of plant lifetime by 4.6 EFPYs based on the criterion of PTS requirement It was investigated that the nuclear characteristics and thermal hydraulic factors at the internal core were only negligibly influenced by the installation of additional shield structure.

• The Transactions of the Korean Institute of Power Electronics
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• v.24 no.2
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• pp.105-110
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• 2019

This study proposed a reactive power variation (RPV) method equipped with positive feedback (PF) for detecting the islanding operation of distributed generation systems. The proposed RPV consists of the constant reactive power component and a certain reactive power term and uses the frequency deviation between the rated and the measured frequencies. The constant reactive power is injected from distributed generation system and power factor is to 0.9975 in grid-connected operation. PF is activated from generation of the frequency deviation and the injected reactive power is continuously increased due to PF when islanding occurs. Consequently, the increasing reactive power causes the point of common coupling frequency to deviate from the maximum/minimum threshold level. Performance of the proposed RPV is verified in a 1.7 kW T-type inverter, and the detection times are 53 and 150 ms.

• 제어로봇시스템학회:학술대회논문집
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• 1993.10a
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• pp.308-313
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• 1993

In this paper, the navigation computer design of RPV(remotely piloted vehicle) using GPS is investigated, and its hardware and software structures are described. The proposed hardware adopts the common PC configuration by using 5016A micro PC card and software is divided into several modules such as navigation module, guidance module and control module, etc. The performance of the navigation computer is verified through PILS(process in the loop simulation).

• Defense and Technology
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• no.1 s.61
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• pp.48-52
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• 1984

• 한국전산유체공학회:학술대회논문집
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• 2009.11a
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.