• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.734-738
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• 2012

A comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals is established in accordance with the United States Nuclear Regulatory Commission Regulatory Guide 1.20 Revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results at the inner barrel assembly top plate in the reactor. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at Inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals.

• Plant Journal
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• v.12 no.4
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• pp.37-43
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• 2016

Power system which provides electricity to the accident mitigation load for nuclear power plant should be verified to maintain the proper voltage level under the various loading and source conditions. For this purpose, it was needed to collect the voltage data of safety related buses during operation of the Reactor Coolant Pump(RCP) motor and Component Cooling Water Pump(CCWP) motor, respectively, under the certain loading condition of the plant. The data (such as, voltage, current, power factor) collected from actual measurement were used to modify the existing ETAP model and then the reanalysis was conducted to simulate the testing conditions. Through these actual measurement and analysis, it ensures that the existing electrical system analysis including assumptions and methods was conducted properly. Finally, the voltage of safety related buses was not dropped below the acceptable level, and the discrepancy between two results was within the limit.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.9
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• pp.643-657
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• 2009

This paper introduces the current nuclear experimental research activities in KAERI, the unique nuclear research institute in Korea, and the universities in Korea to solve and assess the issues which have been faced in the design of new reactors such as APR1400, SMART, GEN-IV reactors as well as fusion reactor. Also the experimental evaluations of safety for operating nuclear plants have been presented. The nuclear thermalhydraulic experiments performed in such organizations are classified the fundamental test, the separated effect test, and the integral effect test with ATLAS and SNUF. Introduction is deployed according to institutes. Finally, the future works and the direction of research voyage in the nuclear thermal-hydraulic field were suggested.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.7
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• pp.767-774
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• 2012

A fluidic device (FD) has been adopted in the safety injection tanks (SITs) of APR1400. A flow control mechanism of the FD was used to vary the flow regime in the vortex chamber corresponding to the SITs water level. The flow regime in the vortex chamber has a different pressure loss from low to high in accordance with the SITs water level. Nitrogen at the top of the SIT could be released owing to inertia of discharge flow when changing from a high flow rate to a low flow rate. This phenomenon is important to design improvement perspective because it can affect the performance of the FD. This paper shows a result of a preliminary numerical study to obtain the transient data related to air release in the flow turn-down period using a two-fluid free-surface model provided from ANSYS CFX 13.0. In conclusion, there is no significant effect on the performance of the FD, though a small quantity of air is released during the flow turn-down period.

• Nuclear industry
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• v.37 no.11
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• pp.30-38
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• 2017

완전 통합 디지털 계측제어 시스템을 갖춘 APR1400 신형경수로 초호기 신고리 3호기가 지난해 연말부터 가동되고 있다. 본고는 이 최첨단 원자로의 종합 계측제어 시스템을 상세히 설명하고 있다. 지금 세계의 많은 원전들은 디지털 계측제어 시스템의 확보를 위해 확실하고, 효율적이며, 무엇보다 안전한 기술 혁신을 추구하는데 매진하고 있다. 기술이 밀집되어 복잡하지 않고, 모듈 방식의 개인별 계기반과 전산화된 공정, 그리고 스마트 경보 발령 시스템 등을 갖춘 첨단 제어실은 더욱 간편하고 안전한 업무 환경을 제공해 주고 있다. 신고리 3,4호기의 성공 사례를 통해서 독립형 계측제어 시스템을 성공적으로 통합시킬 수 있는 혁신적인 방법이 확인됨으로써 이제는 종합 디지털 계측제어 시스템을 적시에, 안전하고, 경제적으로 실용화할 수 있다는 것을 보여주고 있는 것이다.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.9
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• pp.599-605
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• 2015

The revised U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20, rev.3 requires the evaluation of the potential adverse effects from pressure fluctuations and vibrations on piping and components for the reactor coolant, steam, feedwater, and condensate systems. Detailed vibration analyses for the systems attached to the steam generator are very difficult, because these piping systems are very complicated. This paper suggests a screening method for the flow-induced vibration of acoustic resonances and pump-induced vibration of the piping systems attached to the steam generator in order to conduct the APR1400 comprehensive vibration assessment program. This paper seeks to address the areas such as potential vibration sources, and methods to prevent the occurrence of acoustic resonances and pump-induced vibration of piping systems attached to the steam generator, for conducting the APR1400 comprehensive vibration assessment program. The screening method in this paper will be used to estimate the flow-induced vibration of the piping systems attached to the steam generator for the APR1400.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.1
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• pp.1-8
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• 2011

Advanced SITs of the evolutionary PWRs have the advantage that they can passively control the ECC water discharge flow rate. Thus, the LPSI pumps can be eliminated from the safety injection system owing to the benefit of the advanced SITs. In the present study, a passive sealing plate was designed in order to overcome the shortcoming of the advanced SITs, i.e., the early nitrogen discharge through the stand pipe. The operating principle of the sealing plate depends only on the natural phenomena of buoyancy and gravity. The performance of the sealing plate was evaluated using the VAPER test facility, equipped with a full-scale SIT. It was verified that the passive sealing plate effectively prevented the air discharge during the entire duration of the ECC water discharge. Also, the major performance parameters of the advanced SIT were not changed with the installation of the sealing plate.