• Structural Engineering and Mechanics
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• v.2 no.2
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• pp.157-171
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• 1994

A general approach to the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced form earthquake. The dynamic responses such as fuel assembly deflected shapes and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed.

• International Journal of Fluid Machinery and Systems
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• v.1 no.1
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• pp.121-139
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• 2008

For mechanical drive steam turbines, the investigation results of corrosion fatigue phenomena in the transient zone are introduced, including basic phenomena on expansion line and actual design and damage experience. These results were analyzed from the standpoint of stress intensity during the start of cracking. In order to resolve such problems, preventive coating and blade design methods against fouling and corrosive environments are developed. Detailed evaluation test results are given for coating performance using a unique test procedure simulating fouling phenomena and washing conditions. Finally, the results of the successful modification of internals and on-line washing results on site are introduced.

• Journal of Advanced Marine Engineering and Technology
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• v.31 no.1
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• pp.84-94
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• 2007

As a reactor coolant pump(RCP) is operated in the nuclear power system for a long time. so its surface is continuously contaminated by radioactive scales. In order to perform regular or emergency repair about RCP internals a special decontamination process should be used to reduce the radiation from the RCP surface by means of chemical cleaning. In this study, applicable possibility in chemical decontamination for RCP was investigated on the various materials. The STS 304 showed the best electrochemical properties for corrosion resistance than other materials. However, the pitting corrosion was slightly generated in both STS 415 and STS 431 with the increasing numbers of cycle and intergranular corrosion were sporadically observed. The size of their pitting corrosion and intergranular corrosion were also increased with increasing cycle numbers.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1992.10a
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• pp.81-86
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• 1992

원자력발전소 1차 계통 내의 건전성 감시를 위한 설비로는 음향누설 감시계 통(Acoustic Leak Monitoring System: ALMS), 금속파편 감시계통(Loose Parts Monitoring System: LPMS) 및 원자로내부구조물 진동감시계통 (Internals Vibration Monitoring System: IVMS)등이 있다. 현재, 국내의 여 러 원전에는 이들중 일부 계통들이 선택적으로 설치되어 운전중이며, 영광 3,4호기에서는 국내 최초로 이들 3개의 계통을 종합한 핵증기공급계통 건전 성감시계통(Nuclear Steam Supply System Integrity Monitoring System: NIMS)을 설계하였다. 특히, 영광 3,4호기 NIMS에서는 각 계통에 의해 감지 된 1차 계통 내의 이상상태를 하나의 분석컴퓨터(Analysis Computer)를 사 용하여 해석하는 종합결함 탐지해석 기법을 도입하였다.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.907-910
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• 2012

Since the inner barrel assembly of the Advanced Power Reactor 1400 reactor vessel is a new design feature introduced instead of CEA(control element assembly) shroud assembly, the inner barrel assembly can be a significant object of structural integrity assessment. This paper covers the structural responses of top plate, which is a component of the inner barrel assembly, against the deterministic hydraulic load induced by pump pulsation and the random hydraulic load induced by turbulence of coolant. The top plate responds to the deterministic hydraulic load more than to the random hydraulic load and shows enough structural integrity. The results of this paper will be important basis for the selection of instruments and measurement location.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.10a
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• pp.859-862
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• 2002

The Core Support Barrel (CSB) is a major component of Reactor Internals, and is designed to support and protect the Reactor Core. In this study, Reactor Core, Core Shroud and CSB were simplified to coaxial cylinders and then the offset of Reactor Core & Core Shroud to the dynamic characteristic of CSB was analyzed. For the beam modes, natural frequencies of the cantilevered cylinder are compared with those of the cantilevered beam. And it was found out that shear modulus must be used correctly to convert the shell model to the equivalent beam model. From the dynamic characteristics of the beam model, it was found out that natural frequencies are proportional to the length of Reactor Core & Core Shroud and inversely proportional to the mass. From the comparison with the dynamic characteristics of a beam model and a lumped-mass model it was found out that the size of lumped-mass must be determined considering both the length and the mass of Reactor Core & Core Shroud.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.10a
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• pp.306-311
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• 2012

A reactor vessel internals comprehensive vibration assessment program (RVI CVAP) of an advanced power reactor 1400 (APR1400) is being performed as a non-prototype category-2 type of reactor based on the US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure (UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly results show that meet the specified integrity levels of the design acceptance criteria. Also, the measuring locations are set by the analysis results of the UGS assembly and selection criteria of measuring locations prior to this study. These analysis results and measuring locations will be used as fundamental materials to design a measurement system for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.186-195
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• 1997

When asymmetric thermal-hydraulic conditions occur between cold legs, the core inlet temperature will be nonuniform if the coolant is not mixed perfectly in the lower plenum. These uneven core inlet conditions may induce the change in core power distribution. Thus realistic prediction of thermal mixing is important in such abnormal conditions. In this study, reactor internals, which are scaled down as to conserve the flow area ratio, are set up in the model of KORI Unit 1 with the scaling factor of 1/710 by volume and coolant temperatures are measured beneath the lower core plate. Based on experimental results, the ability of COMMIX-1B code to simulate the coolant mixing phenomena in the lower plenum is estimated. The results show that complete mixing never occurs in any conditions and the mixing pattern is characterized according to the plant type.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.576-582
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• 1996

The integrity of reactor vessel support system of the Korean Standard Nuclear Power Plant (KSNPP) is investigated for a postulated reactor vessel closure head drop event. The closure head is disassembled from the reactor vessel during refueling process or general inspection of reactor vessel and internal structures, and carried to proposed location by the head lift rig. A postulated closure head drop event could be anticipated during closure head handling process. The drop event may cause an impact load on the reactor vessel and supporting system. The integrity of the supporting system is directly relevant to that of reactor vessel and reactor internals including fuels. Results derived by elastic impact analysis, linear and non-linear buckling analysis and elasto-plastic stress analysis of the supporting system implied that the integrity of the reactor vessel supporting system is intact for a postulated reactor vessel closure head drop event.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.89-98
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• 2011

The construction technology for reactor vessel internals (RVI) modularization is one of the most important factors to be considered in reducing the construction period of nuclear power plants. For RVI modularization, gaps between the reactor vessel (RV) core-stabilizing lug and the core support barrel (CSB) snubber lug must be measured using a remote method from outside the RV. In order to measure RVI gaps remotely at nuclear power plant construction sites, certain core technologies must be developed and verified. These include a remote measurement system to measure the gaps between the RV core-stabilizing lug and the CSB snubber lug, an RVI mockup to perform the gap measurement tests, and a new procedure and schedule for RVI installation. A remote measurement system was developed previously, and a gap measurement test was completed successfully using the RVI mockup. We also developed a new procedure and schedule for RVI installation. This paper presents the new and improved installation procedure and schedule for RVI modularization. These are expected to become core technologies that will allow us to shorten the construction period by a minimum of two months compared to the existing installation procedure and schedule.