• Journal of the Korean Society for Nondestructive Testing
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• v.16 no.1
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• pp.46-51
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• 1996

원자로는 압력용기 및 내부 구조물로 구분되어 있다. 내부 구조물들의 경련 열화 현상에 따라 결함이 많이 발생하여 이에 대한 초음파검사가 요구되고 있다. 따라서 원자로 내부 구조물에 대한 초음파검사 현황 및 각각 구조물들의 검사 원리를 기술하였다. 특히 원자로 내부 구조물 중 CRDM, core baffle bolt, core barrel bolt, CRGT-support pin 및 fuel alignment pin에 대한 유럽 및 독일을 중심으로 한 검사 현황 및 검사방법을 간략하게 기술하였다. 이 기술에 대한 지침안(guideline)이 독일, 프랑스, 일본을 중심으로 하여 마련되고 있다.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.433-439
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• 1994

영광 3, 4호기는 ABB-CE 사의 System 80 원자로와 비교해서 원자로 내부 구조물의 수력학적 설계 목적과 방법 이 동일하고, 단지 원자로의 크기와 출력이 상대적으로 작아진 내부 구조물이 축소된 형태이다. 따라서 System 80 유동 모델 시험에서 측정된 실험 결과로부터 영광 3, 4호기 연료 집합체 수에 맞게 보간법을 사용하여 보수적으로 유량 분포를 구하고 영광 3, 4호기 유동 모델 시험에서 얻어진 유량 분포와 비교하여 원자로의 수력학적 특성을 검토하고 자각에 대해 열적 여유도를 구하여 이런 경우에 원자로 유동 모델 시험을 수행하지 않고 이전의 실험 결과를 설계에 사용할 수 있는 가에 대해 연구하였다.

• Journal of KSNVE
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• v.7 no.1
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• pp.33-41
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• 1997

To cover a range of possible site conditions where the Korean standard nuclear power plant may be constructed, a range of generic site conditions is selected for geologic and seismologic evaluation. To envelop other Asian countries as well as the Korean peninsula, there is an attempt to increase the seismic level to 0.3g ground motions for the safe shutdown earthquake. The dynamic analyses of the reactor vessel internals and fuel assemblies are performed for the increased motions and the effect of seismic level on the response is investigated. Also the nonlinear response characteristics are discussed by comparing the loads between operating basis earthquake and safe shutdown earthquake excitations. The design adequacy of the reactor vessel internals and fuel assemblies for the increased seismic level is addressed.

• Computational Structural Engineering
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• v.8 no.1
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• pp.4-12
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• 1995

원자로 격납건물(Reactor Containment Bldg)은 정상가동시는 물론 냉각재상실사고(LOCA)를 포함하는 설계기준사고(DBA) 및 설계기준지진(DBE) 발생시 구조물 자체의 건전성 확보는 물론 주기기(NSSS Equipment)를 포함하는 안전관련 계통 및 기기를 안전하게 보호/지지하므로써 핵누출을 방지하여 발전소 종사자를 포함하는 국민의 재산과 생명을 보호하는 역할을 하는 원자력발전소에서 가장 중요한 구조물이다. 원자로 격납건물은 압력용기(Pressure Vessel : 설계내압 5 psi 이상인 용기)로 설계되는 격납용기와 1, 2차 차폐구조 등의 내부구조물로 구성되는데 이 중 본 소고에서는 격납용기의 해석 및 설계 그리고 구조건전성 시험 및 사용중검사에 대해서만 간략하게 기술한다.

• Nuclear industry
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• v.15 no.3 s.145
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• pp.82-92
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• 1995

• Computational Structural Engineering
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• v.6 no.4
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• pp.73-82
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• 1993

Investigated in this paper is the effect of fluid-structure interaction between reactor internal components due to their immersion in a confining fluid on the dynamic responses. A non-linear mathematical model is developed for the dynamic analysis of the reactor internals, which includes lumped masses, stiffnesses and hydrodynamic couplings. The hydrodynamic mass matrix which characterizes the fluid-structure interaction is calculated. Also, the equations of motion containing hydrodynamic mass matrix are presented. The responses of the reactor internals due to seismic and pipe break excitations are obtained for the case of with- and without-hydrodynamic couplings and the different response characteristics are investigated.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.38 no.3
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• pp.271-277
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• 2014

Internal structures, especially those located in the upstream of a reactor core, may have a significant influence on the core inlet flow rate distribution depending on both their shapes and the relative distance between the internal structures and the core inlet. In this study, to examine the effect of the reactor internal structure geometry treatment method on the prediction accuracy for the scale-down APR+ flow distribution, simulations with real geometry modeling were conducted using ANSYS CFX R.14, a commercial computational fluid dynamics software, and the predicted results were compared with those of the porous medium assumption. It was concluded that the core inlet flow distribution could be predicted more accurately by considering the real geometry of the internal structures located in the upstream of the core inlet. Therefore, if sufficient computational resources are available, an exact representation of these internal structures, for example, lower support structure bottom plate and ICI nozzle support plate, is needed for the accurate simulation of the reactor internal flow.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.743-752
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• 1995

This paper investigates the effects of fuel groupings in the coupled internals and core model on the internals and fuel responses due to pipe breaks. The 177 fuel assemblies for Korean standard nuclear power plant are grouped into several stick models and the responses of internals components are calculated. The analysis results show that the fuel model groupings in the coupled internals and core model have no significant effects on the internals and fuel responses for pipe break excitation. Also, in order to determine the feasibility of constructing a single equivalent stick representation of In or more adjacent fuel bundles, the reduced models, each of which employs a different stiffness lumping rule, are constructed. It is shown that the equivalent stiffness calculated to get the first natural frequency of the original model while preserving net gap between grouping centers gives the minimum modelling error.

• Nuclear industry
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• v.21 no.5 s.219
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• pp.84-86
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• 2001

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.9
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• pp.855-862
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• 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.