• 한국전산구조공학회논문집
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• 제26권3호
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• pp.207-212
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• 2013

APR1000(Advanced Power Reactor 1000)은 기존의 OPR1000(Optimized Power Reactor 1000)에 60년 설계수명, 국부주파수제어운전, 0.3g 안전정지지진하중 적용 등의 향상된 설계특성(Advanced Design Feature)을 적용하여 개선한 수출형 1000MW 원전이다. 이 논문에서는 Reg. Guide 1.207에서 요구하는 원자로냉각재 환경을 고려한 피로 평가를 원자로용기에 대하여 평가하였다. 원자로용기에서 비교적 누적사용계수가 높은 출구노즐을 대상으로 평가를 수행하였으며 출구노즐은 구조적 건전성을 만족하는 것으로 평가되었다.

• 대한전자공학회논문지SD
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• 제41권4호
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• pp.15-22
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• 2004

이 논문은 PCB의 PDN(Power Distribution Network) 시스템을 고려한 채널 분석을 나타내었다. 설계자가 원하는 PDN 시스템을 설계하기 위하여, 전체 주파수 범위의 PDN이 요구하는 임피던스를 얻는 유용한 설계방법을 제안하였다. 제안된 방법은 주파수 영역과 관계된 계층적 배치 접관방식과 보트와 decoupling 커패시터 사이의 current 흐름의 간섭을 고려한 path-based equivalent 회로를 기본으로 하였다. 비록 빠르고 쉬운 계산을 위한 lumped model일지라도, 실험 결과는 제안된 모델이 numerical 분석처럼 거의 정확함을 보였다. PDN 시스텐의 분석은 패키지 인덕턴스가 파워 노이즈, 데이터 채널을 통한 신호 이동에 영향을 받는다는 것을 보여주고 있으나, 보드 PDN 또한 정확한 채널 신호를 위해 무시할 수 없다는 것을 보여준다. 따라서 설계자는 반드시 초고속 디지털 시스템의 첫 스팩 설계로부터 보드, 패키지, 칩 등을 동시에 디자인을 해야 한다.

• Journal of Mechanical Science and Technology
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• 제17권9호
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• pp.1380-1387
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• 2003

When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

• Nuclear Engineering and Technology
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• 제37권2호
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• pp.201-206
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• 2005

Fretting, which is a special type of wear, is defined as small amplitude relative motion along the contacting interface between two materials. The structural integrity of steam generators in nuclear power plants is very much dependent upon the fretting wear characteristics of Inconel 690 U-tubes. In this study, a finite element model that can simulate fretting wear on the secondary side of the steam generator was developed and used for a quantitative investigation of the fretting wear phenomenon. Finite element modeling of elastic contact wear problems was performed to demonstrate the feasibility of applying the finite element method to fretting wear problems. The elastic beam problem, with existing solutions, is treated as a numerical example. By introducing a control parameter s, which scaled up the wear constant and scaled down the cycle numbers, the algorithm was shown to greatly reduce the time required for the analysis. The work rate model was adopted in the wear model. In the three-dimensional finite element analysis, a quarterly symmetric model was used to simulate cross tubes contacting at right angles. The wear constant of Inconel 690 in the work rate model was taken as $K=26.7{\times}10^{-15}\;Pa^{-1}$ from experimental data obtained using a fretting wear test rig with a piezoelectric actuator. The analyses revealed donut-shaped wear along the contacting boundary, which is a typical feature of fretting wear.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.111-120
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• 2001

In this study, a dynamic analysis of the integral reactor SMART (System-integrated Modular Advanced ReacTor) under postulated seismic events is performed to review the response characteristics of the major components. To enhance the feasibility of an analysis model, a detailed finite element model is synchronized with the products of concurrent design activities. The artificial time history, which has been applied to the seismic analysis for the Korean Standard Nuclear Power Plant (KSNP), is chosen to envelop broad site specifics in Korea. Responses in the horizontal direction are found slightly amplified, while those in the vertical direction are suppressed. Since amplified response is monitored at the control element drive mechanism (CEDM), minor design provision is considered to enhance the integrity of the subsystem.

• 한국안전학회지
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• 제18권3호
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• pp.39-45
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• 2003

The failure mode of a pipe due to local wall thinning is increasingly more attention in the nuclear power plant industry. To assess the integrity of locally wall thinned pipe, it is necessary to perform many simulations under various conditions. Because the modeling for locally wall thinned elbow is more complicated than that of straight pipe the efficient modeling method for finite element analysis is necessary. In this study, the more simple efficient modeling method of three-dimensional finite element analysis for locally wall thinned elbow has been suggested and verified. And using the method, the failure mode of local wall thinned elbows that have different thinning lengths and circumferential angles is evaluated. From the results, we concluded that the collapse load of elbows has been decreased by the increase of wall thinning shape factors such as thinning lengths and circumferential angles.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1117-1131
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• 2019

Tanks are used extensively in many engineering areas for spent fuel pool structures at nuclear power plants or for water storage tanks in bulk carriers. To ensure the structural integrity of such tanks when under dynamic loads, modal characteristics such as natural frequencies, participation factors and mode shapes should be known. Investigated in this study are the modal characteristics of a square tank by the finite element method. This approach can be used with subsequent dynamic analyses such as a response spectrum analysis or a harmonic analysis. Finite element models are prepared to determine the natural frequencies and mode shapes, which are easy to find the modal characteristics of a fluid-filled square tank. The effects of the fluid contained in the tank and the boundary conditions at top and bottom ends on the modal characteristics are assessed by several finite element analyses.

• 대한기계학회논문집B
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• 제34권9호
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• pp.867-873
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• 2010

수냉각 발전기 고정자 권선에서의 냉각수 누수 및 흡습에 의한 절연파괴 손상사례가 국내 및 국외에서 자주 발생되고 있다. 이러한 사고는 막대한 경제적 피해뿐만 아니라 전력의 안정적 공급 측면에서 매우 심각한 계통 사고로 연결될 수 있다. 특히 국내 발전기는 15년 이상 운전되어 열화가 진행된 발전기가 50% 이상이며, 계획예방정비 기간 중에 권선에서의 누설 및 흡습 권선이 종종 발견되고 있다. 기존에는 누수 시험 전 과정인 권선 건조 과정을 무시한 채 누수 시험 결과만으로 권선 누설 여부를 진단하였으나 본 논문에서는 누수 시험을 위한 준비 단계인 진공 건조 시의 권선 내부의 진공도 패턴 분석을 통해 권선 누설 및 흡습 여부를 예측진단할 수 있는 방법을 실험적으로 증명하였다.

• 전산구조공학
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• 제6권2호
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• pp.79-86
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• 1993

원자력발전소중 안전과 관련된 구조물은 지진의 가능성에 대비하여 그의 구조적 안전성과 가용성이 검증되어야 한다. 본 논문은 원자력발전소 보호시스템 캐비넷을 예를 들어 그에 대한 내진검증 방법을 보였다 캐비넷의 유한요소모델을 작성하여 동특성을 구하였고 그 모우드값을 입력지진스펙트럼과 비교한 결과 구조물의 1차 모우드가 입력 스펙트럼의 peak와 일치함으로써 설계변경의 필요성이 대두되었다. 이 peak값을 피하기 위하여 캐비넷의 구조를 변경하였고 변경된 구조물에 대하여 응답스펙트럼해석과 시간이력해석을 수행하여 구조적 건전성과 가용성을 보임으로써 설계변경된 캐비넷의 내진검증을 확인하였다.

• 한국유체기계학회 논문집
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• 제2권2호
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• pp.13-18
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• 1999

The pump safety related to the functions in nuclear power plants must be designed to meet load conditions considering seismic requirements. In order to satisfy both structural integrity and operability of these pumps, the initial step in the seismic qualification is to establish the resonant frequencies of the structure. Applications we made to the design of the vertical and horizontal type pump. Computational results are analyzed with respect to the dynamic characteristics and are compared to the expected design requirements.