• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.51-55
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• 2014

Based on the results of Probabilistic Safety Assessment(PSA) for a Nuclear Power Plant (NPP), Common Cause Failure(CCF) events have been recognized as one of the main contributors to the risk. Also, the CCF data and estimation method used in domestic PSA models have been pointed out as an issue with respect to the quality. The existing method of MGL and non-staggered testing even widely used were considered conservative in estimating the safety and had a limited capability in uncertainty analyses. Therefore, this paper presents the CCF estimation using a new generic data source and Alpha factor method. The analyses showed that Alpha factor and staggered method are effective in estimating the CCF contribution and risk insights of reference plant. This method will be a common bases for the optimization of new design for the construction plants as well as for the updating of safety assessment on the operating nuclear power plants.

• 한국압력기기공학회 논문집
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• 제5권1호
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• pp.32-37
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• 2009

The emergency diesel generator (EDG) in a nuclear power plant (NPP) shall start within 10 secondss and supply electrical power to engineered safety features within one minute and less if a loss of offsite power (LOOP), A design-basis event, or their combination occur. Each NPP has an EDG set consisting of two diesel generators for redundancy. In addition to the EDG set, an alternate Alternating Current Diesel Generator (AAC DG) is installed and shared by several units to cope with a station black out (SBO), i.e., loss of the offsite power concurrent with reactor trip and unavailability of the EDG set. The objective of this study is to analyze the failure data of emergency diesel generators reported in overseas nuclear power plants.

• 한국마린엔지니어링학회:학술대회논문집
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• 한국마린엔지니어링학회 2005년도 후기학술대회논문집
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• pp.224-225
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• 2005

The lightness of components required on marine and shipbuilding industry is requiring high strength of components. In particular, fatigue failure phenomena, which happen in metal, bring on danger in human life and property. Therefore, antifatigue failure technology takes an important part of current industries. In this study, it was investigated about endurance and fatigue crack propagation rate of according to stress ratio of SMAW commonly using for welding structures in present. Fatigue crack propagation rate(da/dN) of low load(R=0.1) was lower than of high load(R=0.6) for piping weld. And in stage I, ${\Delta}$Kth, the threshold stress intensity factor of the weld under heavy load is higher than under small load. Fatigue life shows more improvement in the weld of stress ratio R=0.l than in the weld of stress ratio R=0.6.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.90-95
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• 2003

Fracture behaviors of pipes with local wall thinning are very important for the integrity of nuclear power plant. In pipes of energy plants, sometimes, the local wall thinning may result from severe erosion-corrosion (E/C) damage. However, the effects of local wall thinning on strength and fracture behaviors of piping system were not well studied. In this paper, the monotonic bending tests were performed of full-scale welded and unwelded carbon steel pipes with local wall thinning. A monotonic bending load was applied to straight pipe specimens by four-point loading at ambient temperature without internal pressure. The observed failure modes were divided into four types; ovalization, crack initiation/growth after ovalization, local buckling and crack initiation/growth after local buckling. Also, the strengths of welded and unwelded piping system with local wall thinning were evaluated.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.33-41
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• 2012

The component cooling water heat exchangers are critical components in a nuclear power plant. As the operation years of the heat exchanger go by, the maintenance costs required for continuous operation also increase. Most heat exchangers have carbon steel shells, tube support plates and flow baffles. The titanium tube is susceptible to flow induced vibration. The damage on carbon steel tube support rod and titanium tube around cooling water entrance area is inevitable. Therefore, analysis of internal flow around the component cooling water entrance and tube channel is a good opportunity to seek for failure prevention practice and maintenance method. The numerical study was carried out by FLUENT code to find out the causes of tube failure and its location.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.246.1-246
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.210.2-210
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• 2002

• 대한토목학회논문집
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• 제31권1C호
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• pp.1-10
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• 2011

본 논문에서는 제방 제체 하부지반에 투수계수 차이가 많이 나는 투수층(전석층)이 존재하는 김포제방에서 발생한 파이핑 현상을 모형실험 중 하나인 원심모형시험을 이용하여 모사하였다. 또한 원심모형실험과 동일 조건을 대상으로 수치해석을 수행하여 원심모형실험 결과와 비교하였다. 우선, 김포 제방을 바탕으로 원심모형실험을 수행할 단면을 결정하고, 간극수압계를 매설한 모형 제방에서 원심모형실험을 수행하였다. 실험을 통해 대부분의 물이 전석층으로 흐르는 것을 확인 하였으며 그로 인해 파이핑 현상이 발생하였다. 비디오 카메라 영상을 통해 기존의 연구 결과와 같이 전석층과 제내지 측 지표면과 거리가 가장 짧은 부분에서의 파이핑 현상을 확인하였다. 간극수압계를 통해 획득된 수압 분포 특성도 전석층으로의 물의 흐름으로 인해 파이핑 현상이 발생한 것임을 확인시켜주었다. 수치해석 결과 또한 원심모형실험 결과와 동일한 위치에서 파이핑 현상이 발생하는 것을 확인하였다. 원심모형시험을 통해 제방에서의 파이핑 현상을 모사하였으며 추후 제방 연구에서 원심모형시험을 이용하여 다양한 실험 및 현상을 평가, 예측 할 수 있을 것이라 판단된다.

• 한국지반공학회논문집
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• 제26권4호
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• pp.59-68
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• 2010

파이핑은 제체 내부 침식의 한 형태로서 토립자의 진행성 유출에 기인하며 균열이 발생한 댐 뿐만 아니라 용해성 재료로 인하여 수년 동안 만족스런 거동을 보인 댐들에서도 발생한다. 파이핑 현상은 필댐의 붕괴 원인의 약 50%를 차지하고 있으며 필댐 관리의 주요항목이나, 이에 대한 체계적인 평가에는 다소 어려움이 있다. 이에 본 연구에서는 파이핑 현상에 대한 비파괴 탐지를 위한 전기비저항 탐사기법의 적용성을 파악하기 위해 대청댐 비상여수로 건설을 위해 해체되는 부댐에 파이핑 모사시험과 전기비저항 모니터링을 수행하였다. 제체에 수평공을 천공하여 인공적인 유로를 만들고 저수지 물을 유입시켜 실규모 파이핑 현상을 모사하였으며, 파이핑 현상 모사 시험전과 시험 중 측정한 전기비저항값을 비교한 결과 파이핑 발생 시 측정한 전기비저항값이 조금 더 큰 변화를 보임으로써 전기비저항 모니터링 탐사는 실제 담수되어 있는 제체의 파이핑 현상을 효과적으로 탐지할 수 있을 것으로 판단된다.

• 한국안전학회지
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• 제18권4호
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• pp.28-34
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• 2003

In the present study, three-dimensional finite element analysis was performed to investigate the effects of internal wall thinning defect on the failure pressure of elbow in the piping system and to develop the failure pressure evaluation model. From the results of finite element analysis, the failure pressure was derived by employing local stress criteria, and the effects of thinning location, bend radius, and defect geometry on the failure pressure of internally wall thinned elbow were investigated. Also, based on these investigations and previous model developed to estimate the failure pressure of elbow with an external pitting defect, the failure pressure evaluation model to be applicable to the elbow containing an internal thinning defect was proposed and compared with the results of finite element analysis. The failure pressure calculated by the model agreed well with the results of finite element analysis.