• 대한기계학회논문집A
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• 제40권2호
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• pp.237-244
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• 2016

산업에서 널리 사용되고 있는 압축 원통 코일 스프링은 제작 시 스프링 고유의 특성을 유지할 수 있는지 실험을 통해 검증해야 한다. 이 과정에서 많은 시간과 비용을 소비한다. 따라서 본 연구에서는 선형 정적 구조 해석 방법을 이용하여 스프링의 구조 건전성 평가를 수행하였다. 본 연구의 타당성을 입증할 목적으로 설계 규격 및 각종 국제적 평가 규격에 의한 실험적 방법과의 비교 및 검증을 수행하였다. 양 끝단 처리를 하지 않은 스프링 형상을 사용하여 구조 해석을 수행한 결과, 해석 모형 제작에 필요한 시간은 절감하고 격자의 질은 향상할 수 있었다. 푸아송 비는 구조 해석 결과에 별다른 영향을 미치지 않았다. 그리고 구조 해석만으로 스프링 구조 건전성을 검증할 수 있는 가능성을 확인하였다.

• 해양환경안전학회:학술대회논문집
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• 해양환경안전학회 2017년도 공동학술발표회
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• pp.111-111
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• 2017

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1320-1329
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• 2024

When wall-thinning occurs in nuclear Class 2 and 3 pipes, reinforcement is typically applied rather than replacement. To analyze the structural integrity of reinforced wall-thinned pipe, stress analysis results using full 3-D FE analysis are not compatible to the design code equation, ASME BPVC Sec. III NC/ND-3650. Therefore, the efficient stress evaluation method for the reinforced wall-thinned pipe, compatible to the design code equation, needs to be developed. In this paper, stress evaluation methods for the reinforced wall-thinned pipe are proposed using the equivalent straight pipe concept. Furthermore, for fatigue analysis of the reinforced wall-thinned pipe, the stress intensification factor of reinforced wall-thinned pipe is presented using the structural stress method given in ASME BPVC Sec. VIII Div.2.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2016년도 춘계 학술논문 발표대회
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• pp.91-92
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• 2016

Structural integrity of reinforced concrete structures including nuclear power plants needs to be evaluated on a regular basis. Deterioration inside the concrete structures can be represented by voids. In this study, the varied volume fraction of voids inside mortar specimens was studied as a parameter using ultrasonic testing equipments. Both direct and indirect measurement methods were employed. The results show that there is a clear distinction between the specimens with different void volume fractions.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.166-171
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• 2004

Since early 1950's fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet bas been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an internet-based cooperative system for integrity evaluation system which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and agent programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet.

• 한국유체기계학회 논문집
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• 제3권4호
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• pp.7-14
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• 2000

This paper provides integrity evaluation and root cause analysis for defects observed at volute tongue, or cutwater, of the operating centrifugal pump in power plant. The cause of the cracks are analyzed and reviewed from the viewpoint of the operation and maintenance of the pumps, and the sample obtained from the cracked volute tongue of the pump are examined. At first, in-situ hardness test and microstructure examination were performed to understand the cause of cracking at volute tongue. The evaluation of structural integrity and the possibility of the crack propagation is also evaluated. Cracks were typical intergranular cracking and propagated along with prior austenite grain boundary. At easing volute tongue, the hardness was higher than ASTM requirement and a large amount of intergranular Cr carbide was precipitated. These were due to high C content in material. P content was also higher than ASTM requirement. Therefore, Cr carbide precipitation and P segregation at grain boundary, caused by higher C and P content in material, resulted in intergranular cracking of casing volute tongue. This procedure for integrity evaluation and root cause analysis is used to guide, and support the pump designer and manufacturer's material selection and process design to avoid a costly, unplanned outage of plant.

• 대한기계학회논문집A
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• 제34권11호
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• pp.1773-1778
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• 2010

본 논문은 원자력 내진등급 원전용 주냉각수 펌프의 지진하중에 대한 구조 건전성을 코드에 따라 평가하였다. 펌프 구조물을 3차원 모델링하여 유한요소법을 사용하여 모달 해석 및 응답스펙트럼 해석을 수행하였다. 모달 해석 결과 고유진동수, 모드형상 및 모드참여계수를 얻을 수 있었다. 응력해석에서 구한 응력들을 조합하여 원자력 규격집에서 제시한 허용응력과 비교하여 펌프의 구조 건전성을 판단하였다. 모든 지진하중에 대한 펌프 구조물에서의 응력은 허용값 보다 작게 분포하므로 구조적 건전성을 유지한다고 평가할 수 있다.

• 한국산학기술학회논문지
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• 제9권6호
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• pp.1500-1505
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• 2008

본 논문은 원자력 내진등급 IIA 글로브 밸브의 지진하중에 대한 구조 건전성을 평가하였다. 밸브 구조물을 3차원 모델링하여 유한요소법을 사용하여 모달 해석 및 등가 정적 응력해석을 수행하였다. 모달 해석 결과 밸브 구조물은 충분히 강건하여 등가 정적해석이 가능하였고 해석에서 구한 응력들을 조합하여 원자력 규격집에서 제시한 허용응력과 비교하여 밸브의 구조 건전성을 판단하였다. 모든 지진하중에 대한 밸의 구조물에서의 응력은 허용값 보다 작게 분포하므로 구조적 건전성을 유지한다고 평가할 수 있다.

• 방사성폐기물학회지
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• 제21권4호
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• pp.559-570
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• 2023

The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

• 소음진동
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• 제10권2호
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• pp.261-268
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• 2000

The instrumented capsule is subjected to flow-induced vibration(FIV) due to the flow of the primary coolant and then the structural integrity of the capsule during irradiation in the HANARO reactor is an issue of major concern. For this purpose the acceleration was measured by four accelerometers attached to the protection tube of the capsule mainbody and the displacement of test holes was calcultated using commercial finite element program ANSYS to evaluate the structural interference with the neighboring flow tubes under the reactor operating condition. The calculated displacements of test holes in the reactor in-core were found to be lower than the values of allowable design criteria.