• Composites Research
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• 제31권4호
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• pp.156-160
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• 2018

본 논문에서는 콘형 복합재 격자구조체의 구조안전성 평가 기법에 대해 연구를 수행하였다. 콘형 복합재격자구조체의 구조안전성 평가는 유한요소해석을 통해 수행되었다. 구조안전성 평가를 위한 유한요소모델은 솔리드 요소를 사용하여 생성하였다. 섬유 교차부와 비교차부의 물성 차이를 고려하기 위해 섬유 체적률을 고려한 기계적 물성을 적용하였다. 구조해석 기법의 검증을 위해 콘형 복합재 격자구조체의 압축 시험을 수행하였다. 시험과 해석의 비교 결과, 약 2%의 변위 오차가 발생하여 잘 일치하는 것을 확인하였다.

• Composites Research
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• 제30권6호
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• pp.338-342
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• 2017

본 논문에서는 원통형 복합재 격자구조체의 구조안전성 평가 기법에 대해 연구를 수행하였다. 구조안전성 평가는 유한요소해석을 통해 수행하였다. 구조안전성 평가를 위한 최적의 유한요소를 확인하기 위해 원통형 복합재 격자구조체 유한요소모델은 빔, 쉘 그리고 솔리드 요소를 사용해 생성하였다. 쉘과 솔리드 모델의 유한요소 해석결과는 서로 유사하게 발생되었다. 그러나 빔 모델의 경우, 쉘과 솔리드 모델의 결과와 큰 차이가 발생하였다. 이것은 빔 요소가 원통형 복합재 격자구조체 섬유 비교차부의 기계적 물성저하를 고려하지 못하기 때문이다. 원통형 복합재 격자구조체의 구조안전성 평가를 위한 유한요소해석은 쉘 또는 솔리드 요소를 사용해야 하는 것을 확인하였다.

• 방사성폐기물학회지
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• 제19권4호
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• pp.491-501
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• 2021

This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 대한기계학회논문집A
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• 제27권6호
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• pp.930-937
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• 2003

The extension of design life among power plants is increasingly becoming a world-wide trend. Kori #1 unit in Korea is operating two cycle. It has two man-ways for tube inspection in a steam generator which is one of the important components in a nuclear power plant. Especially, stud bolts fur man-way cover have damaged by disassembly and assembly several times and degradation for bolt materials for long term operation. It should be evaluated and compared by ASME Code criteria for integrity evaluation. Integrity evaluation criteria which has been made by the manufacturer is not applied on the stud bolts of nuclear pressure vessels directly because it is controlled by the yield stress of ASME Code. It can apply evaluation criteria through FEM analysis to damaged female threads and to evaluated safety fer helical-coil method which is used according to Code Case-N-496-1. From analysis results, we found .that it is the same results between stress intensity which got from FEM analysis on damaged female threads over 10％ by manufacture integrity criteria and 2／3 yield strength criteria on ASME Code. It was also confirmed that the helical-coil repair method would be safe.

• Journal of Mechanical Science and Technology
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• 제14권2호
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• pp.215-225
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• 2000

The structural integrity of the fuel test loop (FTL) in a Korean experimental reactor is evaluated when the FTL, submerged in a water environment, is subjected to a postulated pipe rupture. The analyses are performed under static and dynamic conditions, imposing the thrust force history at each postulated pipe rupture section. Through analysis the following results are found: l) A double ended guillotine can not be expected based on the toughness of the material, 2) the structural integrity of the chimney surrounding the FTL would not impede the structural integrity by the pipe whip. All analyses are performed by finite element methods.

• 대한기계학회논문집A
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• 제24권1호
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• pp.206-214
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Journal of Advanced Marine Engineering and Technology
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• 제40권9호
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• pp.780-785
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• 2016

선박의 연료로 이용되는 중유는 지구 온난화를 일으키는 유해가스를 배출한다. 이를 저감하기 위해 친환경선박으로 표현되고 있는 녹색선박이 등장하게 되었고, 천연가스를 연료로 하는 액화천연가스(LNG) 연료 선박이 대표적인 사례이다. 본 논문에서는 LNG 선박에 사용되는 극저온 환경에서 고압 하중을 받는 기화기와 출입구 배관의 구조 건전성을 ASME 코드에 따라 평가하고 실용화한 최초의 사례이다. 기화기와 배관은 유한요소법을 사용하여 구조해석을 수행하였다. 기화기는 등가응력을 바탕으로 ASME Section VIII Division 2에 제시된 허용응력과 비교하여 건전성 평가를 수행하였고, 배관은 성분별 응력을 조합하여 ASME B31.3에 제시된 허용응력과 비교 및 건전성 평가를 수행하였다. 각 하중에 대한 구조물들의 응력 결과는 허용응력 범위 이내에 있으므로 구조적 건전성을 유지하는 것으로 평가되었다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.264-267
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• 2004

Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle piping systems in Nuclear Power Plants (NPP). Since the mid-1990s, secondary side piping systems in Korean NPPs have experienced wall thinning, leakages and ruptures caused by FAC. Korea Electric power Research Institute (KEPRI) and Korea Hydro & Nuclear Power Co., LTD. (KHNP) have conducted a study to develop the methodology for systematic pipe management and established the Korean Thinned Pipe Management Program (TPMP). To effectively maintain the integrity of piping system, FAC engineer should understand the criterions of the structural integrity evaluation and the safety margin assessment for the thinned pipe component. This paper describes the technical items of TPMP, and shows the example of the integrity evaluation and safety margin assessment for three thinned pipe component of a NPP.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.279-284
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• 2003

A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including regular in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage the integrity issues on a nuclear power plant. In this paper, a web-based fatigue life evaluation system for primary components in nuclear power plant is proposed. This system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant.