• 대한기계학회논문집A
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• 제23권11호
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• pp.2096-2101
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• 1999

This paper presents the effect of shape of external corrosion in pipeline on failure prediction by using numerical simulation. The numerical study for the pipeline failure analysis is based on the FEM(Finite Element Method) with an elastic-plastic and large-deformation analysis. The predicted failure stress assessed for the simulated corrosion defects having different corroded shapes along the pipeline axis are compared with those by methods specified in ANSl/ASME B31G code and a modified B31G code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.795-801
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• 1998

Metallic fatigue of Pressurized Water Reactor(PWR) materials is a generic safety issue for commercial nuclear power plants. It is very important to obtain the fatigue usage factor for component integrity and life extension. In this paper, fatigue usage was obtained at the inside surface of Kori unit 2, 3 and 4 RCP casing weld, based on the design transient. And it was intended to establish the procedure and the detailed method of fatigue evaluation in accordance with ASME Section III Code. According to this code rule, two methods to determine the stress cycle and the number of cycles could be applied. One method is the superposition of cycles of various design transients and the other is based on the assumption that a stress cycle correspond to only one design transient. Both method showed almost same fatigue usage in the RCP casing weld.

• International Journal of Precision Engineering and Manufacturing
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• 제1권2호
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• pp.48-54
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• 2000

This paper presents the effect of shape of external corrosion in pipeline on failure prediction by using a numerical simulation. The numerical study for the pipeline failure analysis is based on the FEM(Finite Element Method)with an elastic-plstic and large-deformation analysis. Corrosion pits and narrow corrosion grooves in pressurized pipeline were analysed. A failure criterion, based on the local stress state at the corrosion and a plastic collapse failure mechanism, is proposed. The predicted failure stress assessed for the simulated corrosion defects having different corroded shapes along the pipeline axis compared with those by methods specified in ANSI/ASME B31G code and a modified B31G code. It is concluded the corrosion geometry significantly affects the failure behavior of corroded pipeline and categorisation of pipeline corrosion should be considered in the development of new guidance for integrity assessment.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2005년도 연구개발 발표회 논문집
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• pp.748-756
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• 2005

The Purpose of this writing is a presentation of small turbine efficiency measuring method, applicable new technology, and several analysis result for real turbine. Measurement methods of hydraulic efficiency written in here are extracted from small turbine applicable international code(IEC, ASME). It includes brief synopsis of 'Current meter method' and 4 other methods and ASFM as a new small turbine applicable technology. And several analysis of test result are for recently performed domestic small turbine result in korea. Through this presentation of extracted code, I hope that it make other small turbine concerner be familiar to perform an efficiency test. for small turbine. And, some analysis of that, make other to feel the importance of efficiency test.

• 대한기계학회논문집A
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• 제39권1호
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• pp.27-35
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• 2015

본 연구에서는 피로 하중 및 크리프 하중을 받는 Mod.9Cr-1Mo (ASME Grade 91)강 시편에 대한 일련의 실험결과로부터 재료물성치인 고온 균열진전 모델을 개발하였다. 이 균열진전 모델은 크리프-피로하중을 받는 균열체의 결함평가에 사용되는 물성치이다. 한국원자력연구원이 수행한 일련의 피로 균열진전(FCG) 속도 실험 및 크리프 균열진전(CCG) 속도 실험 결과로부터 균열진전 모델을 결정하고, 이를 프랑스의 고온 설계 기술기준인 RCC-MRx 와 비교함으로써 설계 물성치의 보수성에 대해 검토하였다. RCC-MRx 는 FCG 모델 및 CCG 모델을 Section III Tome 6 에서 제공하고 있는데, 실험으로부터 결정한 균열진전 모델과 비교한 결과 RCC-MRx 의 FCG 모델은 보수적인 것으로 나타난 반면 CCG 모델은 비보수적인 것으로 나타나 동 물성치에 대한 검증이 필요한 것으로 나타났다. 또한 본 연구에서는 기계적 강도 및 크리프 시험결과에 대해서도 RCC-MRx 의 물성치와 비교 및 분석하였다.

• Journal of Welding and Joining
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• 제31권6호
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• pp.1-7
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• 2013

In nuclear power plants, lined carbon steel pipes or PCCPs (pre-stressed concrete cylinder pipes) have been widely used for sea water transport systems. However, de-bonding of linings and oxidation of PCCP could make problems in aged NPPs (nuclear power plants). Recently at several NPPs in the United States, the PCCPs or lined carbon steel pipes of the sea water or raw water system have been replaced with HDPE (high density polyethylene) pipes, which have outstanding resistance to oxidation and seismic loading. ASME B&PV Code committee developed Code Case N-755, which describes rules for the construction of buried Safety Class 3 polyethylene pressure piping systems. Although US NRC permitted HDPE materials for Class 3 buried piping, their permission was limited to only 10-year operation because of several concerns including the quality of fusion zone of HDPE. In this study, various requirements for fusion qualification test of HDPE and some regulatory issues raised during HDPE application review in foreign NPPs are introduced.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2918-2927
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• 2020

This paper proposes a simplified elastic-plastic analysis procedure using the penalty factors presented in the Code Case N-779 for strain-based fatigue assessment of nuclear safety class 1 components under severe seismic loads such as safety shutdown earthquake and beyond design-basis earthquake. First, a simplified elastic-plastic analysis procedure for strain-based fatigue assessment of nuclear safety class 1 components under the severe seismic loads was proposed based on the analysis result for the simplified elastic-plastic analysis procedure in the Code Case N-779 and the stress categories corresponding to normal operation and seismic loads. Second, total strain amplitude was calculated directly by performing finite element cyclic elastic-plastic seismic analysis for a hot leg nozzle in pressurizer surge line subject to combined loading including deadweight, pressure, seismic inertia load, and seismic anchor motion, as well as was derived indirectly by applying the proposed analysis procedure to the finite element elastic stress analysis result for each load. Third, strain-based fatigue assessment was implemented by applying the strain-based fatigue acceptance criteria in the ASME B&PV Code, Sec. III, Subsec. NB, Article NB-3200 and by using the total strain amplitude values calculated. Last, the total strain amplitude and the fatigue assessment result corresponding to the simplified elastic-plastic analysis were compared with those using the finite element elastic-plastic seismic analysis results. As a result of the comparison, it was identified that the proposed analysis procedure can derive reasonable and conservative results.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3229-3240
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• 2023

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K-P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.

• 에너지공학
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• 제26권2호
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• pp.1-11
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• 2017

최근 합성천연가스(synthetic natural gas, SNG)의 사용과 합성천연가스를 생산하는 플랜트의 실증 운영이 증가하고 있다. SNG 플랜트는 다양하게 개발된 여러 합성 공정 기술이 적용되고 있으며, 이러한 공정의 특성상 고온, 고압의 운전 조건을 가진다. 기존 여러 연구들은 주로 합성천연가스 생산을 위한 화학적 합성 공정의 변수와 공정 최적화에 대한 연구에 집중되어 왔다. 이에 비해, 기존 산업 플랜트와는 다소 차별되는, 공정 특성으로 인한 SNG 플랜트의 기계적 손상과 유지보수 기법에 대한 연구는 많지 않다. 본 연구에서는 SNG플랜트의 주요 배관계통에 대해 ASME B31.3에 의거한 배관 시스템 응력 해석을 수행하였다. 또한 특이 부위에 대해 상세 국부 응력 해석을 수행하였다. 해석 결과로부터 배관 주요부위 중 파손 리스크가 높은 취약부의 위치를 선정하였다. 이 위치들은 배관 위험도 관리 대상으로 활용할 수 있다. 배관 시스템 응력 해석은 설계 운전조건과 실제 운전조건을 고려하여 수행되었다. 배관 시스템 응력 해석을 통해 도출된 주요 부위에 대해서는 국부적 상세 응력 해석을 위해 유한 요소 해석이 수행되었다. 발생되는 상세 응력 값은 가스화 반응기 및 하부 곡관부 대한 ASME B31.3 코드 표준을 만족하였다. 하부 곡관부의 경우 수직 변위를 제한하는 것이 구조적으로 안전 향상에 좋을 것으로 파악되었다. 수행된 해석결과는 향후 위험도 기반 유지 보수 검사 및 안전 운영에 대해 기반 정보로 사용될 수 있을 것으로 판단된다.

• 한국전산구조공학회논문집
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• 제30권3호
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• pp.207-214
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• 2017

국내의 72.5kV 이상, 주파수 60Hz의 송배전설비인 옥내 및 옥외용 가스절연개폐장치(GIS)는 내진 안전성에 대해 국가에서 정한 한전표준규격(ES-6110-0002)을 만족해야 한다. 이 규격에서 명시되지 않은 사항은 IEC 62271-203, 62271-207 등의 관련 기기 규격에 준한다. 한전표준규격에서 기기는 정상사용상태와 특수사용상태에서 건전성이 유지되어야 한다. 안전성 판단을 위해 ASME BPVC SEC.VIII 내압용기 설계 기준에 의해 A6061-T6 재질의 GIS에 대한 정상사용상태 기준과 국내 한전표준규격과 국외 IEC 62271-207에 의한 특수사용상태 기준(지진)에 대한 총체적 응력상태를 판단하였다. 한전표준규격 기준(0.22g) 적용시, 최종응력이 알루미늄인 Part A는 78.2MPa, Part D2의 경우 102.3MPa로, ASME 허용응력 값 181.5MPa를 만족하고 있다. IEC 62271-207 High 0.5g의 경우에도 최종응력은 Part A는 90.5MPa, Part D2는 103.8MPa이다. 본 연구 결과, 72.5kV GIS는 한전표준규격의 구조안전성과 내진성능을 충분히 만족함을 보이고 있다. 내진해석으로 내진시험을 수행할 수 없는 대형 전력기기의 내진성능 실증에 활용될 수 있을 것으로 기대된다.