• 한국압력기기공학회 논문집
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• 제8권1호
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• pp.8-18
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• 2012

In early constructed nuclear power plants, Ni-based Alloys 82/182 had been widely used for dissimilar metal welds (DMW) as a weld filler metal. However, Alloys 82/182 have been proven to be susceptible to primary water stress corrosion cracking (PWSCC) in the nuclear primary water environment. The formation of crack due to PWSCC is also influenced by weld residual stresses. Thus, the accurate estimation of weld residual stresses of DMW is crucial to investigate the possibility of PWSCC and instability behaviors of crack due to PWSCC. In this context, the present paper investigates weld residual stresses of nuclear reactor piping nozzles based on 2-D axi-symmetric finite element analyses based on layer-based approach using maximum molten bead temperature. In particular, the effect of analysis parameters, i.e., a thickness of weld layer, an initial molten bead temperature, convection heat transfer coefficient, and geometric constraints on predicted weld residual stresses was investigated.

• 대한기계학회논문집A
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• 제37권1호
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• pp.47-54
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• 2013

본 논문에서는 Alloy 82/182를 용접재로 이용한 원자로 배관 이종금속 맞대기 용접부(Dissimilar Metal Butt Weld)에서의 PWSCC에 의한 균열성장 거동을 평가하였다. 이를 위해 먼저 유한요소 응력해석을 수행하여 이종금속용접부에서의 응력분포를 결정하였으며, 이때 이종금속용접 및 동종금속용접에 의한 용접잔류응력 외에 수압시험과 정상운전 조건도 고려하여 기계적 하중에 의한 응력 재분배를 고려하였다. 최종적으로 이와 같이 구한 응력 분포를 바탕으로 PWSCC에 의한 축방향 및 원주방향 가상 균열의 균열성장 거동을 평가하여 PWSCC 균열 성장량을 계산하였다. 본 논문의 결과는 향후 PWSCC에 의한 원자로 배관 이종금속 맞대기 용접부의 균열성장 거동 예측에 적용될 수 있다.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1805-1815
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• 2019

While nickel-based alloys have been widely used for power plants due to corrosion resistance and good mechanical properties, during the last couple of decades, failures of nuclear components increased gradually. One of main degradation mechanisms was primary water stress corrosion cracking at dissimilar metal welds of piping and reactor head penetrations. In this context, precise estimation of welding effects became an important issue for ensuring reliability of them. The present study deals with a series of finite element analyses and crack growth rate evaluation of Alloys 690/152. Firstly, variation of residual stresses and equivalent plastic strains was simulated taking into account welding of a cylindrical block. Subsequently, extraction and pre-cracking of compact tension (CT) specimens were considered from different locations of the block. Finally, crack growth curves of the alloys and heat affected zone were developed based on analyses results combined with experimental data in references. Characteristics of crack growth behaviors were also discussed in relation to mechanical and fracture parameters.

• 대한기계학회논문집A
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• 제36권2호
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• pp.137-148
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• 2012

가압경수로의 많은 관통관 중에서 니켈 기저 합금인 Inconel alloy 600 계열의 이종금속용접부는 일차수응력부식균열에 민감하며, 이를 평가하기 위하여 용접부에 작용하는 잔류응력분포를 정확히 예측하는 것이 중요하다. 본 논문에서는 유한요소해석을 이용하여 노즐 맞대기 이종금속용접부에 작용하는 일반적인 잔류응력분포를 예측하였다. 이를 위해 노즐 맞대기 이종금속용접부의 형상을 단순화하여 특정한 형상 변수에 따른 용접부 잔류응력분포를 확인하였으며, 이를 토대로 기존 문헌에 제시된 오스테나이트계 배관 맞대기 용접부 잔류응력 분포식을 수정하여 가압경수로 노즐 맞대기 이종금속용접부에 작용하는 일반적인 잔류응력분포 예측식을 제시하였다.

• 비파괴검사학회지
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• 제21권5호
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• pp.501-509
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• 2001

증기발생기 전열관에서 1차측 응력부식균열의 발생빈도가 증가하고 있으므로 이의 정확한 탐지와 평가를 위해서는 균열 형상에 따른 와전류 신호특성을 규명하고 적합한 탐촉자를 선정하는 것이 매우 중요하다. 본 연구에서는 증기발생기 전열관의 내면 축방향 균열에 대한 와전류 검사의 검출능과 크기예측에 대한 신뢰도를 정량적으로 평가하고 pancake coil과 plus coil과의 신호특성 차이를 비교하였다. 이를 위하여 전열관 내면에 EDM으로 노치를 가공한 시편과 실제 증기발생기에서 1차측 응력부식균열이 발생하여 인출한 전열관을 시험편으로 사용하였다. 본 연구에서 얻어진 결과를 토대로 내면 축방향 균열에 대한 와전류 검사 신뢰도 향상을 위한 방안을 제시하였다.

• 대한기계학회논문집A
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• 제39권3호
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• pp.269-273
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• 2015

일차수응력부식균열 개시 모델과 거시적 현상학적인 손상역학 접근론에 기반한 유한요소 손상해석을 수행하여 Alloy 600TT 로 제작된 원전 증기발생기 전열관에 발생하는 일렬 원주방향 표면 일차수응력부식균열의 성장에 미치는 균열 간격의 영향을 고찰하였다. 기존 연구 결과와의 비교를 통해 손상해석 방법의 타당성을 검증하였다. 검증된 방법을 일렬 원주방향 표면 일차수응력부식균열에 적용하였다. 적용한 결과, 단일 균열에 비하여 일렬 균열의 경우 보다 빠른 합체시간과 관통시간을 보이며 균열 간격이 증가할수록 합체시간과 관통시간은 증가함을 확인하였다. 또한 일정 간격이상으로 두 균열이 떨어지면 합체 이전에 관통될 수 있음을 확인하였다.

• 비파괴검사학회지
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• 제9권1호
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• pp.106-110
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• 1989

Bolting degradation problems in primary coolant pressure boundary applications have become a major concern in the nuclear industry. In the bolts concerned, the failure mechanism was either corrosion wastage(loss of bolt diameter) or stress-corrosion cracking.(3) Here the manual ultrasonic testing of RPV(Reactor Pressure Vessel) and RCP(Reactor Coolant Pump) stud has been performed. But it is difficult to detect indications because examiner can not exactly control the rotation angle and can not distinguish the indication from signals of bolt. In many cases, the critical sizes of damage depth are very small(1-2 mm order). At critical size, the crack tends to propagatecompletly through the bolt under stress, Resulting in total fracture.(3) Automatic stud scanner for studs(bolts) was developed because the precise measurement of bolt diameter is required in this circumstance. By use of this scanner, the rotation angle of probe was exactly controlled and the exposure time of radiations was reduced.

• Corrosion Science and Technology
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• 제14권3호
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• pp.113-119
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• 2015

Alloy 82/182 weld metals had been extensively used in joining the components of the PWR primary system. Unfortunately, there have been a number of incidents of cracking caused by PWSCC in Alloy 82/182 welds during the operation of PWR worldwide. To mitigate PWSCC, optimization of water-chemistry conditions, especially dissolved hydrogen (DH) and Zn contents, is considered as the most promising and effective remedial method. In this study, the PWSCC behaviours of Alloy 182 weld were investigated in simulated PWR environments with various DH content. Both in-situ and ex-situ oxide characterizations as well as PWSCC initiation tests were performed. The results showed that PWSCC crack initiation time was shortest in PWR water (DH: 30cc/kg). Also, high stress reduced crack initiation time. Oxide layer showed multi-layered structures consisted of the outer needle-like Ni-rich oxide layer, Fe-rich crystalline oxide, and inner Cr-rich inner oxide layers, which was not altered by the level of applied stress. To analyse the multi-layer structure of oxides, EIS measurement were fitted into an equivalent circuit model. Further analyses including TEM and EDS are underway to verify appropriateness of the equivalent circuit model.

• 대한기계학회논문집A
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• 제32권9호
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• pp.770-781
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• 2008

In nuclear power plants, ferritic low alloy steel components were connected with austenitic stainless steel piping system through alloy 82/182 butt weld. There have been incidents recently where cracking has been observed in the dissimilar metal weld. Alloy 82/182 is susceptible to primary water stress corrosion cracking. Weld-induced residual stress is main factor for crack growth. Therefore exact estimation of residual stress is important for reliable operating. This paper presents residual stress computation performed by 6" safety & relief nozzle. Based on 2 dimensional and 3 dimensional finite element analyses, effect of welding variables on residual stress variation is estimated for sensitivity analysis.

• Journal of Welding and Joining
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• 제27권4호
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• pp.32-37
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• 2009

Alloy 600 and Alloy 82/182 materials have been used widely in PWR plants. But these materials are known to be susceptible to PWSCC(Primary Water Stress Corrosion Cracking). Recently, there have been several PWSCC events in major components due to repair welding, because repair welding in the dissimilar metal welds during the construction increases residual stress significantly on the inner surface of welds. In this paper, various residual stress analyses for repair welding were performed using FEM to check the effect of repair welding on residual stress distributions in PZR safety/relief nozzle. The results indicate that for inside surface repair welding, high tensile residual stress is developed on the inside surface of the nozzles.