• Nuclear Engineering and Technology
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• 제49권4호
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• pp.817-824
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• 2017

Residual stress is a critical element in determining the integrity of parts and the lifetime of welded structures. It is necessary to estimate the residual stress of a welding zone because residual stress is a major reason for the generation of primary water stress corrosion cracking in nuclear power plants. That is, it is necessary to estimate the distribution of the residual stress in welding of dissimilar metals under manifold welding conditions. In this study, a cascaded support vector regression (CSVR) model was presented to estimate the residual stress of a welding zone. The CSVR model was serially and consecutively structured in terms of SVR modules. Using numerical data obtained from finite element analysis by a subtractive clustering method, learning data that explained the characteristic behavior of the residual stress of a welding zone were selected to optimize the proposed model. The results suggest that the CSVR model yielded a better estimation performance when compared with a classic SVR model.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1060-1068
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• 2019

This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.

• 시스템엔지니어링학술지
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• 제12권1호
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• pp.89-103
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• 2016

Due to safety of the plant, it became very clear the importance of study occurrence reactor coolant system (RCS) issues specially the primary water stress corrosion cracking (PWSCC). The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. Robotic devices have been used for internal inspection, maintenance and performing remote welding and inspection in high-radiation areas. In this paper, PWSCC overview and inlay and over lay welding methodology introduced, concept of robotic device that can be inserted into the piping via Steam Generator (SG) main way to access to primary piping of pressurized water reactor (PWR) is developed based on SE methodology. A 3D model of the inspection system was developed along with the APR1400 (Advanced Power Reactor)reactor coolant systems (RCS) and internals with virtual 3D simulation of the operation for visualization to prove the validity of the concept.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.534-537
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• 2008

In recent years, the dissimilar metal, Alloy 82/182 welds used to connect stainless steel piping and low alloy steel or carbon steel components in nuclear reactor piping system have experienced cracking due to primary water stress corrosion(PWSCC). It is well known that one reason of the cracking is the residual stress by the weld. But, it is difficult to estimate exactly weld residual stress due to many parameters of welding. In this paper, the analysis of 3 FEM models made by ABAQUS Code is performed to estimate exactly the weld residual stress on the dissimilar metal weld. 3 FEM models are Butt model, Repair model and Overlay model and are the plane.strain 2D model. The thermal analysis and the stress analysis are performed on each model and the residual stresses on each model were calculated and compared respectively. Also, the specimen of Butt model was made and the residual stresses were measured by X-Ray method and Hole Drilling Technique. These results were compared with the FEM result of Butt model.

• Corrosion Science and Technology
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• 제3권1호
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• pp.30-33
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• 2004

It has been reported that some events of a rupture of seam generator tube have occurred in nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator(SG) tubings. Primary water stress corrosion cracking(PWSCC) of steam generator tubings have occurred in many tubes in Korean plant, and they were repaired using sleeves or plugs, In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high pressure leak and burst testing system was manufactured. Various types of Electro Discharged Machined (EDM) notches were developed on the SG tubes. Leak rate and burst pressure were measured on the tubes at room temperature. Burst pressure of the part through wall defected tubes depends on the defect depth, Water flow rates after the burst were independent of the t1aw types; tubes having 20 to 60 mm long EDM notches showed similar flow rates regardless of the defect depth. A fast pressurization rate gave the tube a lower burst pressure than the case of a slow pressurization.

• 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.

• Corrosion Science and Technology
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• 제21권6호
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• pp.466-475
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• 2022

Surface oxides and intergranular (IG) oxidation phenomena in Alloy 600 depending on hydrogen concentration were characterized to obtain clear insight into the primary water stress corrosion cracking (PWSCC) behavior upon exposure to pressurized water reactor primary water. When hydrogen concentration was between 5 and 30 cm³ H₂/kg H₂O, NiFe₂O₄ and NiO type oxides were found on the surface. NiO type oxides were found inside the oxidized grain boundary when hydrogen concentration was 5 cm³ H₂/kg H₂O. However, only NiFe₂O₄ spinel on the surface and Ni enrichment were observed when hydrogen concentration was 30 cm³ H₂/kg H₂O. These results indicate that the oxidation/reduction reaction of Ni in Alloy 600 depending on hydrogen concentration can considerably affect surface oxidation behavior. It appears that the formation of NiO type oxides in a Ni oxidation state and Ni enrichment in a Ni reduction (or metallic) state are common in primary water. It is believed that the above different oxidation/reduction reactions of Ni in Alloy 600 depending on hydrogen concentration can also significantly affect the resistance to PWSCC of Alloy 600.

• 대한기계학회논문집A
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• 제38권6호
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• pp.637-647
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• 2014

국내 가압경수로형 원자로의 압력용기 상부헤드에는 많은 제어봉구동장치(CRDM) 노즐이 분포한다. 이들 노즐은 억지끼워맞춤(Shrink fitting) 방식으로 결합되어 용접 처리 된다. 용접에 의해 발생되는 인장잔류응력은 일차수응력부식균열을 발생시키는 주요 요인이다. 이러한 이유로 최근 15 여 년 동안 관통노즐 용접부 부위에서 균열 발생 사례가 증가하고 있으며, 이를 극복하기 위해 다양한 방안이 모색되고 있다. 또한 용접과정에서 발생되는 불필요한 결함은 일차수응력부식균열(PWSCC)을 가속화 시키는 원인이 되기도 한다. 원자로 제작과정에서 용접에 의한 결함은 보수용접에 의해 즉시 수리가 이루어 진다. 기존의 연구에서는 정상적인 용접과정에서 발생되는 잔류응력을 예측하였으나, 본 연구에서는 용접과정에서 발생되는 결함을 보수하기 위해 실시되는 보수용접이 용접잔류응력에 미치는 영향을 분석하였다.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2010년도 춘계학술발표대회 초록집
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• pp.84-92
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• 2010

현재 가동 중인 몇몇 가압 경수로 원전 안전 1등급 설비의 이종금속 용접부는 일차수응력부식균열(PWSCC : Primary Stress Corrosion Cracking) 발생의 세가지 조건(민감 재질, 부식 환경, 인장응력)을 동시에 충족하고 있다. 즉, 이종금속 용접부는 PWSCC에 민감한 재질인 Alloy 600 계열 합금으로 제작 또는 용접되어 있으며 고온 수화학 부식 환경 하에 놓여있다. 아울러 오스테나이트 스테인리스 강의 예민화 예방을 위한 용접 후열처리 미실시로 높은 인장 용접 잔류응력이 작용하고 있다. 이러한 이종금속 용접부의 특성상 PWSCC가 발생할 잠재성이 있을 뿐만 아니라 국내외적으로 Alloy 600 계열 합금으로 제작 및 용접된 가압 경수로 원전 안전 1등급 설비의 이종금속 용접부에 실제 PWSCC가 발생된 사례들이 다수 보고되고 있다. 운전 환경 및 재질 변화 없이 PWSCC 발생을 예방하기 위해서는 인장 잔류응력을 이완시켜 낮은 인장 또는 압축 응력화하여야 한다. 이러한 인장 잔류응력 이완방법들로는 PWOL(Pre-emptive Weld Overlay), 레이저 피닝(Laser Peening), MSIP(Mechanical Stress Improvement Process), 워터 제트 피닝(Water Jet Peening), IHSI(Induction Heating Stress Improvement) 방법들이 있는데 공정 시간이 짧고 열 에너지 원이 필요 없으며 전체적인 소성 변형을 야기시키지 않는 레이저 피닝을 본 연구의 대상 방법으로 한다. 본 연구에서는 동적 유한요소 해석을 통해 용접 잔류응력을 이완시키는 레이저 피닝의 효과를 검증하고 용접 잔류응력에 미치는 레이저 피닝 변수의 영향을 고찰하고자 한다. 내부 보수용접이 수행된 경수로 원전 가압기 노즐 이종금속 용접부에 레이저 피닝을 적용한 경우에 대해 상용 유한요소 해석 프로그램인 ABAQUS를 이용하여 동적 유한요소해석을 수행한 결과, 고온 수화학 일차수와 접하는 Alloy 600 계열 합금 내면에서의 인장 잔류응력이 상당히 이완됨을 확인하였다. 또한, 최대충격 압력이 증가할수록, 충격압력 지속시간이 증가할수록, 레이저 스팟 직경이 증가할수록 내표면 인장 잔류응력 이완 정도는 감소하나 이완되는 영역의 깊이는 증가함을 알 수 있다. 또한, 레이저 피닝 방향이 잔류응력 이완에 미치는 영향은 미미함을 알 수 있다.

• 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.