• 대한기계학회논문집A
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• 제34권3호
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• pp.333-340
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• 2010

본 논문은 Alloy 82/182를 용가재로 사용한 SA508 Gr.3 노즐과 F316L 스테인리스강 안전단 사이의 이종금속 용접부에서 상온의 인장물성치와 파괴물성치를 평가하였다. 이를 위해서 두 모재와 열영향부, 버터링부, 그리고 용접부 내의 각 위치에서 채취된 시편을 이용하여 인장시험과 J-R 시험을 수행하였다. 시험 결과 Alloy 82/182 용접 루터부가 상부에 비해 높 은 강도를 보였다. 용접 루터부에서는 항복강도와 인장강도가 위치에 따라 크게 변화하였다. 버터링 부분이 가장 낮은 강도를 보였으며, F316L 스테이리스강 인접부로 갈수록 강도가 증가하였다. 반면, 용접 상부에서는 위치에 따른 강도의 변화가 크지 않았다. Alloy 82/182 용접부는 모재와 열영향부에 비해 낮은 파괴인성을 보였다. Alloy 82/182 용접부 내에서는 용접부 중심이 용접 경계부와 버터링부에 비해 약간 낮은 파괴인성을 보였으며, 용접 루터부가 용접 상부에 비해 높은 파괴인성을 보였다.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.228-237
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• 2019

The crack growth responses of as-received and as-welded Alloy 600/182 and Alloy 690/152 welds to constant loading were measured by a direct current potential drop method using compact tension specimens in primary water at $325^{\circ}C$ simulating the normal operating conditions of a nuclear power plant. The as-received Alloy 600 showed crack growth rates (CGRs) between $9.6{\times}10^{-9}mm/s$ and $3.8{\times}10^{-8}mm/s$, and the as-welded Alloy 182 had CGRs between $7.9{\times}10^{-8}mm/s$ and $7.5{\times}10^{-7}mm/s$ within the range of the applied loadings. These results indicate that Alloys 600 and 182 are susceptible to cracking. The average CGR of the as-welded Alloy 152 was found to be $2.8{\times}10^{-9}mm/s$. Therefore, Alloy 152 was proven to be highly resistant to cracking. The as-received Alloy 690 showed no crack growth even with an inhomogeneous banded microstructure. The cracking mode of Alloys 600 and 182 was an intergranular cracking; however, Alloy 152 was revealed to have a mixed (intergranular + transgranular) cracking mode. It appears that the Cr concentration and the microstructural features significantly affect the cracking resistance and the cracking behavior of Ni-base alloys in PWR primary water.

• 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.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.107-115
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• 2018

To ensure the structural integrity of nuclear power plants, it is essential to predict the lifetime of Alloy 182 weld, which is used for welding in nuclear reactors. The lifetime of Alloy 182 weld is directly related to the crack initiation time. Owing to the large time scatter in most crack initiation tests, a probabilistic model, such as the Weibull distribution, has mainly been adopted for prediction. However, since statistically more advanced methods than current typical methods may be applied, we suggest a statistical procedure for parameter estimation of the crack initiation time of Alloy 182 weld, considering right-censored data and the covariate effect. Furthermore, we suggest a procedure for uncertainty evaluation of the estimators based on the bootstrap method. The suggested statistical procedure can be applied not only to Alloy 182 weld but also to any material degradation data set including right-censored data with covariate effect.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1909-1923
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• 2021

We experimentally investigated the effects of thermal aging and cold work on the microstructure, mechanical properties, and primary water stress corrosion cracking (PWSCC) initiation time for Alloy 182 welds. The effects of thermal aging and cold work on the PWSCC initiation time of Alloy 182 were modeled based on the plastic energy concept and the PWSCC initiation data of this study and previous reports by considering censored data. Based on the results, it is estimated that the PWSCC resistance of the Alloy 182 weld firstly increases and then decreases with thermal aging time when the applied stress is kept constant.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2009년 추계학술발표대회
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• pp.16-16
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• 2009

Alloy 82/182로 용접된 원자력 발전소 주기기의 이종 금속 용접부는 장기간 운전 후 응력부식균열(SCC : Stress Corrosion Cracking)에 의한 결함이 나타나게 된다. 2000년대 이후로 원자력 주기기 Alloy 82/182 용접부에서 PWSCC(Primary Water Stress Corrosion Cracking)에 의한 Degradation이 급격히 증가하는 추세를 보이고 있으며, 국내에서도 이와 관련하여 원자력 발전소의 안전성에 대한 Issue 및 대비책에 대한 관심이 고조되고 있다. 이러한 Alloy 600 용접부에 대한 결함을 예방하기 위한 대표적인 기술로써 수명연장 오버레이 기술이 있다. 원자력 주기기 노즐부는 저탄소강으로 제작되어 있으며, 저탄소강에는 제작 시 용접후열처리가 적용된다. 후열처리를 하는 주된 이유는 Tempering을 통해 열영향부의 인성 및 연성의 회복과 강도를 감소시켜 모재와 동등 또는 이 이상의 물성을 갖도록 하는 데 그 목적이 있다. 그러나 수명연장 오버레이의 경우 현장 작업 시에 후열처리가 어렵기 때문에, 이를 대체하기 위한 기술로 템퍼비드 용접을 적용할 경우 후열처리를 면제해 주고 있다. 본 연구에서는 수명연장 오버레이 기술 개발의 일환으로써 저 탄소강에 대한 템퍼비드 용접 기술을 확립하였다. 실험에 사용된 모재는 원자력 주기기의 노즐에 사용되는 SA508 Gr.3 Cl.1을 사용하였으며, 용가재는 Alloy 52 및 52M을 사용하였다. 최적 조건 도출을 위해서 실험 매트릭스를 이용하여 기본 실험을 수행하였으며, 실험에는 자동 GTAW 용접을 적용하였다. 기본 실험을 통해 얻은 최적 조건을 사용하여 PQ 시험을 수행하여 WPS를 확보하였다. 분석은 용접 후 조직 및 경도 시험, 물리시험(인장시험, 굽힘시험 및 충격시험)을 수행하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.73-78
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• 2008

The paper presents the characteristics of mechanical properties within the heat affected zone (HAZ) of dissimilar metal weld between SA508 Gr.1a and F316 stainless steel (SS) with Alloy 82/182 filler metal. Tensile tests were performed using small-size specimens taken from the HAZ regions close to both fusion lines of weld, and the micro-structures were examined using optical microscope (OM) and transmission microscope (TEM). The results showed that significant gradients of the yield stress (YS), ultimate tensile stress (UTS), and elongations were observed within the HAZ of SA508 Gr.1a. This was attributed to the different microstructures within the HAZ. In the HAZ of F316 SS, however, the welding effect dominated the YS and elongation rather than UTS, and TEM micrographs conformed the strengthening in the HAZ of F316 SS was associated with a dislocation-induced strain hardening.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.102-114
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• 2015

Background: A technique developed to produce artificial intergranular stress corrosion cracks in structural components was applied to thick, forged alloy 600 base and alloy 182 weld metals for use in the qualification of nondestructive examination techniques for welded components in nuclear power plants. Methods: An externally controlled procedure was demonstrated to produce intergranular stress corrosion cracks that are comparable to service-induced cracks in both the base and weld metals. During the process of crack generation, an online direct current potential drop method using array probes was used to measure and monitor the sizes and shapes of the cracks. Results: A microstructural characterization of the produced cracks revealed realistic conformation of the crack faces unlike those in machined notches produced by an electrodischarge machine or simple fatigue loading using a universal testing machine. Conclusion: A comparison with a destructive metallographic examination showed that the characteristics, orientations, and sizes of the intergranular cracks produced in this study are highly reproducible.

• Journal of Welding and Joining
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• 제33권2호
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• pp.40-46
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• 2015

This study was performed to investigate the effect of ultrasonic nano crystal surface modification (UNSM) on residual stress mitigation after Weld Inlay repair for butt dissimilar metal weld with Alloy 82/182 in reactor vessel In/Outlet nozzle. As-welded and Weld Inlay specimens were made in accordance with design standard of ASME Code Case N-766, and two planes of their weld specimens were peened by the optimum UNSM process condition. Peening characteristics for weld specimens after UNSM treatment were evaluated by surface roughness and Vickers hardness test. And, residual stress for weld specimens developed from before and after UNSM treatment was measured and evaluated by instrumented indentation technique. Consequently, it was revealed that the mitigation of residual stress in weld metal after Weld Inlay repair of reactor vessel In/Outlet nozzle could be possible through UNSM treatment.

• 한국안전학회지
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• 제23권6호
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• pp.28-33
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• 2008

This paper presents the characteristics of mechanical properties within the heat affected zones(HAZs) of dissimilar metal weld joint between SA508 Gr.1a and F3l6 stainless steel(SS) with Alloy 82/182 filler metal. Tensile tests were performed using small-size specimens taken from the heat affected regions close to both fusion lines of weld, and the micro-structures were examined using optical microscope(OM) and transmission microscope(TEM). The results showed that significant gradients of the yield stress(YS), ultimate tensile stress(UTS), and elongations were observed within the HAZ of SA508 Gr.1a. This was attributed to the different microstructures within the HAZ developed during the welding process. In the HAZ of F316 SS, however, the welding effect dominated the YS and elongation rather than UTS. TEM micrographs demonstrated these characteristics of the HAZ of F316 SS was associated with a dislocation-induced strain hardening.