유한요소법을 이용한 용접공정 모사 시 입열 방법에 따른 용접잔류응력의 영향

양준석(Jun-Seog Yang),박치용(Chi-Yong Park),이경수(Kyoung-Soo Lee) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.11

This paper is to discuss distribution of welding residual stresses of a ferritic low alloy steel nozzle with dissimilar metal weld using Alloy 82/182. Two dimensional (2D) thermo-mechanical finite element analyses are carried out to simulate multi-pass welding process on the basis of the detailed and fabrication data. On performing the welding analysis generally, the characteristics on the heat input and heat transfer of weld are affected on the weld residual stress analyses. Thermal analyses in the welding heat cycle process is very important process in weld residual stress analyses. Therefore, heat is rapidly input to the weld pass material, using internal volumetric heat generation, at a rate which raises the peak weld metal temperature to 2200 ℃ and the base metal adjacent to the weld to about 1400 ℃. These are approximately the temperature that the weld metal and surrounding base materials reach during welding. Also, According to the various ways of appling the weld heat source, the predicted residual stress results are compared with measured axial, hoop and radial through-wall profiles in the heat affected zone of test component. Also, those results are compared with those of full 3-dimensional simulation.

원자력발전소 1,2등급 배관의 LBB 소급 적용 시 PWSCC 선별기준에 대한 고찰

양준석(Jun-Seog Yang),박치용(Chi-Yong Park),김태룡(Tae-Ryong Kim) 대한기계학회 2005 대한기계학회 춘추학술대회 Vol.2005 No.5

This paper presents methodology and procedure of the Leak-Before-Break (LBB) screening criteria of primary water stress corrosion crack (PWSCC) to be met necessarily to backfit LBB technology for nuclear power plant class 1, 2 piping in operation. Those reviewed in this paper are the susceptibility of pipe rupture due to the PWSCC. This paper considered the possibility of new flaws to be founded by the new examination method, even though no such flaws have been identified by nondestructive tests. Crack growth evaluation on PWSCC was carried out for several flaw shapes of both axial and circumferential postulated flaws by using the steady-state stresses including residual stresses.

원자력발전소 가압기 하부헤드 구조건전성 평가

양준석(Jun-Seog Yang),박치용(Chi-Yong Park),강선예(Seon-ye Kang) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.5

원자력발전소 가압기와 고온루프배관의 운전조건에 따른 계통온도차로 인해서 밀림관에서는 서로 다른 밀도차에 의해서 층이 분리된 채 존재하는 열성층 현상이 발생한다. 원자력발전소 운전 중 가장 큰 온도차는 가열 및 냉각 운전기간에 발생하며 이 때 원자로냉각재가 가압기로 유입될 때 가압기로 들어오는 차가운 유체는 가압기 밀림노즐과 하부헤드를 냉각하면서 온도구배를 야기시키며 반대로 유출이 되는 경우 같은 부위를 다시 가열시킬 수 있다. 본 논문에서는 가열 및 냉각 운전기간 동안 원자로냉각재가 유입 또는 유출되는 과도조건을 작용하여 ASME Code Section Ⅲ 방법론에 따라 가압기 하부헤드의 구조건전성을 평가하였다. 평가 결과 가압기 하부헤드 모든 영역에서 설계수명기간 동안 ASME Code Sec. Ⅲ의 허용요건을 만족하였다. When the temperature difference between the reactor coolant system and the pressurizer during heat-up and cool down in the nuclear power plant is large, these out of limit conditions can produce significant temperature transients in the pressurizer lower head. This paper evaluates the impact of pressurizer out-of-limit insurge/outsurge transients on the structural integrity of the pressurizer lower head in the nuclear power plant. The stresses to be considered are those resulting from the specified loadings of internal pressure, deadweight, thermal, seismic, and thermal gradients. The evaluation on the primary plus secondary stress intensity ranges and the cumulative fatigue usage factors were included. The calculated stress intensity range and usage factor values are compared with the applicable limits of Section Ⅲ of the ASME Boiler and Pressure Vessel Code. According to the evaluation results, all requirements of the ASME Section Ⅲ, Subsection NB are satisfied.

Type CF8M 강 열취화 건전성 평가

양준석(Jun-Seog Yang),박치용(Chi-Yong Park),정일석(Il-Seog Jeong),배시연(Shi-Yeon Bye) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.11

가압경수형 원전의 주배관 엘보 제작에 사용된 스테인리스주강은 원자력발전소 가동온도에서 재료 열화의 종류인 열취화에 민감하다. 스테인리스주강의 열취화는 페라이트함량에 따라 민감도를 판단할 수 있다. SA351 CF8 계열의 스테인리스주강은 몰리브덴의 함량에 따라 열취화의 정도를 판단할 수 있다. 스테인리스주강으로 제작된 배관이 원자력발전소 가동기간 동안 고온에서 장시간 노출되면 배관재의 인장강도는 증가한 반면 파괴인성은 감소한다. 따라서 원자력발전소 주배관 엘보에 사용되는 스테인리스주강의 결함 평가시 반드시 열취화 현상을 고려해야 한다. 본 논문에서는 CF8M으로 제작된 원자력발전소 1차계통 주배관 엘보의 재질에 대한 열취화 건전성을 평가하고 그 결과를 토의하였다. Cast stainless steel used in the fabrication of the primary loop elbows of pressurizer water reactors is subject to embrittlement due to thermal aging at the reactor service temperature. The susceptibility of the material to thermal aging increases with increasing ferrite contents. The American Society of Mechanical Engineers (ASME) Specification SA-351 Grade CF-8M contains a larger amount of molybdenum and shows increased susceptibility to thermal aging, compared with the other cast stainless steel grades such as CF-8 and CF-8A. Because of the potential for embrittlement, flaw evaluation of Type CF8M cast stainless steel components in reactor applications should consider the effects of thermal aging to ensure structural integrity of the reactor coolant pressure boundary. This paper presents the integrity and life assessment methodologies of Type CF8M cast stainless steel elbow and discuss the results of them.

협개선 용접부에 대한 용접잔류응력 예측

양준석(Jun-Seog Yang),허남수(Nam-Su Huh) 대한기계학회 2010 大韓機械學會論文集A Vol.34 No.1

최근 원자력발전소 배관계의 용접 시 결함 발생의 원인이 되는 용접변형과 용접잔류응력을 감소시키고 용접 효율도 증가시키기 위해 협개선 용접법이 적용되고 있다. 협개선 용접법은 기존 용접법에 비해 상대적으로 입열량이 적고 용접변형과 용접잔류응력도 감소시킬 수 있기 때문에 배관 건전성 관점에서 많은 이점이 있을 수 있다. 그러나 실제 협개선 용접부 결함에 대한 정확한 파괴역학적 분석 등을 위해서는 용접 특성을 고려하여 변형 및 용접잔류응력 특성을 정확하게 예측해야 한다. 따라서 본 논문에서는 협개선 용접부에 대한 상세 2차원 유한요소해석을 수행하여 협개선 용접부의 용접잔류응력 분포를 결정하였으며, 이를 일반적인 용접법에 대한 특성과 비교하였다. 본 논문의 결과는 향후 협개선 용접부의 결함 평가와 용접방법 개선 등을 위해 적용될 수 있다. The conventional welding technique such as shield metal arc welding has been mostly applied to the piping system of the nuclear power plants. It is well known that this welding technique causes the overheating and welding defects due to the large groove angle of weld. On the other hand, the narrow gap welding(NGW) technique has many merits, for instance, the reduction of welding time, the shrinkage of weld and the small deformation of the weld due to the small groove angle and welding bead width comparing with the conventional welds. These characteristics of NGW affect the deformation behavior and the distribution of welding residual stress of NGW, thus it is believed that the residual stress results obtained from conventional welding procedure may not be applied to structural integrity evaluation of NGW. In this paper, the welding residual stress of NGW was predicted using the nonlinear finite element analysis to simulate the thermal and mechanical effects of the NGW. The present results can be used as the important information to perform the flaw evaluation and to improve the weld procedure of NGW.

가압기 밀림관 환경피로평가를 위한 피로보정계수 적용에 관한 연구

양준석(Jun-Seog Yang),박치용(Chi-Yong Park),강선예(Seon-Ye Kang) 대한기계학회 2009 大韓機械學會論文集A Vol.33 No.10

Nuclear power plants applying for the continued operation over design life are required to address the effects of reactor water environment in fatigue design requirement of the ASME Code. Reactor water environmental effects are generally evaluated by calculating fatigue correction factors on fatigue usage. This paper describes the application for pressurizer surge line of environmental fatigue correction factors and the strain rate impact in the application. From this paper, the environmental fatigue correction factors resulted from the assumption of a step change in temperature are especially compared with those calculated from the data measured during plant startup. As a conclusion of this paper, the design transient conditions applied to the fatigue design may be conservative in case of the environmental fatigue evaluation.

협개선 용접부 용접잔류응력 분석

양준석(Jun-Seog Yang),허남수(Nam-Su Huh) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.5

The conventional welding technique such as shield metal arc welding has been mostly applied to the piping system of the nuclear power plants. It is well known that this welding technique causes the overheating and welding defects due to the large groove angle of weld. On the other hand, the narrow gap welding (NGW) technique has many merits, for instance, the reduction of welding time, the shrinkage of weld and the small deformation of the weld due to the small groove angle and welding bead width comparing with the conventional welds. These characteristics of NGW affect the deformation behavior and the distribution of welding residual stress of NGW, thus it is believed that the residual stress results obtained from conventional welding procedure may not be applied to structural integrity evaluation of NGW. In this context, the welding residual stress of NGW was predicted using the nonlinear finite element analysis to simulate the thermal and mechanical effects of the NGW in the present study. The present results can be used as the important information to perform the flaw evaluation and to improve the weld procedure of NGW.

LBB 적용 배관의 용접 잔류응력이 배관 균열열림변위에 미치는 영향 평가

양준석(Jun-Seog Yang),박치용(Chi-Yong Park),이경수(Kyung-Su Lee) 대한기계학회 2008 대한기계학회 춘추학술대회 Vol.2008 No.5

Leak-Before-Break (LBB) concept has been applied to various high energy lines in primary water reactors as an alternate way of addressing the assumption of double ended guillotine break. Following the recent primary water stress corrosion crack (PWSCC) events. the application of LBB at Alloy 82/182 locations has been questioned. The one of the LBB issues on the Alloy 82/182 welds is the high residual stress after welding. Therefore it is very important to evaluate the effect of the weld residual stresses on the crack opening area to apply LBB concept to the nuclear class I piping. Fristly this paper reviewed the effect of the residual stresses of the Alloy 82/182 weld performed in the shop. Secondly stainless steel weld location done at field was investigated incorporated with first Alloy 82/182 weld. Crack opening area was evaluated based on each residual stress result.

원자로냉각재 환경영향을 고려한 고리1호기 가압기 밀림관 피로평가

양준석(Jun-Seog Yang),박치용(Chi-Yong Park),강선예(Seon-Ye Kang) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.5

The original design of Kori Unit 1 did not consider the thermal stratification phenomena in the surge line piping and the insurge / outsurge out of limit in the pressurizer lower head during plant heat up and cooldown condition. For the plant life extension over 20 years, the metal fatigue evaluation considering the effects of reactor water environment was performed to determine the impact of the insurge / outsurge out of transients and thermal stratification transients, in conjunction with design transient effects, on the environmental fatigue usage factors at the critical locations of the surge line and pressurizer surge nozzle. In this paper, the effects of the environmental fatigue were evaluated and discussed by confining to the thermal stratification transients in the surge line and insurge / outsurge transients in the pressurizer surge nozzle. Also, the use of fatigue life correction factor to incorporate the effects of environment into the ASME Code fatigue evaluation for the surge line piping and pressurizer surge nozzle was discussed in this paper.

밀림관 형상에 따른 배관내 열성층 현상 연구

조인수(In-Su Jo),양준석(Jun-Seog Yang),박치용(Chi-Yong Park) 대한기계학회 2010 대한기계학회 춘추학술대회 Vol.2010 No.5