• 한국압력기기공학회 논문집
• /
• 제6권1호
• /
• pp.83-88
• /
• 2010

Domestic nuclear steam generators with Alloy 600 HTMA tubes have experienced axial cracking at eggcrate tube support plates(TSPs). The axial stress corrosion cracks were observed at the crevice between outside of tubes and eggcrate TSPs. The root cause of axial cracking was investigated by thermal hydraulic analysis and sludge distribution diagnosis. It is suggested that deposition of sludge at eggcrate TSPs could increase the outside surface temperature of tube and promote the enrichment of impurities at crevice, and thus accelerate cracking. Additionally strategy for reducing the sludge ingress to steam generators is discussed.

• 한국안전학회지
• /
• 제15권3호
• /
• pp.57-65
• /
• 2000

The stress corrosion crack growth is simulated assuming a small axial surface crack inside a S/G tube. Internal pressure and residual stresses are considered as applied forces. Stress intensity factors along crack front, variation of crack shape and crack growth rate are obtained and discussed. It is noted that the aspect ratio of the crack is not depend on the initial crack shape but depend on the residual stress distribution.

• 한국안전학회지
• /
• 제30권4호
• /
• pp.168-173
• /
• 2015

A steam generator tubing is one of the main pressure boundary of the reactor coolant system in the nuclear power plants. Structural integrity refers to maintaining adequate margins against failure of the tubing. Burst pressure of a tube at tube support plate can be higher than that for a free-span tube because failure behaviors could be interfered from the tube support plate. Alternative repair criteria for out-diameter stress corrosion cracking indications in tubes to the drilled type tube support plate were developed, however, there are very limited information to the eggcrate type tube support plate. This paper discussed reinforcement effect of steam generator tube burst pressure with axial out-diameter stress corrosion cracking within an eggcrate type tube support plate. A series of tube burst tests were performed under the room temperature and it was found out that there is no significant but marginal effects.

• 한국공작기계학회:학술대회논문집
• /
• 한국공작기계학회 1999년도 춘계학술대회 논문집
• /
• pp.163-169
• /
• 1999

Residual stresses causing stress corrosion cracking (SCC) of thin-walled steam generator U tubes were investigated. The residual stresses were measured by hole drilling methods, and the applied stresses resulting from the internal pressure and the temperature gradient in the steam generator were estimated theoretically. In U-bent regions, the residual stresses at extrados were induced with compressive stress(-), and its maximum value reached -319MPa in axial direction at $\phi$= $0^{\circ}$ in position. Maximum tensile residual stress of 170MPa was found to be at the flank side at position of $\phi$= $90^{\circ}$, i.e., at apex region. Hoop stress due to the pressure and temperature differences between primary and secondary side were analyzed to be 76 MPa and 45 MPa, respectively.

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

• Corrosion Science and Technology
• /
• 제13권3호
• /
• pp.112-119
• /
• 2014

In this work, it is aimed to develop the fabrication method of axial stress corrosion cracking (SCC) defects having various sizes, on the outer diameter surface of the steam generator (SG) tubings. To control the length of the artificial SCC defect, the specific area of the SG tubing samples was exposed to an acidic solution after a sensitization heat treatment. During the exposure to an acidic solution, a direct current potential drop (DCPD) method was adopted to monitor the crack depth. The size of the SCC defect was first evaluated by an eddy current test (ECT), and then confirmed by a destructive examination. From the comparison, it was found that the actual crack length was well controlled to be similar to the length of the surface exposed to an acidic solution (5, 10, 20 or 30 mm in this work) with small standard deviation. From in-situ monitoring of the crack depth using the DCPD method, it was possible to distinguish a non-through wall crack from a through wall crack, even though the depth of the non-through wall crack was not able to be precisely controlled. The fabrication method established in this work was useful to simulate the SCC defect having similar size and ECT signals as compared to the field cracks in the SG tubings of the operating Korean PWRs.

• 한국압력기기공학회 논문집
• /
• 제6권2호
• /
• pp.34-41
• /
• 2010

This work is concerned with the analysis of stress corrosion cracking(SCC) behavior of Alloy 600 nozzle penetration mock-up according to a residual stress induced by a dissimilar metal welding(DMW) in a nuclear reactor pressure vessel. The effects of the dimension and materials of the nozzle penetration on the deformation and the residual stress induced by DMW were investigated using a finite element analysis(FEA). The inner diameter(ID) change of the nozzle by DMW and its dependance on the design variables, calculated by FEA, were well consistent with those measured from the mock-up. Accelerated SCC tests were performed for three mock-ups with different wall thicknesses in a highly acidic solution to investigate mainly the effect of the residual stress on the SCC behavior of Alloy 600 nozzle. From a destructive examination of the mock-up after the tests, the SCC behavior of the nozzle was fairly related with the residual stress induced by DMW : axial cracks were found in the ID surface of the nozzle within the J-weld region where the highest tensile hoop stress was predicted by FEA, while circumferential cracks were observed beyond both J-weld root and toe where the highest tensile axial stress was expected.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2008년도 추계학술대회A
• /
• pp.725-726
• /
• 2008

This paper proposes elastic stress intensity factors and crack opening displacements (CODs) for a slanted axial through-wall cracked cylinder under an internal pressure based on detailed 3-dimensional (3-D) elastic finite element (FE) analyses. Based on the elastic FE results, the stress intensity factors along the crack front and CODs through the thickness at the center of the crack were provided. These values were also tabulated for three selected points, i.e., the inner and outer surfaces and at the mid-thickness. The present results can be used to evaluate the crack growth rate and leak rate of a slanted axial through-wall crack due to stress corrosion cracking and fatigue. Moreover, the present results can be used to perform a detailed Leak-Before-Break analysis considering more realistic crack shape development.

• 비파괴검사학회지
• /
• 제29권4호
• /
• pp.374-382
• /
• 2009

원자력발전소의 1차측 및 2차측 냉각계의 장벽 역할을 하는 핵심 설비중 하나인 증기발생기(steam generator, SG) 전열관은 공공의 사회적 안전성과 효율적인 발전 용량을 유지하기 위해 구조적 건전성을 유지하여야 한다. 또한 결함을 함유하고 있는 전열관은 해당결함을 조기에 검출, 정량적으로 결함을 평가하여 필요한 경우에는 보수조치를 수행하여야 한다. 이러한 결함의 검출 및 정량화를 위해서 검사관련 고시 및 강화된 SG 관리프로그램(SGMP)에 근거하여 와전류탐상검사법(eddy current testing, ECT)을 적용, 검사를 수행하고 있다. SG 전열관에서 검출되고 있는 결함중 응력부식균열(stress corrosion cracking, SCC)은 미세한 경우 결함의 검출이 어려울 뿐 아니라 생성된 결함의 성장속도가 빠르기 때문에 SG 전열관의 건전성을 위협하는 주요결함 기구중 하나로 분류하고 있다. 본 논문에서는 다양한 결함 깊이 및 길이별로 방전가공(electric discharge machining, EDM)된 축방향 ODSCC에 대해 pancake, +point 및 shielded pancake 코일 등이 탑재된 3 coil형태의 +PT MRPC(motorized rotating pancake coils)를 적용하여 결함의 검출가능 여부 및 크기 측정을 위한 검사를 수행하였으며 본 실험결과를 통해 SG 전열관의 건전성 및 원전 운전의 안전성을 진단하는 공학적 평가 자료로써의 활용 가능성 뿐 아니라 와전류탐상검사의 신뢰도 향상을 도모하고자 하였다.

• 한국압력기기공학회 논문집
• /
• 제8권1호
• /
• pp.19-24
• /
• 2012

This paper presents the results of finite element analysis for the effects of hydrostatic test and operating condition on the weld residual stress at dissimilar metal weld of a pressurizer safety nozzle in a nuclear power plant. For the study, the weld residual stress at ambient condition was analyzed using ABAQUS in the first place. After the weld residual stress analysis, the hydrostatic test condition and operating condition was applied to the same model one after another. The weld residual stress was observed to change due to the successive hydrostatic test and operating condition. The axial residual stresses on inner surface of the dissimilar metal weld and HAZ region were decreased by hydrostatic test and operating condition, which gives beneficial effect on preventing primary water stress corrosion cracking.