• Journal of Welding and Joining
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• 제23권6호
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• pp.43-48
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• 2005

This work was carried out to obtain sound welds and to select a most suitable binary metal joint among three different dissimilar binary metal combinations such as Zr-4/Ta, Mo/Ta and Ti/Ta(seal tube/sensor sheath) joints fur the instrumented nuclear fuel irradiation test. To do this, Taguchi experimental method was employed to optimize the experimental data. In addition, metallography, micro-focus x-ray radiography and hardness test were conducted to examine the welds. From the weld bead appearance, penetration depth and bead width as well as weld defects standpoint, Zr-4/Ta joint is suggested for the circumferential joining between a seal tube and a sensor sheath. The optimized welding parameters based on Zr-4/Ta joint are suggested as well.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.20-24
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• 2015

Steam generator(S/G) tubes in pressurized water reactor (PWR's) are subject to several types of degradation. This degradation includes denting, pitting, intergranular attack(IGA), intergranular stress corrosion cracking(IGSCC), fatigue, fretting and wear. Degradation can be derived from either the primary side(inside) or the secondary side(outside) of the tube. Recent issue for tube degradation in domestic steam generator is the tube end cracking on seal weld region. The seal weld region at the tube end and tube itself is regarded as a pressure boundary between the primary side and the secondary side. One of the Westinghouse Model-F S/G has experienced tube end cracking and its number of plugging approximately becomes to the operating limit up to 5% due to tube end cracking which was reported as SAI/MAI(single/multiple axial indication) or SCI/MCI(Single/multiple circumferential indication) from the results of eddy current testing. Eddy current mock-up test was carried out to determine the origin of cracking whether it is from weld zone area or parent tube. This result was helpful to analyze crack location on ECT data. Correct action on this problem was the installation of tube-end sleeve. Last year, after removing 340 installed plugs from tubes, selected 269 tubes took tube-end sleeve installation. Tube-end sleeve brought pressure boundary from parent tube to installed sleeve tube. Tube-end sleeve has the benefit of reducing outage period and increasing more revenue than replacing S/G. This paper is provided to assist interest parties in effectively understanding this issue.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.408-422
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• 1999

To reduce the fast neutron fluence which deteriorates the RPV integrity, additional shields were assumed to be installed at the outer core structures of the Kori Unit 1 reactor, and its reduction effects were examined. Full scope Monte Carlo simulation with MCNP4A code was made to estimate the fast neutron fluence at the RPV. An optimized design option was found from various choices in geometry and material for shield structure. It was expected that magnitude of fast neutron fluence would be reduced by 39% at the circumferential weld of the RPV, resulting in extension of plant lifetime by 4.6 EFPYs based on the criterion of PTS requirement It was investigated that the nuclear characteristics and thermal hydraulic factors at the internal core were only negligibly influenced by the installation of additional shield structure.

• 한국안전학회지
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• 제15권2호
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• pp.51-58
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• 2000

Recently, the J/T analysis based on elastic-plastic finite element analysis is popularly used in the nuclear industry to assess the integrity of a cracked pipe. The objective of this paper is to evaluate the effect of stress-strain curve for weld metal, variation of crack incremental length(${\delta}a$), and crack face pressure on the J/T analysis result. For this purpose, a parametric analysis was performed and the results calculated from finite element analysis were compared with those from the piping experimental data(stainless steel weldment pipe with circumferential through-wall crack). The numerical result using base metal material property is in agreement with the experimental one and the maximum load is decreased as the ${\delta}a$ for J/T analysis is increased.

• 비파괴검사학회지
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• 제4권1호
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• pp.5-10
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• 1984

One of the cylindrical reactor vessels in petrochemical plants was examined by acoustic emission method. The vessel was quiet in view of A.E. activity throughout the pressure range $12-44kg.f/cm^{2}.G$. Above the pressure of $44kg.f/cm^2$, some events were appeared lower than 30 counts. In order to verify the events, other Nondestructive testing methods were performed and a surface crack, 10mm in length and 0.8mm in depth, was found on the outside surface of circumferential weld.

• 대한기계학회논문집A
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• 제35권10호
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• pp.1337-1345
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• 2011

가압경수로형 원자로의 원자로압력용기 상부헤드에는 많은 제어봉구동장치(CRDM) 노즐이 분포한다. 최근 10 여 년 동안 제어봉구동장치 alloy 600 CRDM 노즐에서 균열 발생 사례가 증가하고 있으며, 이는 용접과 연관성이 매우 깊은 것으로 알려져 있다. CRDM 노즐에서 발생하는 축 및 원주방향 균열은 유럽과 미국의 원자력 발전소에서 발견되었으며, 사고의 원인은 용접 잔류응력 및 작용하중에 기인하는 일차수응력부식균열(PWSCC)임이 확인되었다. 이러한 이유로 본 연구에서는 유한요소해석을 통해 한국형 원자로의 CRDM 관통 노즐 용접부를 대상으로 용접 잔류응력을 예측하였으며, 특히, 관통노즐의 위치와 형상, 용접부 필렛 형상 및 인접노즐 용접에 의한 영향을 분석하였다.

• 비파괴검사학회지
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• 제31권5호
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• pp.536-542
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• 2011

원자력발전소의 이종금속 용접부는 PWSCC 결함에 민감한 것으로 알려져 있으며 기량검증된 검사자가 기량검증된 절차서를 사용하여 가동중검사 기간 중에 주기적인 검사를 수행하고 있다. 국내 원자력발전소 이종금속 용접부의 형상 조사 결과에 따르면 대부분의 이종금속 용접부가 경사진 노즐부나 인접부에 위치하는 것으로 나타났다. 일반적인 초음파 탐촉자를 사용하여 경사부위에 위치한 이종금속 용접부의 검사를 수행할 경우 초음파 탐촉자의 접근성이 제한되어 검사체적을 모두 검사하기가 어렵다. 특히 축방향 결함 검출을 위한 원주방향 주사에서는 초음파 탐촉자가 경사면에 위치하게 되면 반사체로부터 결함 신호를 얻기 가 어려우며 이에 따라 결함 검출이 어렵게 된다. 이러한 문제점을 극복하기 위해서는 경사면을 고려하여 비틀림 각도를 적용한 초음파 탐촉자를 사용하는 것이 필요하다. 모델링을 통하여 비틀림 굴절 종파탐촉자 를 설계하고 축방향 결함 검사용 비틀림 굴절 종파탐촉자를 제작하여 실험을 통하여 결함으로부터 신호를 취득하였다. 일반 탐촉자와 비틀림 각도가 적용된 탐촉자의 초음파 응답신호를 비교한 결과 비틀림 각도가 적용된 탐촉자의 초음파 응답신호가 훨씬 뛰어난 것으로 나타났다.

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

• 한국압력기기공학회 논문집
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• 제6권1호
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• pp.16-22
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• 2010

Cast austenitic stainless steel(CASS) is used in the primary cooling piping system of nuclear power plant for it's relative low cost, corrosion resistance and easy of welding. However, the coarse-grain structure of cast austenitic stainless steel can strongly affect the inspectability of ultrasonic testing. The major problems encountered during inspection are beam skewing, high attenuation and high background noise of CASS component. So far, the best inspection performance involving CASS components have been achieved using low frequency TRL(Transmitter/Receiver side-by-side L wave) angle beam probe. But TRL technique could not detect shallow defect and it contains an uncertainty for sizing capability. Currently, most of researchers are studying to overcome these challenge issue. In this study, low-frequency phased array TRL technique used to detect and sizing the flaws in CF8A cast austenitic stainless steel.As conclusion, we could detect and size not only axial flaw but also circumferential flaw using low frequency phased array technique.

• 대한기계학회논문집
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• 제19권4호
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• pp.929-940
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• 1995

The fatigue strength and fracture topography in the friction welded interface of high speed steel (HSS-Co) to SM55C carbon steel have been investigated through the fatigue test, SEM fractograph and EDS (energy dispersive spectrometer) analysis. Three kinds of specimens used in this research are the friction welded joints, HSS-Co and SM55C carbon steel with circumferential notch, saw notch and smooth, respectively. The notch sensitivity factor, .eta. of the friction welded joints is lower than that of the base materials, and that represents a superiority of the joint performance of FRW. Fracture topography of the FRW specimens with a notch showed a cleavage or brittle appearance, while that of the FRW smooth specimen appeared to be ductile. Furthermore, although fatigue crack likely initiated near the weld interface of the FRW smooth specimen, crack propagation continued into the HAZ of SM55C steel. Finally, fatigue fractures of the base materials were associated primarily with the inclusions located at the outer periphery of the specimen.