• 한국압력기기공학회 논문집
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• 제8권3호
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• pp.7-12
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• 2012

Steam generator tubes are critical boundary of the primary and secondary side in nuclear power plants. Eddy current testing is commonly used as the method of non-destructive testing for the safety and integrity of steam generator tubes in the nuclear power plants. Changes in the geometric shape act as a stress concentration factor likely to cause a defect during the steam generator operation. The mixed-signals with the geometric shape are distorted and attributes that are difficult to detect signals. An example is bending stress due to compression process at a U-bend occurring in the intrados region which has a small radius of curvature. The resulting change in the geometric shape may lead to a dent like occurrences. The dent can cause stress concentration and generates stress corrosion cracks. In this study, the steam generator tubes of nuclear power plant were selected to study for analysis of mixed-signal containing dent and stress corrosion cracks.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 춘계학술대회논문집A
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• pp.25-30
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• 2000

Residual stresses induced in U-bending and tube-to-tubesheet joining processes of PWR's steam generator row-1 tube were measured by X-ray method and Hole-Drilling Method(HDM). The stresses resulting from the Internal pressure and the temperature gradient in the steam generator were also estimated theoretically. In U-bent lesions, the residual stresses at extrados were induced with compressive stress(-), and its maximum value reached -319 MPa in axial direction at ${\psi}=0^{\circ}$ in position. Maximum tensile residual stress of 170MPa was found to be at the flank side at Position of${\psi}=90^{\circ}$, i.e., at apex region. In tube-to-tubesheet fouling methods, the residual stresses induced by the explosive joint method were found to be lower than that by the mechanical roll method. The gradient of residual stress along the expanded tube was highest at the. transition region, and the residual stress in circumferential direction was found to be higher than the residual stress in axial direction. Hoop stress due to an internal pressure between primary and secondary side was analyzed to be 76 MPa and thermal stress was 45 MPa.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1995년도 가을 학술발표회 논문집
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• pp.61-67
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• 1995

The steam generator, one of the major components in the reactor coolant system, plays an important role in transferring the thermal energy made in the reactor during normal operation to the secondary side and producing steam to drive turbine. A hydraulic snubber system is used in order to protect the steam generator under the dynamic loading condition and to absorb the thermal expansion transmitted by the reactor coolant piping due to high temperature and pressure during normal operation. In this study, the model for a geometrical linkage system is presented to analyze the snubber stroke of the steam generator and the parameters in the snubber stroke analysis are investigated. A method to analyze lever ratio of the linkage system which is required in the process of determining the snubber stiffness value is also presented. To discuss the validation of the suggested analysis, the analysis results are compared with the measured data during the hot functional test for the standardized 1000 Mwe pressurized water reactor plant under the construction.

• 한국산학기술학회논문지
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• 제11권8호
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• pp.3112-3118
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• 2010

최근 경수로형 원전 2차 계통의 건전성 유지를 위해 수처리제를 암모니아에서 에탄올아민으로 전환하였으나, 적용 후 복수 및 저압급수가열기 영역에서의 pH가 감소하므로 본 연구에서는 최적의 pH 제어제로 사용 할 수 있는 아민을 조사하였다. 대체아민 조사 결과 최적 조건을 만족시키는 단일 아민은 존재하지 않았다. 암모니아는 상대휘발도가 높아 증기에 많이 분포되어 증기 응축수인 복수에서 pH가 높으며, 상대휘발도가 낮은 에탄올아민은 습증기 영역의 pH를 높여 유체가속부식을 억제하므로 증기발생기 철 슬러지 유입을 감소하는데 효과적인 것으로 나타났다. 따라서 복수 및 저압급수계통에서 pH가 높은 암모니아와 습증기영역의 유체가속부식 측면에서 특성이 우수한 에탄올아민(ETA)을 혼합 주입하는 복합아민을 선택하면 2차 계통 재질의 손실을 최소화하여 증기발생기 건전성을 확보할 수 있을 것이다.

• 한국소음진동공학회논문집
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• 제12권4호
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• pp.261-271
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• 2002

The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
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• pp.379-380
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• 2002

Steam generators (S/G) of pressurized water reactors are large heat exchangers that use the heat from the primary reactor coolant to make steam in the secondary side for driving turbine generators. Reciprocating sliding wear experiments have been performed to examine the wear properties of Incoloy 800 and Inconel 690 steam generator tubes in high temperature water. In present study, the test rig was designed to examine the reciprocating and rolling wear properties in high temperature (room temperature - $300^{\circ}C$) water. The test was performed at constant applied load and sliding distance to investigate the effect of test temperature on wear properties of steam generator tube materials. To investigate the wear mechanism of material, the worn surfaces were observed using scanning electron microscopy. At $290^{\circ}C$, wear rate of Inconel 690 was higher than that of Incoloy 800. It was assumed to be resulted from the oxide layer property difference due to the a\loy composition difference. Between 25 and $150^{\circ}C$ the wear loss increased with increasing temperature. Beyond $150^{\circ}C$, the wear loss decreased with increasing temperature. The wear loss change with temperature were due to the formation of wear protective oxide layer. From the worn surface observation, texture patterns and wear particle layers were found. As test temperature increased, the proportion of particle layer increased.

• 비파괴검사학회지
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• 제32권5호
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• pp.545-550
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• 2012

원자력발전소 증기발생기 세관은 방사성물질이 외부로 누출되지 않도록 압력경계 역할을 하는 주요부품이다. 설비 운전기간이 증가함에 따라 이차측에서 유입된 슬러지가 증기발생기 2차측 유체 흐름을 따라 상부로 이동하면서 유체비등과 난류에 의해 세관 외면에 스케일이 부착되어 세관열화, 유로홈 막힘 및 열전달을 감소시키는 파울링을 유발하는 원인으로 작용한다. 따라서, 원전 운영자는 세관 외면에 쌓인 스케일의 두께를 확인하여 일정시점이 되면 화학세정 등의 정비를 수행한다. 본 논문에서는 보빈코일 와전류신호를 이용하여 세관 외면에 부착된 스케일 두께를 정량적으로 평가하는 기술을 개발하고자 스케일 시험편을 제작하여 스케일 두께와 와전류신호 진폭 간의 상관관계를 분석하였고, 이를 바탕으로 스케일의 두께를 정량적으로 평가하는 기법과 대량의 와전류 데이터를 평가할 수 있는 프로그램을 개발하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.353-358
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• 2004

Tracing the study of the burst test of steam generator tube, few studies have been reported to effect of external pressure acting on secondary-side in service condition. In this study the burst tests of Inconel 690TT were conducted in order to evaluate burst strength characteristics under the effect of external pressure. We obtained the result that the burst strength of Inconel 690TT increased when external pressure increased while both total circumferential elongation and uniform burst elongation were not affected. Also, according to the increased of external pressure, the size of the burst opening became smaller and the tear was getting severe.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.56-63
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• 2021

The feedwater ring is an assembly in steam generator internal piping, which distributes feedwater into the secondary side of the steam generator. It consists of an assembly of carbon steel piping, pipe fittings and J-nozzles which are inserted into the top of the feedwater ring and welded to the diameter of the ring. The feedwater ring at the attachment region of the J-nozzle may be susceptible to flow accelerated corrosion (FAC) due to flow turbulence which increases local fluid velocities. If a J-nozzle becomes a loose part, it can cause damage to tubing near the tube sheet. In this paper, the structural stress analysis for a wall thinned feedwater ring and integrity evaluations under assumed loading conditions are carried out in compliance with ASME B&PV SecIII, NB-3200.

• 비파괴검사학회지
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• 제26권1호
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• pp.12-17
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• 2006

원자력발전소의 증기발생기는 수천 개의 매우 얇은 두께의 튜브로 구성되어 있다. 이러한 증기발생기의 튜브는 원자력발전소의 1차 계통과 2차 계통의 압력경계를 유지하는 데에 매우 중요한 역할을 하고 있으며, 고온 고압의 열 수력적 상호작용으로 인한 가혹한 운전조건으로 인하여 손상되기 쉽다. 따라서 증기발생기의 구조적 건전성을 평가하기 위하여 많은 시간과 노력이 투입되고 있다. 와전류 검사 방법이 증기발생기 튜브의 건전성을 평가하기 위한 가장 보편적인 비파괴 방법이지만, 와전류 검사의 특성상 결함의 전체 체적에 의하여 신호의 특성이 나타나게 되어 정확한 결함의 크기를 평가하기에는 한계가 있다. 본 연구에서는 증기발생기 튜브의 결함 검출과 정확한 측정을 위하여 초음파 검사기법의 적용 가능성을 확인하였으며, 연구결과를 증기발생기 튜브 검사에 적용 할 경우 검사결과가 크게 향상 될 것으로 기대된다.