• 한국정보통신설비학회:학술대회논문집
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• 한국정보통신설비학회 2008년도 정보통신설비 학술대회
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• pp.509-512
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• 2008

본 논문에서는 원전 증기발생기세관 진단을 위한 와전류탐상의 전자기 수치해석을 수행하였다. 전자기적 특성을 해석하기 위하여 맥스웰 방정식을 이용하여 지배방정식을 유도하였고, 3차원 전자기 유한요소 프로그램인 OPERA 3D를 이용하여 전자기 수치 해석을 수행하였다. 신호해석을 위해 사용된 프로브의 종류는 배열와전류프로브이며, FBH 결함의 신호를 해석하였다. 결함의 깊이는 세관 두께의 40[%], 60[%] 및 100[%]로 하였다. 시험주파수는 100[kHz], 300[kHz], 400[kHz]를 사용하였고, 각각의 결함 및 시험주파수에 대한 결과를 비교 분석하였다. 본 논문의 결과는 앞으로 배열와전류프로브를 이용하여 원전 증기발생기세관 진단을 할 경우 신호 해석에 도움이 될 것으로 사료된다.

• 한국압력기기공학회 논문집
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• 제5권2호
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• pp.1-6
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• 2009

Nuclear power plants (NPP) using Alloy 600 (Ni 75wt%, Cr 15wt%, Fe 10wt%) as a heat exchanger tube of the steam generator (SG) have experienced various corrosion problems by ageing such as pitting, intergranular attack (IGA) and stress corrosion cracking (SCC). In spite of much effort to reduce the material degradations, SCC is still one of important problems to overcome. Especially lead is known to be one of the most deleterious species in the secondary system that cause SCC of the alloy. Even Alloy 690 (Ni 60wt%, Cr 30wt%, Fe 10wt%) as an alternative of Alloy 600 because of outstanding superiority to SCC is also susceptible to leaded environment. An oxide on SG tubing materials such as Alloy 600 and Alloy 690 is formed and modified expanding to complex sludge throughout hideout return (HOR) of various impurities including Pb. Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that an oxide property such as a passivity of an oxide formed on steam generator tubing materials is deeply related to PbSCC and an inhibitor to hinder oxide modification by lead efficiently can be found. In the present work, the SCC susceptibility obtained by using a slow strain rate test (SSRT) in aqueous solutions with and without lead was discussed in view of the oxide property. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), an energy dispersive x-ray spectroscopy (EDXS), an x-ray photoelectron spectroscopy (XPS) and an electrochemical impedance spectroscopy (EIS).

• 대한기계학회논문집A
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• 제37권5호
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• pp.665-671
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• 2013

제 4 세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX 와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX와 FHX는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다.

• 비파괴검사학회지
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• 제21권5호
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• pp.518-525
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• 2001

증기발생기 전열관이 노후화 됨에 따라 새롭고 판단이 애매한 결함이 발생되기 시작하고 있다. 대부분의 결함은 전열관 외부에서 발생되어 진전된다. 일반 와전류탐상에서는 외부결함으로부터의 신호가 표피효과로 인해 내부결함으로부터의 신호보다 매우 약하기 때문에, 본 논문에서는 자성체 관 검사에서 내부 및 외부결함에 거의 같은 민감도를 보인 바 있는 원격장와전류 탐상을 비자성체인 전열관의 검사에 적용하기 위한 연구를 수행하였다. 유한요소 모델링을 통한 연구결과는 비자성체인 전열관에서 원격장와전류 효과가 나타나려면 탐상주파수가 수백 kHz가 되어야 하며, 여자코일과 센서코일간의 간격은 자성 관 검사시의 절반인 관 외경의 1.5배 정도가 되어야 함을 보였다. 이렇게 설계된 탐촉자를 사용하여 예측한 결함신호들은 이 검사방법이 내부 및 외부결함에 동일하게 민감하며, 위상신호의 세기와 결함깊이간에는 선형적인 비례관계가 존재함을 보여 주었다. 이러한 결과들은 비자성 증기발생기 전열관이라 할지라도 원격장와전류 탐상이 가능함을 말해 주고 있다.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.869-874
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• 2014

It was discovered in a Korean PWR that an extensive number of very short and shallow cracks in the SG tubes were undetectable by eddy current in-service-inspection because of the masking effect of sludge deposits. Axial stress corrosion cracks at the outside diameter of the steam generator tubes near the line contacts with the tube support plates are the major concern among the six identical Korean nuclear power plants having CE-type steam generators with Alloy 600 high temperature mill annealed tubes, HU3&4 and HB3~6. The tubes in HB3&4 have a less susceptible microstructure so that the onset of ODSCC was substantially delayed compared to HU3&4 whose tubes are most susceptible to ODSCC among the six units. The numbers of cracks detected by the eddy current inspection jumped drastically after the steam generators of HB4 were chemically cleaned. The purpose of the chemical cleaning was to mitigate stress corrosion cracking by removing the heavy sludge deposit, since a corrosive environment is formed in the occluded region under the sludge deposit. SGCC also enhances the detection capability of the eddy current inspection at the same time. Measurement of the size of each crack using the motorized rotating pancake coil probe indicated that the cracks in HB4 were shorter and substantially shallower than the cracks in HU3&4. It is believed that the cracks were shorter and shallower because the microstructure of the tubes in HB4 is less susceptible to ODSCC. It was readily understood from the size distribution of the cracks and the quantitative information available on the probability of detection that most cracks in HB4 had been undetected until the steam generators were chemically cleaned.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.7-14
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• 2014

There has been extensive experience associated with the operation of SGs wherein it was believed, based on NDE, that throughwall tube indications were present within the tubesheet. The installation of the SG tubes usually involves the development of a short interference fit, referred to as the tack expansion, at the bottom of the tubesheet. The tack expansion was usually effected by a hard rolling process and thereafter, in most instance, by the expansion of a urethane plug inserted into the tube end and compressed in the axial direction. The rolling process by its very nature is considered to be intensive with regard to metalworking at the inside surface of the tube and would be expected to lead to higher residual surface stresses. Alternate repair criteria(ARC) in the tack expansion area have been developed and applied to nuclear power plants in USA, however domestic nuclear power plants have not applied ARC for tubes in tubeheet area yet. In consideration of the degradation characteristics of tubes in the Steam Generator tubesheet, this paper suggests ARC application for tubes in the steam generator tubesheet of the domestic nuclear power plants in order to assure life time of the steam generator as well as nuclear power plants.

• 조명전기설비학회논문지
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• 제24권1호
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• pp.63-70
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• 2010

본 논문에서는 원자력발전소 증기세관 크기 예측을 위한 신경회로망 구조에 대해서 연구한다. 와류탐상 시험(ECT) 신호로부터 특징을 추출한 후, 결함크기 예측을 위해서 다층퍼셉트론 신경회로망을 이용한다. 결함크기 예측성능을 최대화하기 위해서는 신경회로망의 구조, 특히 은닉층 내의 뉴런의 개수를 신중히 결정하여야 한다. 본 논문에서는, 결함크기 예측을 위한 은닉층 내의 뉴런의 개수를 교차검증을 이용하여 매우 효과적으로 결정할 수 있음을 보인다.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• Nuclear Engineering and Technology
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• 제33권2호
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• pp.145-155
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• 2001

This simulation first allows us to define a transition zone from a bubble to jet mode of the argon out-flow and hereinafter to define a similar area for water-steam leak in the KALIMER SG (Korea Advanced Liquid Metal Reactor Steam Generator) using a water mock-up system, taking into account the KALIMER leak classification and tube bundle design, as a simulation of a real water-steam into sodium leak. in accordance with leak conditions in the KALIMER SG, the transition from bubbling to jetting is studied by means of turbulence regime simulation for argon out-flow through a very small orifice, which has the equivalent diameter of about 0.253 mm. finally the noise generation mechanism is explained from the existing experimental data. We also confirmed the possibility of micro-leak detection from the information of the bubbling mode through simulations and the experiment in this study.

• Corrosion Science and Technology
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• 제10권4호
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• pp.118-124
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• 2011

Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) has been degraded by various corrosion mechanism during the long-term operation. Especially lead (Pb) is known to be one of the most deleterious species in the secondary system causing outer diameter stress corrosion cracking (ODSCC). Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that a property change of the oxide formed on SG tubing materials by lead addition into a solution is closely related to PbSCC. In the present work, the SCC susceptibility was assessed by using a slow strain rate test (SSRT) in caustic solutions with and without lead for Alloy 600 and Alloy 690 (Ni 60 wt%, Cr 30 wt%, Fe 10 wt%) used as an alternative of Alloy 600 because of outstanding superiority to SCC. The results were discussed in view of the oxide property formed on Alloy 600 and Alloy 690. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDXS).