• Journal of Energy Engineering
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• v.21 no.4
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• pp.362-372
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• 2012

The steam generator tubes of nuclear power plants are pressure boundaries, and if tubes are leaked, the coolant with the radioactive materials was flowed out from the primary system to the secondary system and polluted the plant and the air. Recently most crack defects of tubes are stress corrosion cracks and these defects are located in expansion transition area, sludge pile-up region, and U-bend area. The most effective one of crack initiation factors in expansion transition area and U-bend area is the residual stress. According to the experiences of Korea standard nuclear plants(Optimized Power Reactor-1000), they had the stress corrosion cracks at the tube expansion transition area in early operating stage and especially lots of circumferential cracks were occurred. Therefore in this study, the distributions and conditions of residual stresses by tube expansion methods were compared and the dominant reason of a specific direction was examined.

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.5
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• pp.465-471
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• 2013

A steam generator in nuclear power plant is a heatexchager which is used to convert water into steam from heat produced in a nuclear reactor core, and the steam produced in steam generator is delivered to the turbine to generate electricity. Because of damage to steam generator tubing may impair its ability to adequately perform required safety functions in terms of both structural integrity and leakage integrity, eddy current testing is periodically performed to evaluate the integrity of tubes in steam generator. This assessment is normally performed during a reactor refueling outage. Currently, the eddy current testing for steam generator of nuclear power plant in Korea is performed in accordance with KEPIC & ASME Code requirements, the eddy current testing system is consists of remote data acquisition unit and data analysis program to evaluate the acquired data. The KEPIC & ASME Code require that the electrical properties of remote data acquisition unit, such as total harmonic distortion, input & output impedance, amplifier linearity & stability, phase linearity, bandwidth & demodulation filter response, analog-to-digital conversion, and channel crosstalk shall be measured in accordance with the KEPIC & ASME Code requirements. In this paper, the measurement requirements of electrical properties for eddy current testing instrument described in KEPIC & ASME Code are presented, and the measurement results of newly developed eddy current testing instrument by KHNP(Korea Hydro & Nuclear Power Co., LTD) are presented.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.175.1-175
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.555-560
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• 1997

원자력발전소의 전열관 보수용접기술은 원전시설의 수명연장을 목적으로 개발된 기술이며, 손상된 Inconel 600 전열관 내부에 Inconel 690 재질의 sleeve tube을 삽입한 후 용접을 하는 방법으로 수행된다. 증기발생기 전열관 sleeve 레이저 용접의 기본개념은 방사능지역 외부의 발진기로부터 발진된 레이저빔을 광섬유를 이용하여 장거리(200미터 이상) 전송하여 전열관 내부의 광학계 및 회전장치를 이용하여 전열관을 용접하는 방법이다. 본 연구에서는 펄스형 레이저의 용접 변수인 펄스폭, 반복율, 첨두출력 및 용접속도를 변화시키면서 용입상태를 측정하였고, 용접된 시편에 열처리 여부와 부식실험을 위한 C-ring 시편을 제작하여 caustic test를 위해 auto-clave vessel에서 1,000 시간 실험을 실시하였다.

• Journal of the Korean Society for Nondestructive Testing
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• v.34 no.5
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• pp.356-361
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• 2014

The eddy current testing performance directly affects the results of a steam generator tube integrity assessment because the integrity assessment of defected tubes is conducted based on eddy current testing results. This means that it may not be possible to accurately discriminate between adjacent flaws. This paper presents an investigation on the resolution of rotating pancake coils with multiple cracks and the effects on the structural integrity assessment of steam generator tubes.

• Korean Chemical Engineering Research
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• v.56 no.3
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• pp.388-403
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• 2018

We investigated design requirements of vent lines for Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We developed design requirements of areas of the rupture disks of the steam generator, a diameter of the gas vent line of the sodium dump tank, a diameter of the gas vent line of the water dump tank, a diameter of the water dump line of the steam generator. With the design requirements, we calculated the time to vent fluid inside the steam generator and analyzed the transient pressure behavior, also evaluated the close pressure value of the isolation valve of the water dump line. Our results are expected to be used as basis information to design Sodium-Water Reaction Pressure Relief System of Prototype Generation IV Sodium-Cooled Fast Reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.327-332
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• 1996

대구경유압식스너버의 잠재적 기능상실은 미국원자력규제위원회(USNRC)에서 인정한 주요안 전성 문제중의 하나이다. 본 보고서는 한국표준형원전의 증기발생기와 원자로냉각재펌프에 설치되어있는 대구경유압식스너버 중 단일 스너버의 기능상실에 대하여 지진과 가상분기관파단의 시간이력해석을 수행하여 구조적영향을 보였다. 지진 입력은 SRP 3.7.1 에 따른 가상적 시간이력이며, 분기관파단 입력은 파단전누설기법(Leak Before Break) 적용에 의하여 배제되지 못한 분기관파 단인 증기발생기 주증기배관과 증기발생기 주급수배관의 가상파단을 이용하였다.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.1
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• pp.9-15
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• 2008

A new ECT probe to separate the ECT signal distortion due to PVC (permeability variation clusters) and ordinary defects created in SG tubes has been developed. The hystersis loops of PVC which are extracted from retired SG (steam generator) tubes of Kori-1 NNP were measured. The tensile tests were performed to identify the mechanism of PVC creation. The conditions detecting the PVC created in 56 tubes were investigated using computer simulation, and the signal processing circuits were inserted in the probe for the digital signal transmission. The new Probe can measure and separate the PVC signal which is created in the SG tubes, and also measures the defects in Ni-sleeving part of SG tubes. furthermore the new ECT probe can measure the defects as fast as bobbin probe, and enhance the testing speed as well as reliability of the defect detection of SG tubes.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.211-216
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• 1995

1994년 11월에 나타난 고리 1호기 증기발생기의 전열관 누설에 대한 원인을 조사하기 위하여 인출 전열관의 파손 분석과 슬러지 분석 및 pH 분석 등을 수행하였다. 손상원인은 국부적인 염기도 상승과 부식전위 상승에 따른 2차측 응력부식균열(ODSCC)로 밝혀졌다. 전열관 표면과 접한 관판 상부의 퇴적슬러지 끝단에 형성된 틈새에서 나타나는 비등현상으로 $Na^{+}$ 등의 양이온이 농축하게 되며, Cl$^{-}$ 등의 음이온 증발로 인하여 국부적으로 염기도의 상승현상이 야기되었다. 또한 재 가동시 전열관 표면에 침착된 잔류 구리와 용존산소의 결합으로 부식전위가 상승되었다. 이와 같은 ODSCC 발생환경은 1990년이래 지속적으로 형성된 것으로 판단된다.

• Journal of Energy Engineering
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• v.7 no.1
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• pp.44-56
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• 1998

Separate heat pipe exchanger is considered as the high thermal transportation equipment, because evaporator and condenser are separately positioned in the long distance. Its characteristics are that the working fluid is circulated naturally by the position height of two exchangers. But the operating characteristics are restricted by the temperature of hot and cold fluid, flow pattern and diameter of vapor line, etc. in this study, the vapor pressure and the minimum height of two exchangers are studied about the factors restricting the operating characteristics.