• Proceedings of the KSME Conference
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• 2003.04a
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• pp.49-54
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• 2003

Most of the operating pressurized water reactors (PWRs)has chosen Inconel 600 as steam generator tubing. The long-term operation of steam generators showed that the use of this material induced localized corrosion damages. The current trend is using Inconel 690 as a tube material for the replacement steam generators. Based on the current trend, we have chosen Inconel 690 for the advanced Power Reactor 1400 (APR1400) steam generator tube material. In this paper, we examined the technical consideration in this modification: the effect of chemical composition, thermal conductivity, corrosion resistance and wear characteristics

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2004.05a
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• pp.115-118
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• 2004

In this study, amplitudes of turbulence excitation are obtained for selected tubes inside the KSNP SG and their normal work-rates are investigated to estimate the magnitude of fretting-wear. From the results of numerical calculation, row 40&41 tubes show the maximum work-rates. Up to this row number, the work-rates inside the row 41 have much larger values than those of outside tubes. This phenomenon reveals the particular central one which has larger normal work-rate than that of outside zone. It turns out that both of the higher local mode at the U-bend region and the larger value of effective mass in the central region Increase the normal work-rate enormously.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.1-8
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• 2010

Nuclear steam generator tubes have experienced wear degradation at tube support structure. Morphology of wear scar was analyzed by using eddy current signal. A burst test facility for steam generator tubes was established and tubes with 3 types of defects were tested. The burst test results show that the depth of wear scar is the main factor influencing the burst pressure of tubes, meanwhile, both the longitudinal length and the angle also have effect on the burst pressure. Based on test results, the burst pressure equation for wear degradation was proposed.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.16 no.12
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• pp.1107-1116
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• 2004

In this paper, the effects of diverse parameters on the operation of loop heat pipe (LHP), such as particle diameter of sintered porous wick, wick porosity, vapor line diameter, thickness of wick and heating capacity were investigated by a theoretical analysis. A LHP has a wick only in its evaporator for the circulation of working fluid, and utilizes a porous wick structure of which pore size is very small to obtain a large capillary force. The working fluid is water and the material of sintered porous wick is copper. For these different parameters, capillary pressure, pressure drop in wick, pressure drops and temperature distribution were analyzed by a theoretical design method of LHP.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11b
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• pp.515-520
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• 1996

원자력 발전소 증기발생기의 1차측 및 2차측 응력부식균열에 대한 온도감소 효과를 고리 1호기의 현장 데이터를 근간으로 분석하였다. 고리 1호기의 경우 출력을 100%에서 85%로 감소시키므로써, 고온관 온도는 320.5$^{\circ}C$에서 313.5$^{\circ}C$로 7$^{\circ}C$ 감소하였으며, 이와 같은 온도감소 효과로 PWSCC 손상률은 약 40%, ODSCC 손상률은 약 33% 감소하는 것으로 산출되었다. PWSCC의 경우 Weibull 기울기는 b = 5.6 에서 b : 3.8로 감소한 것으로 나타났다. PWSCC의 억제방안으로는 출력감발에 의한 온도감소가 가장 효과적이지만, ODSCC의 경우에는 틈새 분위기의 변환이 큰 역할을 하는 것으로 나타났다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.449-454
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• 1996

원자력발전소 증기발생기 전열관의 레이저 슬리브 용접시, 레이저 전송 및 용접상태의 광학적 감시방법을 개발하였다. 전열관 레이저용접은 용접 중의 레이저 출력, 레이저 전송 광학계의 파손여부, 광학 정렬상태 등을 정확히 감시하며 수행하여야 하지만, 작업공간의 협소함과 방사능 공간이라는 어려움 때문에 적절한 감시방법이 없었다. 본 연구에서는 레이저 빔 전송을 위한 광섬유 광학계를 그대로 이용하여, 용접시 발생되는 radiation과 용접 표면에서 반사되는 Nd:YAG 레이저 빔을 측정하여 레이저 및 광학계 상태를 실시간 감시할 수 있는 기술을 실험적으로 확인하였다. 실험은 Inconel plate를 시편으로 이루어졌으며, 레이저 펄스길이, 레이저 반복률에 따른 감시 조건과 초점확인 기능에 대해서도 논의하였다.

• The Magazine of the Society of Air-Conditioning and Refrigerating Engineers of Korea
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• v.28 no.5
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• pp.389-395
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• 1999

냉동공조설비, 발전설비, 화학플랜트설비 등에 사용되는 응축기는 주로 증기가 관의 외부에서 응축을 하고 냉각수가 관 내부로 흐르는 쉘-튜브(shell and tube)형 태를 취하고 있다. 초기투자비용 및 운전비용을 줄이기 위해서는 응축기의 열교환 성능을 향상시키는 일이 필수적이며 이를 위해 코팅 표면(coated surfaces), 거친 표면(rough surfaces), 코일 튜브(coiled tubes), 선회 흐름장치(swirl flow), 전열면적을 넓힌 낮은 핀관과 3차원 형상을 갖는 열전달 촉진관의 사용이 제시되고 있다.

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.1
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• pp.12-17
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• 2006

A steam generator of nuclear power plant has thousands of thin tubes. These tubes play an important role in maintaining the pressure boundary between the primary and secondary side of nuclear power plant. The steam generator tube is easy to be damaged because of the severe operating conditions such as the high temperature and pressure. Therefore, tremendous efforts are made to assess the structural integrity of the steam generator tubes. The eddy current test is the most popular non-destructive test to assess the integrity of the tubes. However, the eddy current test has the limitation to size the flaw accurately because the eddy current signal behavior depends on the total volume of flaw. This paper shows the possibility that the ultrasonic test method can be applied to detect the flaws in the steam generator tubes and to measure them quantitatively. From the test results, it is expected that if the ultrasonic test is put to practical use in the steam generator tube inspection, the inspection results will be improved.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.344.1-344
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.281.2-281
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• 2001