• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.799-800
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• 2004

한전전력연구원에서는 한국표준형 원전 증기발생기인 시스템 80 모델의 튜브시트 상부에 침전된 슬러지를 제거하기 위한 고압 분사식 슬러지 세정장비를 개발하였다. 국내에는 영광 3,4,5,6 호기 및 울진 3,4,5,6 호기 등 총 8 기의 한국표준형 원전이 운전중에 있으나 기 사용하고 있던 증기발생기 세정 장비가 자동화가 미흡하여 작업자의 방사선 쪼임량이 과대하고 장비가 전열관에 충돌하여 전열관이 손상될 가능성이 상존하며 세정 방식이 증기발생기 가장자리인 annulus 에서 중앙의 center stay cylinder 로 향해 슬러지가 증기발생기 중앙 부분에 집중적으로 침적될 가능성이 있는 등 문제점을 해결하기 위하여 개발에 착수하였다. 개발된 증기발생기 세정 장비는 2004 년 4 월 중 울진 3 호기 증기발생기 세정 작업에 성공적으로 활용되었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.561-565
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• 1997

증기발생기 전열관 보수를 위하여 sleeve pipe를 삽입하여 레이저 용접을 하는 과정에서 전열관과 sleeve pipe 의 간격을 최소한으로 줄여 용접 품질을 높이고, 균등하게 하기 위하여 sleeve pipe에 대하여 확관이 수행된다. 확관은 sleeve pipe의 상단부와 하단부에 각각 수행되는데 정확한 확관규격을 유지하기 위하여 컴퓨터로 확관압력의 미분치를 비교분석하여 압력펌프를 제어하였다. 압력신호의 변화가 크고 안정되지 못하여 전후신호와 비교분석하여 안정화시킨 후 미분치를 추출하여 제어함으로써 전열관의 확관이 0.02mm 이내가 되도록 하여 전열관의 과도한 확관을 방지하였다.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.98-103
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• 2009

The nuclear power plants have been safely operated by plugging the steam generator tubes which have the crack indications. Tube rupture events can occur if analysts fail to detect crack indications during in-service inspection. There are various types of crack indication in steam generator tubes and they have been detected by the eddy current test. The integrity assessment should be performed using the crack-sizing results from eddy current data when the crack indication is detected. However, it is not easy to evaluate the crack-depth precisely and consistently due to the complexity of the methods. The current crack-sizing methods were reviewed in this paper and the suitable ones were selected through the laboratory tests. The retired steam generators of Kori Unit 1 were used for this study. The round robin tests by the domestic qualified analysts were carried out and the statistical models were introduced to establish the appropriate depth-sizing techniques. It is expected that the proposed techniques in this study can be utilized in the Steam Generator Management Program.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.437-443
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• 1995

국내, 외에서 원자력발전소의 주요 구성 기기인 증기발생기의 세관 건전성과 관련 설계개선을 위한 연구가 활발히 진행되고 있다[2,3,4,5,6]. 현재 가동중인 발전소에서는 개선된 증기발생기로 교체하고자 하는 검토가 이루어지고 있으며, 설계중인 발전소에서는 중기발생기의 건전성을 향상시키기 위한 노력이 진행되고 있다. 본 논문에서 기존에 조사되고 검토된 자료를 바탕으로 [2] 현재까지 주로 사용되어온 증기발생기의 세관 재질을 인코넬 600 MA(mill annealed)로 사용할 때 40년 수명동안 증기발생기의 건전성을 보장할 수 있는 고온관의 온도를 분석한 결과 적절한 온도가 607$^{\circ}$F(319.4$^{\circ}C$)임을 알았다. 그리고 이 온도를 반영할 때 계통설계에 영향을 주는 설계사항에 대하여 검토하였고, 추가로 인코넬 600 MA보다 고온조건에서 건전성이 양호한 인코넬 690 TT(thermal treatment)를 사용할 때 설계에 미치는 영향도 검토하였다. 이러한 분석결과는 추후 국내 원자력발전소에서 보다 증기발생기의 건전성을 보장하기 위해 설계개선을 하고자 할 때 기초 자료가 되리라 판단한다.

• Journal of Energy Engineering
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• v.22 no.2
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• pp.148-155
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• 2013

The steam generators of OPR-1000 plants have Alloy 600 and Alloy 690 as the tube material and its tube expansion method is the explosive expansion method. According to the experience of these plants, circumferential cracks were largely occurred in steam generator tubes expanded by the explosive expansion method and their locations were the outer surface of tube expansion transition region surrounding with piled-up sludge. But even though tubes have the same conditions, tubes with the hydraulic expansion method shows the prevail trend of axial cracks compared to circumferential cracks. Therefore in this study, in order to identify the difference of such phenomena as above, configurations of tube and tubesheet were modeled and at operating conditions, stress values applied in the tube expansion transition area in accordance with tube expansion methods were calculated by using computational program and the direction and the predominance of cracks were evaluated.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.518-525
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• 2001

As steam generator (SG) tubes have aged, new and subtle flaws have appeared. Most of them start growing from outside the tubes. Since signals from outer diameter (OD) defects are very weak compared to those from inner diameter (ID) defects in the conventional eddy current testing due to skin effect, this paper studies the feasibility of using remote field eddy current (RFEC) technique, which has shown equal sensitivity to ID and OD defects in the ferromagnetic pipe inspection. Finite element modeling studies show that the operating frequency needs to be increased up to a few hundred kHz in order for RFEC effects to occur in the nonmagnetic SG tube. The proper distance between exciter and sensor coils is also found to be about 1.5 OD, which is half the distance used in the ferromagnetic pipe inspection. Defect signals obtained by the designed RFEC probe show equal sensitivity to ID and OD defects and the existence of linear relationship between defect depth and phase signal strength. These results tell us that RFEC inspection is feasible even in nonmagnetic steam generator tubes.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.501-509
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• 2001

For the enhancement of ECT reliability on the primary water stress corrosion cracks of nuclear steam generator tubes, of which the occurrence is on the increase, it is important to comprehend the signal characteristics on crack morphology and to select an appropriate probe type. In this paper, the sizing accuracy and the detectability for the inner wall axial cracks of tubes were quantitatively evaluated using the following specimens: the electric discharge machined notches and the corrosion cracks which were developed on the operating steam generator tubes. The difference of eddy current signal characteristics between pancake and axial coil were also Investigated. The results obtained from this study provide a useful information for more precise evaluation on the inner wall axial tracks oi stram generator tubes.

• Korean Chemical Engineering Research
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• v.57 no.1
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• pp.28-41
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• 2019

We carried out performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We analyzed transient-dynamic behavior of fluids inside the steam generator to vent into a sodium dump tank or a water dump tank when tubes in the steam generator were broken to cause a large-water-leak accident. Accordingly, we preliminarily evaluated design requirements of our system. Our results showed that sodium in the shell side of the steam generator and in Intermediate Heat Transport System was completely vented within 50 s and feed water in the tube side of the steam generator was completely vented within 2.5 s. It was analyzed that pressure of the tube side of the steam generator was higher than pressure of the shell side of the steam generator, which showed that sodium in the shell side did not flow into the tube side. Our results are expected to be used as basis information to performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor.

• Journal of the Korea Institute of Information Security & Cryptology
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• v.31 no.5
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• pp.941-950
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• 2021

Single Sign-On (SSO) service manages user's account passwords from multiple websites so that security in a high level is required. Users who use the SSO service are authenticated through the Identity Provider (IdP) when logging into the website. We present the security requirements that IdP can take in order to minimize the user's risk whose IdP account is compromised. We describe the security threats that arise when the security requirements are not satisfied. Through evaluation, we prove that the attacker's session cannot be canceled even if the user recognizes the attack if the IdP does not satisfy the security requirements.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.237.2-237
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• 2001