• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.6
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• pp.511-518
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• 2013

Steam Generator (SG) tube, as a barrier isolating the primary coolant system from the secondary side of nuclear power plants (NPP), must maintain the structural integrity for the public safety and their efficient power generation. So, SG tubes are subject to the periodic examination and the repairs if needed so that any defective tubes are not in service. Recently, corrosion related degradations were detected in the tubes of the domestic OPR-1000 NPP, as a form of axially oriented outer diameter stress corrosion cracking (ODSCC). According to the studies on the factors causing the heat fouling as well as developing corrosion cracking, densely scaled deposits on the secondary side of the SG tubes are mainly known to be problematic causing the adverse impacts against the soundness of the SG tubes [1]. Therefore, the processes of various cleaning methods efficiently to dissolve and remove the deposits have been applied as well as it is imperative to maintain the structural integrity of the tubes after exposing to the cleaning agent. So qualification test (QT) should be carried out to assess the perfection of the chemical cleaning and QT is to apply the processes and to do ECT. In this paper, the chemical cleaning processes to dissolve and remove the scaled deposits are introduced and results of ECT on the artificial crack specimens to determine the effectiveness of those processes are represented.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.3
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• pp.432-439
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• 2003

Inconel 690 for nuclear steam generator tube has more Chromium than the conventionally used Inconel 600 in order to increase the corrosion resistance. To evaluate the tribological characteristics of Inconel 690 under fretting condition the fretting tests were carried out in air and elevated temperature water. Fretting tests of the cross-cylinder type were done under various vibrating amplitudes and applied normal loads in order to measure the friction forces and wear volumes. From the results of fretting wear tests. the wear of Inconel 690 can be predictable using the work rate model. The amounts of friction forces were proportional to relative movement between two fretting surfaces. The friction coefficients were decreased as increasing the normal loads and deceasing the vibrating amplitudes. Depending on fretting environment, distinctively different wear mechanisms and often drastically different wear rates can occur It was found that the fretting wearfactors in air and water at 2$0^{\circ}C$, 5$0^{\circ}C$, and 8$0^{\circ}C$ were 7.38 $\times$ $10^{-13}$$Pa^{-1}$, 2.12 $\times$$10^{-13}$$Pa^{-1}$, 3.34$\times$$10^{-13}$$Pa^{-1}$and 5.21$\times$$10^{-13}$$Pa^{-1}$, respectively flexibility to model response data with multiple local extreme. In this study, metamodeling techniques are adopted to carry out the shape optimization of a funnel of Cathode Ray Tube, which finds the shape minimizing the local maximum principal stress. Optimum designs using two metamodels are compared and proper metamodel is recommended based on this research.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.204-211
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• 2015

Probabilistic safety assessment (PSA) has had a significant role in quantitative decision-making by finding design and operational vulnerabilities and evaluating cost-benefit in improving such weak points. In particular, it has been widely used as the core methodology for risk-informed applications (RIAs). Even though the nature of PSA seeks realistic results, there are still "conservative" aspects. One of the sources for the conservatism is the assumptions of safety analysis and the estimation of failure frequency. Surveillance, diagnosis, and prognosis (SDP), utilizing massive databases and information technology, is worth highlighting in terms of its capability for alleviating the conservatism in conventional PSA. This article provides enabling techniques to solidify a method to provide time- and condition-dependent risks by integrating a conventional PSA model with condition monitoring and prognostics techniques. We will discuss how to integrate the results with frequency of initiating events (IEs) and probability of basic events (BEs). Two illustrative examples will be introduced: (1) how the failure probability of a passive system can be evaluated under different plant conditions and (2) how the IE frequency for a steam generator tube rupture (SGTR) can be updated in terms of operating time. We expect that the proposed model can take a role of annunciator to show the variation of core damage frequency (CDF) depending on operational conditions.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.53-66
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• 1995

Since the current licensed system codes for Non-LOCA safety analysis are applicable only for a specific type PWR, it is necessary to develope a new system analysis code applicable for all apes of PWRs. As a R&D program, KAERI is developing TASS code as an interactive and faster-than-real-time code for the NSSS transient simulation of both CE and Westinghouse plane. It is flexible tool for PWR analysis which gives the user complete control over the simulation through convenient input and output options. In this paper the code applicability to Westinghouse ape plants was verified by comparing the TASS prediction to plant data of loss of AC power and loss of load transients, and comparing to the prediction of RELAP5/MOD3 for feedline break, locked rotor, steam generator tube rupture and steam line break accidents.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.31 no.7 s.262
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• pp.784-791
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• 2007

Fretting, which is a special type of wear, is defined as small amplitude tangential oscillation along the contacting interface between two materials. In nuclear power plants, fretting wear caused by flow induced vibration (FIV) can make a serious problem in a U-tube bundle in steam generator. In this study, substructure method is developed and is verified the feasibility for the finite element model of fretting wear problems. This method is applied to the two-dimensional finite element analyses, which simulate the contact behavior of actual tube to support. For these examples, computing time can be reduced up to 1/5 in comparisons with conventional finite element analyses.

• Journal of the Korean Society for Precision Engineering
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• v.17 no.4
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• pp.85-96
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• 2000

An Eggcrate assembly is an important part to hold and support 16,000 tubes containing hot and contaminated water in the steam generator of nuclear power plant. As a great number of tubes should be inserted into the eggcrate assembly, the dimensions of each eggcrate grid are one of the critical factors to determine the availability of tube insertion. in this paper. we propose a machine vision algorithm for measuring the inner-circle diameter of each eggcrate grid whose shape is not exact quadrangular. The overall procedure of the algorithm is composed of camera calibration, eggcrate image preprocessing, grid height adjustment, and inner-circle diameter estimation. The algorithm is tested on real specimens and the results show that the algorithm works fairly well.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 1995.10a
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• pp.125-128
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• 1995

The remote sleeve repair-welding technology using the pulsed Nd:YAG laser for increasing the lifetime of the steam generator tube in the nuclear power plant has been developed. The laser welding has many advantages on deep penetration depth and narrow heat affect zone(HAZ). The inconel 600 tube and inconel 690 sleeve used high temperature and high pressure service have been welded as round lap welds. It is found that the relation between the connection width and welding parameters. It is found that the shear strength in proportion to the connection width by conducting tensile-shear tests.

• Nuclear industry
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• v.30 no.3
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• pp.52-56
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• 2010

증기발생기 2차측 상태는 발전소 운영에 있어 중요한 역할을 수행한다. 기기, 배관 및 열교환기에서 발생되는 침적물은 증기발생기 튜브의 부식 원인이 될 수 있으며, 튜브와 튜브 지지판 사이의 공간을 차단한다. 2차측 침적물에 의한 튜브 파손으로 인해 일부 발전소에서는 강제적으로 발전을 정지하는 사례가 발생하였다. 또한 튜브 지지판의 침적물 축적으로 인해 정상 운전 동안 전력 생산을 감소하게 되는 결과를 초래한 발전소도 있었다. 따라서 증기발생기 2차측 상태 감시와 더불어 증기발생기 부품의 청결 유지는 필수 항목이라 할 수 있다. 웨스팅하우스에서는 증기발생기를 초기 제작 상태로 복구하고 2차측 침적물을 제거하기 위해 EPRI SGOG 증기발생기 화학 세정을 수 년간 이용하고 있다. 본고는 35개 이상의 발전소에서 성공적으로 이용하고 있는 화학 세정 프로세스 개요 및 프로세스를 적용하면서 취득한 경험을 요약한 것이다.

• Proceedings of the Korean Institute of Intelligent Systems Conference
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• 1997.10a
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• pp.441-444
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• 1997

증기 발생기 수위 제어기의 성능 향상은 발전소의 정기 횟수를 줄여 발전소 신뢰도 및 가동률을 향상시키고 또한 기타 여러 부품의 수명에도 영향을 주어 경제적으로 보다 효율적인 발전소 운영에 기여한다. 이러한 수위 제어의 발전을 위해서 본 연구에서는 E. Irvingd의 모델을 사용하였다. E. Irving이 모델이 단순화한 관계로 단점을 가지고는 있으나 프로그램화가 편리하고, 또한 증기 발생기의 특성을 잘 표현하기 때문에 이용하였다. 먼저 시스템의 출력, 즉 증기 발생기의 수위를 안정화시키기 위하여 퍼지 제어기를 Case by Case로 선정하여 제어를 하였으며, 그 다음으로 시스템의 두 입력, 증기량과 퍼지 제어기에서 선택되어진 급수 유량, 그리고 전 단계의 출력인 증기 발생기의 수위를 입력으로 하는 신경 회로망을 이용하여 시스템을 규명하였다.

• Proceedings of the KIEE Conference
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• 2001.11c
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• pp.11-14
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• 2001

원자력 발전소의 증기 발생기는 증기량과 급수량에 대한 비 최소위상 특성과 비선형성, 그리고 입력 제한 특성을 가지고 있다. 이러한 특성들은 증기 발생기의 효과적인 수위 제어에 어려움을 주고 있다. 본 논문에서는 게인 스케줄링 기법과 변형된 슬라이딩 모드 제어 기법을 이용한 원전 증기 발생기 제어기를 제안한다. 또한 앞먹임 구조를 가진 PI 제어기를 설계하여 저출력 영역에서 제안된 슬라이딩 모드 제어기와 성능을 비교한다. 모의 실험 결과 제안된 슬라이딩 모드 제어기가 최대 수위, 최소 수위, 그리고 안정화 시간 면에서 개선된 성능을 보였다.