• Proceedings of the KSME Conference
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• 2003.04a
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• pp.629-636
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• 2003

The purpose of this study is localization of safety relief valves for Nuclear Service through technical development with overall design, fabrication, inspection, capacity certification test and functional qualification test of safety relief valves in accordance with ASME Section III and KEPIC Code. Safety relief valve is the important equipment used to protect the pressure vessel, the steam generator and the other pressure facility from overpressure by discharging the operating medium when the pressure of system is reaching the design pressure of the system. But we're depending on technology of the other country up to the present time. Because we don‘ have our own technologies, we have been spent the great time and money on installing and repairing safety relief valve at nuclear power plant. Therefore we have to achieve the development of safety relief valves for Nuclear Service with our own technologies.

• Nuclear Engineering and Technology
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• v.32 no.3
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• pp.227-234
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• 2000

It is known that loose parts in the reactor coolant system (RCS) cause serious damage to the systems. This paper is concerned with estimating the mass of a loose part in the steam generator of a nuclear power plant. We developed the mass estimation algorithm based on the Hertz theory in order to estimate the mass of the loose parts and applied the algorithm to the impact test data of YGN3. The mass estimation values were compared with real values in order to verify the algorithm. The result showed that the average error of the mass estimation value is less than 27%.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1118-1124
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• 2023

Experimental study on the fluidelastic instability (FEI) of a curved tube bundle in single phase downward cross flow is investigated for the design qualification and analysis input preparation of helical coiled steam generator tubing. A 6×9 normal square curved tube array with equal and different vertical/horizontal pitch-to-diameter ratio was under-tested up to 6 m/s in term of gap flow velocity to measure the critical velocity for FEI. The critical velocity for FEI was measured at the turning point from the vibration amplitude plot along the gap flow velocity. Our test results were compared with straight tube results and published data in the design guideline. The applicability of the current design guidelines to a curved tube bundle is also assessed. We found that introducing frequency difference in a curved tube array increases the critical velocity for fluidelastic instability.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.14 no.4
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• pp.327-335
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• 2004

This study investigates the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the helical type tubes with various conditions. The wear rate of helical type tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the external pressure on the vibration and fretting-wear characteristics of the tube.

• Proceedings of the KIEE Conference
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• 2003.07d
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• pp.2603-2605
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• 2003

증기발생기(SG : Steam Generator)는 다수의 세관으로 구성되어 원자로에서 발생한 열을 이용하여 발전기 터빈을 구동시키는 원동력인 증기를 생성해 주는 기능을 하는 원자력발전소의 핵심 설비이다. 증기 발생기 세관의 건전성을 확보하기 위해 매주기 계획예방 정비, 즉 가동중검사마다 정기적인 와전류검사를 수행하고, 검사결과에 따라 전열관 보수 등과 같은 제반 조치를 취하고 있다. 현재 검사데이터 DB 구축은 일부 발전소에 개발되어 운영중에 있고, 세관 DB와는 별도로 통계정보만을 관리하는 증기발생기 성능관리시스템이 운영되고 있으며, 또한 각 발전소마다 수질을 계측하여 수화학 성분을 감시하는 수질관리시스템이 운용되고 있다. 이러한 이원화된 DB 및 시스템을 통합하고 연계하여 전 원전의 증기발생기를 종합적으로 관리 할 수 있는 시스템의 필요성이 대두되었다. 따라서 본 논문에서는 현장에 보관되어 있는 모든 세관 검사데이터를 취득하여 대용량 데이터베이스를 설계 및 구축하고 이기종의 분산된 수질관리시스템 DB를 연계하여, 증기발생기의 설계/제작부터 검사결과 Mapping, 추이 분석을 통한 수명평가에 이르는 전 과정을 통합 관리한 수 있는 시스템을 개발하고 그 구현방안을 제시한다.

• 제어로봇시스템학회:학술대회논문집
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• 2001.10a
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• pp.157.4-157
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• 2001

An effective approach to detect and diagnose multiple failures in a dynamic system is proposed for the case where the measurement noise is correlated sequentially in time. It is based on the modified interacting multiple-model (MIMM) estimation algorithm in which a generalized decorrelation process is developed by employing the autoregressive (AR) model for the correlated measurement noise. Numerical example for the nuclear steam generator is provided to illustrate the enhanced performance of the proposed algorithm.

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.550-550
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• 2000

This paper proposed a web-based open architecture monitoring system using the Java language. The proposed system can be implemented in any platform and the status of the process can be monitored in a remote station. The proposed scheme have been applied to a steam generator level controller for a nuclear power plant. The result shows the feasibility of the proposed system.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2003.05a
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• pp.399-402
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• 2003

국내에서는 현재 총 18기의 원자력발전소를 운전하고 있다. 2002년 6월 기준으로 원자력발전설비 용량은 총 14,716㎾로서 전체발전설비 용량의 28%를 차지하며 2001년 원자력 발전량은 58,222 백만㎾h로서 전체의 39%를 차지할 만큼 국내 에너지공급원으로서 큰 기여를 하고 있다. 원자력발전소가 향후에도 지속적으로 주요 에너지공급원으로서 역할을 수행하기 위해서는 무엇보다도 안전성에 대한 신뢰성을 확보해야 할 것이다.(중략)

• Proceedings of the KAIS Fall Conference
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• 2005.05a
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• pp.151-154
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• 2005

In this paper, we present a ultrasonic speckle suppression method for centrifugal-casted stainless steel sample by computer simulations of a flaw enhancement algorithm. Because of their practical importance in welds, the ultrasonic signal obtained from heat-affected zone or welds are investigated for computer simulation. The results for computer simulation present the more enhanced flaw-visibility and speckle suppression than the compared two techniques.

• Proceedings of the KIEE Conference
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• 1997.07b
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• pp.547-549
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• 1997

원자력 발전소용 증기발생기의 수위제어에 기존의 경우 주로 PI 제어기를 이용하는데 반해 본 논문에서는 주급 수유량에 대한 주증기 유량의 변화를 신경망을 이용해 학습함으로서 PI제어기의 각 파라메터를 뉴닝하는 문제를 연구하였다. 적용하여 고찰한 결과 기존의 PI 제어기에 비해 수위변화에 대한 추종성능이 우수함을 나타내었다.