• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.267.2-267
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• 1999

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.268-268
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• 1999

• 한국전기전자재료학회논문지
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• 제20권3호
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• pp.273-280
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• 2007

In this paper, the ZnO surge arrester performance of power distribution class has been studied under different manufacturing conditions such as housing materials(polymeric, porcelain), interface sealants and one-body molding type. In the recent years, the polymeric ZnO surge arresters have been developed and put into operations based on their excellent characteristics. For polymeric surge arresters, the inner gas volume is extremely small, especially in solid insulation polymeric arresters there are not any gas volume inside arresters in the structure due to polymeric materials are filled into the internal gas volume. The sealing integrity is related to safe operation of surge arrester, the prime failure reason of porcelain housed arresters is moisture ingress. In this paper, the sealing integrity of polymeric surge arresters is investigated with moisture multi-aging test and ingress test. The evaluation techniques are used to inspect the sealing integrity of polymeric arresters, including leakage current, surface temperature, reference voltage and dissipation factor.

• 대한기계학회논문집A
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• 제25권7호
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• pp.1031-1038
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• 2001

In order to verify the analytical methods predicting failure behavior of cracked piping, full-scale pipe tests are crucial in nuclear power plant piping. For this reason, series of international test programs have been conducted. However, full-scale pipe tests require expensive testing equipment and long period of testing time. The objective of this paper is to develop a test system which can economically simulate the full-scale pipe test regarding the integrity evaluation. This system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system was developed for the integrity evaluation of nuclear piping based on the methodology of hardware-in-the-loop (HiL) simulation. Using this simulator, the piping integrity can be evaluated based on the elastic-plastic behavior of full-scale pipe, and the high cost full-scale pipe test may be replaced with this economical system.

• 한국안전학회지
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• 제28권6호
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• pp.42-48
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• 2013

Forensic engineering is the application of engineering principles covering the investigation of constructed facilities and systems that fail to perform as intended, causing personal injury or damage to property, environmental, economy etc. In the year 2012, two collapsed accidents of the large scaffolding system in national thermal power station occurred one after another, causing many casualties. In this study, we had performed to investigate the collapsed accident of scaffolding system occurred in the a thermal power station of two accidents. First, the investigation about the collapsed accidents site had performed to understand collapsed state and structures of the scaffolding system. Second, reviewing the materials concerning about the applied weight on the scaffolding system had performed. The applied weight is sum of the weights of the 15 workers, additional materials for coating work and dispersed and loaded shot ball on the foothold etc. the applied weight that calculated exceed more three times than the safe working load. Third, we had confirmed the install state of the materials of the scaffolding system by reviewing the quantity of the materials on the manual and the real system. Last, structural analysis had performed to evaluate structural integrity of the scaffolding system using Ansys. Through a series of this processes, the definite accidents causes of the collapsed scaffolding system revealed. Through these studies, the collapse accident that may occur in the scaffolding system in thermal power station can be minimized by performing specialized and systematic investigation on the accidents in terms of Forensic engineering.

• Corrosion Science and Technology
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• 제3권5호
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• pp.216-221
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• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1998년도 하계학술대회 논문집 B
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• pp.813-815
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• 1998

Engineered Safety Features-Component Control Systems(ESF-CCS) are those I&C systems that control safety equipment used to maintain the integrity of reactor coolant pressure boundary. This paper illustrates distinctive features and improved design concepts of Korea Next Generation Reactor(KNGR) based on the experience obtained through prototyping of ESF-CCS.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2006년 추계학술발표대회 개요집
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• pp.26-28
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• 2006

When materials such as metal are into the secondary side of steam generator, they, so called foreign objects, may have influences on the integrity of the steam generator tubes. They cause the tube wear due to the relative motion between the tubes and foreign objects and the tube impact due to flow. The best way to avoid the effects is to remove all the foreign objects. However, it is not easy to remove the foreign materials thoroughly due to their condition such as the location. Considering the wear and impact by the foreign materials, KEPRI(Korea Electric Power Research Institute) developed the methodology to evaluate the foreign materials analytically. This methodology was described with a computer program in order to obtain the fast results. The program informs whether the tubes have the structural integrity when the foreign material strikes the tubes. Moreover, this gives us the remaining life of the steam generator tubes. In this paper, the program, which evaluates the effects of the foreign objects in the secondary side of steam generator, is introduced.

• 한국지진공학회논문집
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• 제22권2호
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• pp.87-94
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• 2018

In this study, seismic performance of bushings and their connection parts was analyzed by performing shaking table tests for various types of bushings widely used as auxiliary equipment of main transformers in domestic substations. As a result of the seismic tests of five types of 154 kV bushings according to the manufacturers, all the bushings secured the structural integrity even at the acceleration of 1.4 g and it was found that leakage of insulating oil didn't occur. Also, the average acceleration amplification rate at the upper part of the bushings was about 2.5 to 3.0 times higher than the lower one. On the other hand, when a representative 345 kV bushing was subjected to the seismic test, the structural integrity was secured even at 1.0 g acceleration similar to the design earthquake load level, but in this test, leakage of insulating oil occurred. However, when a stiffener restricting the connection of the bushing is installed in the same 345 kV bushing, the displacement of the bushing connection is controlled and the stiffener prevent the oil from leaking even at the acceleration of the designed seismic level.

• 한국안전학회지
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• 제22권1호
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• pp.13-18
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• 2007

Structural integrity assessment of thin-walled pipes and pipe items has become one of the major issues in the nuclear power plant. ASME Section XI Code Case N-597-2 provides a criterion for acceptance of the pipes. But the code case has several limitations for application and sometimes gives too conservative or non-conservative results. So it is necessary to understand fully the technical bases of the code case. In the code case N-597, the allowable local thicknesses of thinned straight pipes are given for three different cases. Because of the different technical base, each case gives different thickness values and sometimes gives contradictory values. In this paper attempts were made in order to propose a unified rule for the allowable local thickness and in order to remove or relax the restrictions on the application of the code case. For this purpose elastic stress analyses were made using the finite element method and the stress results were examined. Based on the obtained bending stress results, a very simple procedure was proposed to obtain the consistent allowable local thickness for the thinned straight pipes.