• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2606-2611
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• 2008

The high energy fluid leakage from the high temperature and high differential pressure drop system of NPPs (Nuclear Power Plants) decreases efficiency and consequently leads to considerable economic loss due to less power production. Also, the leakage possibly damages critical parts of components such as valve and trim with the effect of cavitation, flashing, and erosion, etc. and deteriorates its performance. Thus, in this study, we diagnosed the as-is leakage for four (4) main steam drain valves and two (2) steam traps of Yonggwang 1,2 units during normal operation by using multi-measuring technique and observed the occurrence of fine leakage. In the course of measuring fluid leakage, the sign of fine leakage is estimated to be the leakage from orifice. By converting the leakage to energy loss, it is equivalent to the amount of several hundred thousand won per each unit, which supports the basis for the justification of fine leakage.

• Proceedings of the KAIS Fall Conference
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• 2004.06a
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• pp.82-85
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• 2004

EPDM 봉인물질을 가진 SOV의 작동실패 사례 및 원인을 분석해보면 많은 SOV 고장(SOV failures)은 결국 EPDM 봉인물질이 원인인 것으로 나타났다. SOV 고장은 SOV가 그것의 설계 범위를 초과하는 주위 온도(ambient temperature)에 노출되어 있었기 때문에 발생되었다. 그러한 SOV 고장은 증기누설, 주위 온도의 부정확한 평가 등의 원인에 의해 발생되었다. 또한 SOV 고장은 연속적인 coil의 여기시 발생하는 가열(heatup from energization)로 인해 수명이 단축되는 영향이 서비스 수명 기간에 적절하게 포함되지 않았기 때문에 발생되었다 많은 발전소 운용자는 이러한 사실을 알고 있지 못하였으며 결과적으로 서비스 수명 기간이 과도 예측되는 결과가 초래되었다 SOV의 또 다른 연성(Soft) 봉인물질로, 열저항 특성이 EPDM에 비해 약 $50^{\circ}C$ 정도 우수한 Viton 봉인물질에 대한 특성이 제시되었다. 더불어 현재 많은 국외 솔레노이드구동밸브 전문 생산업체들이 Viton 재료를 옵션으로 채택하고 있는 추세이다. 그러나 고온주변에서 지속적으로 노출되어 있는 환경에서 Viton이 EPDM의 문제점을 어느 정도 해결할 수 있는 지의 여부는 현재 결론 내리기 어려울 것으로 판단되며, 향후 이를 채택한 SOV의 운전경험을 계속 주시할 필요가 있을 것으로 판단된다.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.40 no.1
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• pp.81-86
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• 2016

The valves used to control or shut off the flow through a pipeline can be divided into many different types, including gate valves, globe valves, and check valves. Globe valves, in particular, have excellent properties, and because they can easily control the flow under high-pressure conditions, they are generally used in LNG ship and steam pipelines. In this paper, a method for changing the shape of a seat was suggested to solve the valve leakage problem from a structural perspective. In addition, the stress distribution and directional deformation were compared for each model. The suggested models were thus validated, and the optimized seat structure, which includes a self-supporting capability for decreasing the amount of leakage, was determined.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.10
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• pp.1075-1081
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• 2011

A leak-detection method has been developed by measuring the pressure variation between the inner and outer heattransfer tubes of a double-wall tube steam generator. An experiment was carried out to measure the leak rate in the gap between two surfaces pressed with a hydraulic press in order to simulate the phenomena, and a correlation was determined for the leak rate in a micro gap. However, in the correlation, the gap width and friction coefficient were coupled with the surface roughness, which affects the two parameters. The two parameters were separated using a surface-contact model to develop a correlation for the friction coefficient. The correlation was compared with the existing correlations used for crack analysis. Although the applied ranges of Reynolds numbers were different, the developed correlation for Reynolds numbers of 0.1.0.35 showed similar tendencies to existing correlations used for higher Reynolds numbers.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.4
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• pp.295-299
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• 2016

Residual foreign objects at the secondary side (top of the tubesheet and tube support plates) of a steam generator are likely to cause a leak by causing wear in the tube. The extent of wear is significantly affected by the material, shape, and size of the foreign object, and the corrosion properties of the tube. The presence of foreign objects at the top of the tubesheet and tube support plates has been identified using remote visual inspection methods such as the foreign object search and retrieval and eddy current test (ECT). The detection of the residual foreign object at the secondary side of a steam generator has limitations that depend on the material properties and the condition of contact with the tube. In this study, which is vertical and horizontal from the upper tubesheet, the corresponding bobbin ECT signals were collected and analyzed to measure its ability to detect foreign objects.

• Journal of the Korean Institute of Intelligent Systems
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• v.26 no.3
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• pp.246-252
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• 2016

Since thermal power plant (TPP) equipment is operated under very high pressure and temperature, failures of the equipment give rise to severe losses of life and property. To prevent the losses, fault detection method is, therefore, absolutely necessary to identify abnormal operating conditions of the equipment in advance. In this paper, we present Mahalanobis distance (MD) based fault detection method for steam boiler tube in TPP. In the MD-based method, it is supposed that abnormal data samples are far away from normal samples. Using multivariate samples collected from normal target system, mean vector and covariance matrix are calculated and threshold value of MD is decided. In a test phase, after calculating the MDs between the mean vector and test samples, alarm signals occur if the MDs exceed the predefined threshold. To demonstrate the performance, a failure case due to boiler tube leakage in 200MW TPP is employed. The experimental results show that the presented method can perform early detection of boiler tube leakage successfully.

• Journal of the Korean Society of Marine Environment & Safety
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• v.24 no.6
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• pp.825-833
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• 2018

In this study, one calculated the range of damage to the combustion characteristics according to the composition of LNG and the size of leaking holes, and analyzed the damage effect in case of leakage accidents caused by pipe damage in the re-gasification process for the LNG supply system. In order to confirm the combustion characteristics according to LNG composition, there was no significant difference in the result of risk analysis by LNG-producing areas. However, the higher the methane content of the components, the lower the risk of flash fire, hazardous areas of overpressure due to explosion, and thermal radiation damage caused by jet fire. In addition, one investigated the effect of leakage, holes, and ruptures on the risk range and explosions according to the size of the pipe-leakage hole. Also, the influence of overpressure and the range of damage from radiant heat could be predicted. One confirmed the effect of LNG composition and pipe-leakage size on fire and explosion.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.41-53
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• 1994

A SGTR accident postulated at Kori unit 1 is simulated with Mihama unit experience, which occurred on February 1991, to evaluate the capability of plant to cope with the transient. The system design and plant conditions of Kori Unit 1 are much similar with those of Mihama Unit 2. Therefore, special concern has been given to evaluate the sequence and the resulting consequence of the postulated SGTR accident at the Kori unit 1 An analysis is peformed as realistically as possible, with following the EOP of Kori unit 1. The result indicates that the leak through tube break terminates within about forty minutes, and the Kori unit 1 may be sufficient to cope with SGTR accident with same type of sequence. However, the reconsideration may be required for the design of Kori unit 1 which disconnects non-safety AC power from off-site power on SI signal generation. It may be pointed out that the content of EOP for SGTR accident is not enough to require operator's proper judgements. An analysis of SGTR accident tested in the LSTF which simulated the SGTR accident at the Mihama Unit 2 is peformed using the RELAP5/MOD3. The results indicates that the code yields in general good agreement with the test, except the break flowrate at the early stage of the event.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.6
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• pp.1843-1853
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• 1996

The objective of this paper is to evaluate the crack stability of the nuclear main stresm pipes, considering the load reduction effect due to the presence of circumferential throuth-wall crack. Also, the optimization techniques are adoped tosimulate the crack effect on the elbow component of the piuping system. By using a general beam elemetn which contains a discontinuous cross-section, the piping analysis is accomplished to acquire the reduced load. Considering this reduced load, it is feasible for the LBB application in nuclear main stresm pipe. Also, by combining an optimization program and a genaral finite element analysis program, the appropriate dimensions of the simplified beam elemtn which represents the effect of crack in elbow could be successfully determined.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.93-100
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• 2016

Since the sodium-cooled fast reactor is operated in a hostile environment due to the use of liquid sodium as its coolant, advanced techniques for in-service inspection are required to periodically verify the integrity of the reactor. This paper presents the development of in-service inspection techniques for Proto-type Generation IV Sodium-cooled Fast Reactor. First, the 10 m long plate-type ultrasonic waveguide sensor has been developed for in-service inspection of reactor internals, and its feasibility was verified through several under-water and under-sodium experiments. Second, the combined inspection system for in-service inspection of ferromagnetic steam generator tubes has been developed. The remote field eddy current testing and magnetic flux leakage testing can be conducted simultaneously by using the developed inspection system, and the detectability was demonstrated through several damage detection experiments. Finally, the electro-magnetic acoustic transducer which can withstand high temperature and be installable in the remote operated vehicle has been developed for in-service inspection of the reactor vessel, and its detectability was investigated through damage detection experiments.