• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.407-412
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• 1996

본 논문은 중수로 내의 중수증기 회수 계통을 효율적, 경제적으로 운전하기 위하여 현 계통의 문제점을 분석하고, 이를 해결할 수 있는 제어알고리즘에 대해 연구하였다. 또한 PLC로 시제품을 개발하고 이를 기존 제어시스템과 병행설치하여 운전함으로써, 원자로 건물내의 습분 정도에 따라 변하는 DRYER 베드의 흡착능력에 연계된 운전 및 휴지상태(Idle running)를 없애는 최적 운전 방안을 도출하였다.

• Journal of Energy Engineering
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• v.27 no.4
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• pp.27-35
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• 2018

The passive containment cooling system (PCCS) has been designed to remove the released decay heat during the accident by means of the condensation heat transfer phenomenon to guarantee the safety of the nuclear power plant. The heat removal performance of the PCCS is mainly governed by the condensation heat transfer of the steam-air mixture. In this study, the heat removal performance of the PCCS was evaluated by using the MARS-KS code with a new empirical correlation for steam condensation in the presence of a noncondensable gas. A new empirical correlation implemented into the MARS-KS code was developed as a function of parameters that affect the condensation heat transfer coefficient, such as the pressure, the wall subcooling, the noncondensable gas mass fraction and the aspect ratio of the condenser tube. The empirical correlation was applied to the MARS-KS code to replace the default Colburn-Hougen model. The various thermal-hydraulic parameters during the operation of the PCCS follonwing a large-break loss-of-coolant-accident were analyzed. The transient pressure behavior inside the containment from the MARS-KS with the empirical correlation was compared with calculated with the Colburn-Hougen model.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.349.2-349
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• 1999

• JOURNAL OF ELECTRICAL WORLD
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• s.443
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• pp.29-34
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• 2013

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.108-108
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• 2002

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.10a
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• pp.457-458
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• 2012

• The magazine of the Korean Society for Advanced Composite Structures
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• v.5 no.1
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• pp.19-25
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• 2014

본 연구에서는 국내외 항공기 충돌에 대한 원자력발전소의 안전성 평가 및 규제 현황과 연구용원자로의 안전성 평가 및 규제 현황을 살펴보았다. 이러한 현황과 평가와 관련하여 연구용원자로에 적용할 수 있는 항공기 충돌 안전성 평가 및 대비설계 기준을 원자력발전소에 적용되는 기준을 기반으로 정리하였다. 본 연구를 바탕으로 후속되는 연구에서는 연구용원자로에 대한 실질적인 항공기 충돌 안전성 평가 및 대비설계 기준을 도출할 수 있으며, 평가 및 대비설계 방법을 상세히 정립할 수 있을 것이다. 결과적으로, 이를 바탕으로는 항공기 충돌에 대비한 연구용원자로 건물의 예비개념설계 모델을 개발할 수 있을 것으로 예상된다.

• Journal of Energy Engineering
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• v.23 no.1
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• pp.90-94
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• 2014

Environmental qualification (EQ) for safety-related equipment is required to ensure that those equipment will perform their required function even under the harsh environment conditions arising from design basis accident in the nuclear power plant. As a part of EQ program, the room temperature analysis in case of a loss of Heating, Ventilation, and Air Conditioning(HVAC) system was carried out to ensure the operability of the safety-related equipment of a nuclear power plant randomly chosen among the Korean nuclear power plants. In this paper, this analysis was performed in the conservative perspective using GOTHIC code. The room temperature analysis includes selecting the rooms in which the safety related equipment are located but not supported by safety related HVAC and determining the temperature of the selected rooms. Target rooms for the analysis consist of W229/W237 (Aux. feedwater pump room), W232 (Aux. feedwater tank room) and W230 (Equipment passageway). The results showed the temperature range from $43^{\circ}C$ to $83^{\circ}C$, in 72 hours after a loss of HVAC. Those values are far below of generic EQ temperature($171^{\circ}C$). Therefore, it is satisfied with EQ requirement of temperature limits on safety related equipment.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.5
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• pp.535-542
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• 2015

A reactor containment acts as a final barrier to prevent leakage of radioactive material due to the possible reactor accidents into external environment. Because of the functional importance of the containment building, the SIT(Structural Integrity Test) for containments shall be performed to evaluate the structural acceptability and demonstrate the quality of construction. In this paper, numerical analyses are presented, which simulate the results obtained from the SIT for a prestressed concrete(PSC) structure. A sophisticate structural analysis model is developed to simulate the structural behavior during the SIT properly based on various preliminary analysis results considering contact condition among structural elements. From the comparison of the analysis and test results based on the acceptance criteria of ASME CC-6000, it can be concluded that the construction quality of the containment has been well maintained and the acceptable performance of new design features has been verified.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.5
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• pp.523-533
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• 2015

A reactor containment acts as a final barrier to prevent leakage of radioactive material due to the possible reactor accidents into external environment. Because of the functional importance of the containment building, the SIT(Structural Integrity Test) for containments shall be performed to evaluate the structural acceptability and demonstrate the quality of construction. An initial numerical analysis was performed to simulate the results obtained from the SIT for a prestressed concrete(PSC) structure. But the analysis results by the initial model expected smaller displacements than the measured ones by 30% at some locations. Accordingly, the research and development to improve the initial model to corelate the measured results of the SIT more properly have been performed. In this paper, the effects of the loss of concrete due to duct for tendons and the contact of duct and tendons in un-bonded tendon system are mainly evaluated based on the preliminary analysis results. In addition, the importances of the proper definition of mesh connectivity among structural elements of concrete, liner plates, rebars and tendons are discussed.