• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.824-829
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• 1998

미국의 개량형 원자력 발전소 개념설계단계에서 중대사고시 사고완화를 위한 전략으로 원자로 압력용기 외부냉각 개념이 제안되었다. 중대사고 진행과정에서 노심용융물이 원자로 압력용기 하부헤드로 재배치 되었을 때 압력용기 외벽을 냉각함으로서 노심용융물을 압력용기 내부에 가두어 두어 격납건물 내로의 유출을 방지하는 방식이다. 이 연구에서는 원자로 압력용기 하부헤드 내의 노심용융물 거동중 자연 순환에 의한 거동을 수치적으로 모의하여 보았다. 연구결과, 정상상태의 온도 및 속도분포는 현상학적으로 적절하게 모의되나 고화와 액화의 경우에는 고유모델의 필요성이 요구되었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.556-561
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• 1995

원자력발전소 중대사고시 격납건물의 건전성을 위협할 수 있는 현상들 중의 하나인 MCCI에 대한 분석을 목적으로 MCCI 관련 실험인 SWISS 및 SURC 실험에 대하여 MELCOR 1.8.2 를 이용하여 계산을 수행하였다. 에어로졸 생성량을 제외한 MCCI 진행과정의 주요 예측대상에 대하여 실험결과와의 비교를 통하여 콘크리트 침식 진행과정 및 침식을, 노심용융물의 온도분포 및 열유속, 반응에 의해 유출되는 각종 가스 생성을, 그리고 노심용융물의 냉각에 따른 각질층 형성 등을 파악하였다. 콘크리트 침식과정 및 노심용융물의 온도 예측은 적절하며 콘크리트 분해에 따른 각종 방출가스는 열수력 조건에 따라 큰 불확실성을 보여주는 것으로 나타났다. 아울러 노심용융물의 냉각에 따른 각질층의 동적 거동해석은 MELCOR 1.8.2의 모델로서는 불가능하였다. 보다 많은 검증계산을 통하여 적절한 해석방법의 도출 및 새로운 모델 제시의 필요성이 있다고 판단된다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.392-397
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• 1996

지반특성에 따라 지진발생시 면진구조물과 비면진 구조물의 응답특성이 어떠한가를 평가하기위해, 1940 El Centro 지진을 입력지진으로 하고, 면진구조물로는 가압경수형 원자로격납건물을 이용하여 수평(NS) 및 수직지진입력에 대한 시간이력해석을 수행하였다. 0.5Hz 수평면진 구조물의 경우 수평방향 가속도응답은 지반특성에 무관하게 거의 변화가 없으며, 또 2Hz 이상에서 비면진구조물의 수평지진가속도응답보다 현저히 낮은 가속도응답을 갖는다. 면진베어링의 수직방향 21Hz 고유진동수는 풍화암의 경우 수직방향 가속도응답에 영향을 주지 않으나. 경암의 경우 원자로지지점에서의 수직방향 가속도응답을 전반적으로 증가시킨다. 비면진 구조물의 경우 지반의 강성이 약할수록 가속도응답이 비교적 큰 폴라크레인위치에서 수평 및 수직방향 가속도응답이 감소되는 것으로 나타났으며, 특히 수직방향의 가속도응답이 크게 감소하는 것으로 나타났다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.95-100
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• 1996

현재 방사선 비상훈련에 사용하는 TID-14844 의 방사선원항은 너무 보수적이고, 각 사고경로별로 방사선원항의 특성을 나타낼 수 없으므로, 원전의 비상사고 발생시 주민의 피폭선량을 최소화하기 위한 발전소 요원의 신속, 정확한 대처능력을 배양하기 위하여 현실적인 방사선원항 평가자료의 필요성이 대두되어 왔다. 본 연구에서는 보수성을 배제한 최적 분석기법을 이용하여 선정된 사고경로에 대해 MAAP 전산코드로 사고진행 및 방사선원항을 분석하였고, 격납건물내 방사선계측기의 예측치를 평가할 수 있는 방법론을 개발하였으며, 이를 통해 사고경로별 안전변수 및 방사선 계측기 등에서의 사고 진행에 따른 예측치 등을 계산함으로써 효과적인 비상대책 수립을 위한 실질적인 방사선원항 데이타 베이스를 구축하였다.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2014.11a
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• pp.205-206
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• 2014

It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. The purpose of this study attempts to evaluate the single-stage workability of the system given a change in the height of the setting of RCB exterior wall formwork to be used in nuclear power plant construction. As a result of this study, it is possible height of 3.5m~4m uses formwork when analyzing the construction period and material costs an increase in formwork by concrete lateral pressure, to ensure the workability of the RCB exterior wall formwork. Through this study, I want to provide as basic data for the improvement of workability and RCB shortening the construction period.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.1-9
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• 2020

Large-scale industrial facility structures continue to deteriorate due to the effects of operating and environmental conditions. The problems of these industrial facilities are potentially causing economic losses, environmental pollution, casualties, and national losses. Accordingly, in order to prevent disaster accidents of large structures in advance, the necessity of diagnosing structures using non-destructive inspection techniques is being highlighted. The defect occurrence, location and defect type of the structure are important parameters for predicting the remaining life of the structure, so continuous defect observation is very important. Recently, many researchers have been actively researching real-time monitoring technology to solve these problems. Structure Health Monitoring Inspection is a technology that can identify and respond to the occurrence of defects in real time, but there is a limit to check the degree of defects and the direction of growth of defects. In order to compensate for the shortcomings of these technologies, the importance of defect imaging techniques is emerging, and in order to find defects in large structures, a method of inspecting a wide range using guided ultrasonic is effective. The work presented here introduces a calculation for the shape factor for evaluation of the damaged area, as well as a variable β parameter technique to correct a damaged shape. Also, we perform research in modeling simulation and an experiment for comparison with a suggested inspection method and verify its validity. The curved structure image obtained by the advanced RAPID algorithm showed a good match between the defect area and the shape.

• Journal of the Earthquake Engineering Society of Korea
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• v.10 no.6 s.52
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• pp.137-145
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• 2006

Recently, the interests in structural monitoring of civil infrastructures are increased. Especially, as the civil infrastructures such as bridges, tunnels and buildings become large-scale, it is necessary to monitor and maintain the safety state of the structures, which requires smart systems that can supply long-term monitoring during the service time of the structures. In this paper, we investigated the possibilities of fiber optic sensor application to the various structures. We investigate the possibility of using fiber optic Bragg grating sensors to joint structure. The sensors show good response to the structural behavior of the joint while electric gauges lack of sensitivity, durability and long term stability for continuous monitoring. We also apply fiber optic structural monitoring to the composite repaired concrete beam structure. Peel-out effects is detected with optical fiber Bragg grating sensors and the strain difference between main structure and repaired carbon sheets is observed when they separate each other. The real field test was performed to verify the behaviors of fiber Bragg grating sensors attached to the containment structure in Uljin nuclear power plant in Korea as a part of structural integrity test which demonstrates that the structural response of the non-prototype primary containment structures. The optical fiber Bragg grating sensor smart system which is the probable means for long term assessments can be applicable to monitoring of structural members in various civil infrastructures.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2005.04a
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• pp.183-189
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• 2005

Nuclear containment structure is the last barrier for being secure from any nuclear power plant accident. Even though the safety requirements of nuclear power plant have been focused on removing accidental situations, nuclear containment structure must reserve the sufficient resisting capacity to any accident because it works as the last barrier. The acceptable nuclear containment structure makes possible to limit the effect of internal accidents and to avoid radioactive release. In this study, to conduct the numerical analysis for the structural safety of a containment structure, loss of coolant accident (LOCA) is considered as the basic accidental load, and Wolsong containment structure is considered as a target structure. The CANDU containment structure, such as Wolsong containment structure, is a prestressed concrete shell structure which has dome and is reinforced with bonded tendons. The evaluation of ultimate pressure capacity was conducted by nonlinear analysis of a prestressed concrete containment structure.

• 유체기계공업학회:학술대회논문집
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• 2001.11a
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• pp.404-410
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• 2001

A new thermal type fluid level transmitter was designed and tested at the HITROL R&D institute. The relation of heat transfer and electric resistance was adopted as an operation principle. The length of a fabricated level transmitter was two meters and a water under normal temperature was used as a working fluid for the experiment Finally, the new product could have a high precision, acceptable accuracy and reasonable response time. Foreign-made level transmitter of this type is under in use for measuring water level of containment building in nuclear power plants so far. It is expected that new product will substitute it.

• Applied Chemistry for Engineering
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• v.16 no.6
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• pp.809-814
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• 2005

In this work, the thermal properties of epoxy coating system on the liner plate in the containment structure of nuclear power plants had been examined by irradiation and design basis accident (DBA) conditions. The effect of immersion in hot water on adhesion strength of the coating system had been also studied. The glass transition temperature ($T_g$) and thermal stability of ET-5290/carbon steel A 32 epoxy coating systems were measured by DSC and TGA analyses, respectively. Contact angle measurements were used to determine the effect of immersion on the surface energetics of epoxy coating system, with a viewpoint of surface free energy. Adhesion tests were also executed to evaluate the adhesion strength at interfaces between carbon steel plate and epoxy resins. As a result, it was found that the irradiation led to an improvement of internal crosslinked structure in cured epoxy systems, resulting in significantly increasing the thermal stability, as well as the $T_g$. Also, the immersion in hot water made a role in the post-curing of epoxy resins and increased the mechanical interlocking of the network system, resulting in increasing the adhesion strength of the epoxy coating system.