• Journal of Energy Engineering
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• v.25 no.2
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• pp.1-20
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• 2016

A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.

• 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.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 1997.11a
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• pp.229-234
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• 1997

밀폐공간내에서 일어나는 자연대류 열전달은 태양열 집열판, 연료탱크, 원자로의 핵반응로, 초전도 자성체의 냉각 및 건물이나 방의 효율적인 열설계, 화재시의 안전대책등 공업적으로 그 응용성이 매우 광범위하고 다양하게 생활주변에서 흔히 볼 수 있다. 밀폐공간내에 경계층 유동이 존재할 경우, 이 경계층 흐름은 일반적으로 외부유동에서의 경계층 유동과는 달리 코어영역(Core region)에서의 흐름과 서로 밀접한 관계가 있다. (중략)

• Journal of the Korean Professional Engineers Association
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• v.34 no.3
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• pp.41-45
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• 2001

The nuclear power plant(NPP) construction is a mammoth project, which requires a huge construction budget and the long construction duration of about 7 years. Recently there is a trend to shorten the construction period, to save the construction cost and to ensure the competitiveness through adopting the open top construction method. In this method, the major equipment are moved into the building through its top, and the internal works of the reactor building are being in process without installation of a dome roof.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.937-941
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• 1995

한국원자력연구소는, 지금까지 연구로 1호(250 ㎾)는 33년째, 연구로 2호(2 ㎿)는 23년째 운전하고 있다 그러나 이들 연구로는 노후하였고, 새로운 연구로 하나로(30 ㎿)가 금년 4월에 준공하였으며, 현재의 연구로 1, 2호가 있는 부지와 건물은 한국전력공사의 소유물이라는 이유로, 금년 말까지 연구로 1, 2호는 폐로할 계획으로 있다 이 논문에는 연구로 1,2호기에 대한 1)폐로 계획의 배경, 2)연구로의 이력, 3)폐로 계획 등을 기술하고 있다.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2014.11a
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• pp.212-213
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• 2014

The purpose of this study is to suggest analysis model of RCB construction in nuclear power plant. For the objective, This study drew the risk factors of RCB construction from existing literature. The results of the study proposed analysis model made hierarchy in rebar, form, and concrete work. These will be baseline data for risk management in construction project of nuclear power plant.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.592-599
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• 1997

중대사고관리 전략의 하나로서 피동형-설계에 적용되고 있는 용기내부보존(IVR)기념 - 이 논문에서는 실제적으로 원자로 압력용기벽 외부냉각(ERVC)방법을 사용한다 -이 규제측면에서는 용융물의 냉각가능성 쟁점의 해결이라는 문맥에서 조감되었다; 기술측면에서는 IVR개념의 신빙성 및 유융성이 언급되었다. 덧붙여서, 이 ERVC방법들이 개량형-설계에 적용되기 위하여 요구되는 점들이 규제측면과 기술측면에서 각각 검토되었다. 이 검토결과의 바탕위에서 용융물 냉각가능성/급냉가능성의 쟁점과 관련하여 전력연구원(KEPRI) 신형원전개발센타(CARD)에서 개발중인 한국차세대원전(KNGR)-설계에서 선택될 수 있는 대안적 전략들이 제안되었다: (1) 전략1A: 젖은공동방법의 신빙성에 기반을 두는 것; (2) 전략1B: 젖은공동방법/격납건물건전성에 기반을 두는 것; (3) 전략2A : ERVC방법의 신빙성에 기만을 두는 것, (4)전략2B: ERVC방법/격납 건물건전성의 균형된 접근법에 기반을 두는 것. 마지막으로, 신형-설계적용의 관점에서 각각 규제측면과 기술측면에서 본 현황파악 및 대책마련의 권고사항이 제시되었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.881-886
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• 1998

한국형 표준 원진(울진 원전 3,4호기)화해 사건에 대한 2 단계 확률론적 안전성평가 (Level 2PSA) 에서 격납건물 파손모드에 큰 영향을 준다고 판단되는 현상들에 대한 불확실성 분석을 수행하였다. 불확실성 분석 대상은 주로 민감도분석 및 기존 2단계 PSA수행결과 중요한 인자로 선정된 8가지 주요 현상들로 국한하였다. 수행 방법은 성층화 추출방식 (Latin Hypercube Sampling)으로부터 발생된 1000개의 표본을 사용하였고, 분석결과는 두가지 불확실성 측도로 제시하였으며, 사용된 코드는 2 단계 PSA 분석용 전산코드인 CONPAS 이다. 불확실성 관리측면에서. 제일 불확실성이 높은 격납건물 파손모드인 원자로 공동바닥관통의 불확실성 인자를 줄이기 위해서는 CR-EJECT 현상에 대한 불확실성 을 줄여야 할 것이다.