• Proceedings of the Korea Water Resources Association Conference
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• 2011.05a
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• pp.250-250
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• 2011

지정학적으로 3면이 바다로 접하고 있는 우리나라 해안가 주변의 원자력발전소는 관류형 냉각 방식(Once-through cooling System)을 채택하고 있기에, 이 계통을 통과한 냉각수는 주변해역으로 흘러들어가 주변 환경에 영향을 줄 수 있다. 또한 동해 주변 인접국들의 활발한 원자력 이용과 방사능 물질의 수송 등으로 인해 방사능 사고 위험이 증대됨에 따라 동해 원전 주변의 해양방사성물질 거동에 대한 필요성이 제기되고 있다. 이에 따라 동해의 해양환경을 조사 분석하여 해역에서의 오염 물질의 거동 파악이 중요하고, 산업시설 주변 연안의 오염물 이동 및 해석을 위해 오염물의 시 공간적으로 농도를 추정하는 것이 중요한 사항이다. 따라서 본 연구에서 수직 수평적으로 2차원 및 3차원화가 용이한 EFDC 모델을 사용하였다. 국립해양조사원에서 발간하는 DC103의 수치해도를 이용하여 모델영역은 동서방향으로 171km, 남북방향으로 235km로 설정하고, 격자간격은 $1km{\times}1km$의 정방향 격자를 사용하여 동해 월성 주변의 조위 시계열 검증 및 조화분석을 통한 검증을 실시한 결과 관측한 결과와 양호한 재현성을 나타내었다. 또한 국립수산과학원에서 제공되고 있는 년별 수온, 염분 자료 등의 해양환경과 지형적 특성을 고려하여 오염 물질의 이동 경로를 파악함으로써 사고시 피해를 줄이고자 한다.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.525-530
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• 1995

최적 열수력 전산 코드인 CATHARE2 Vl.3u 코드를 이용하여 영광 3/4호기 midloop 운전중 잔열제거(RHR) 기능 상실사고를 해석하였다. 본 연구의 주된 목적은 사고시 계통에서 발생하는 열수력 현상의 이해 향상 및 증기발생기 열제거 능력 평가에 있다. 사고 복구 절차 관점에서 노심 비등, 노출 시점 및 계통압력 등이 중요한 인자이다. 본 계산 수행시 사용한 가정은 다음과 같다. 가) 초기 계통 수위는 고온관 중간에 위치하며 그 윗 부분은 질소 가스로 차 있다. 나) 3/4 인치 크기의 방출 밸브가 원자로 용기 상부 및 가압기 상부에 각각 설치되어 있으며, RHR 흡입구에 수위지시계가 설치되어 있다. 다) 증기발생기의 이차측은 U-튜브가 잠기도록 물로 차있다. 라) 두 증기발생기의 대기 방출 밸브(ADV)는 항상 열려 있어 사고시 이차측 압력을 대기압으로 유지하기에 충분하다. 사고는 원자로 정지 2일 후 발생하였다고 가정한다. 이와 같은 조건하에서 사고시 주된 계통 열제거 수단은 증기발생기 U-튜브내의 응축 작용이며 이는 전체 열제거량의 94%로 나타났다. 노심 비등 시점온 사고후∼300초 이후이며, 계통압력은 10,800초 이후에 최고 압력인 0.25MPa에 도달한 후 그 값을 계속 유지하고 있다. RHR 배관에 연결된 수위지 시계를 통해 10,200초 이후부터 냉각수가 방출되었다. 2개의 방출밸브 및 수위지시계를 통하여 방출된 유량에 근거하여 원자로 용기 냉각재 수위가 고온관 바닦까지 낮아지는 시점을 계산하면 사고 약 6.4 시간 이후가 된다.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.265-275
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• 1993

A vapor explosion has been a concern in nuclear reactor safety due to its potential for a destructive mechanical energy release. In order to properly assess the hazard of a vapor explosion, it is necessary to accurately estimate the conversion efficiency of the thermal energy to mechanical energy. In the absence of a complete model to determine the explosive energy yield, one may have to rely on a simpler upper bound estimate such as a thermodynamic model. This paper discusses various thermodynamic models and presents a clarification of each model in their mathematical formulation and the thermodynamic work conversion. It is shown that the work release in the shock adiabatic model of Board and Hall is essentially equal to that of Hicks-Menzies thermodynamic model. The effect of coolant void fraction on the explosion efficiency is also predicted based on these thermodynamic models. Finally, the Hicks-Menzies model is modified to account for the chemical reaction between a metallic fuel and water and the resultant effects on the explosion expansion work are discussed.

• Proceedings of the Korean Vacuum Society Conference
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• 2015.08a
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• pp.92.1-92.1
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• 2015

열전소재의 열전성능을 평가하기 위해서는 샘플 내 상하부 온도구배가 필요하다. 하지만 진공챔버 내에서는 대류효과가 제한되어, 1 mm 이하 두께의 얇은 샘플은 상하부 방향으로 온도 구배를 조성하기 어렵다. 온도 구배를 조성하기 위해서는 샘플의 두께 방향을 관통하는 열유속이 필요하며, 진공 분위기에서 열유속을 조성하기 위해서는 히터뿐만 아니라 별도의 열배출기가 요구된다. 본 연구에서는 열전특성 측정 장비 내 수냉식 열배출기의 설계를 위해, 열배출기의 층수를 달리하며 열전달거동을 수치해석적인 방법으로 연구하였다. 열배출기의 층수에 따른 영향을 평가하기 위해서 동일한 채널길이를 유지시키면서 층수를 달리하는 기하학적인 구조들을 설계하였다. 수치해석을 용이하게 진행하기 위해, 열배출기의 형태는 단순한 bar 형태를 가진 1-5층의 다층 구조 로 설계하였다. 열배출기들 각각의 열전달 효율을 평가하기 위해, 수냉식 열배출기의 열배출량에 가장 큰 영향을 미치는 질량유량을 0.1-1 g/s로 변화 시키면서 열전달 거동을 확인하였다. 또한 냉각수의 기화 현상을 방지하기 위해 발열체의 온도를 290-370 K로 바꿔 가며 열전달 거동을 확인하였다. 수치 해석결과, 5층의 열배출기가 최대 120 W/cm2 로 높은 단위면적당 열배출량을 가지는 것을 확인하였으나, 열배출기 전체의 열배출량을 기준으로하는 열배출효율은 0.6 정도로 낮은 효율을 가짐을 확인하였다. 반면에 3층의 열배출기의 경우, 열배출 효율이 0.8에 달하며, 2층의 열배출기 보다 열배출 효율이 좋다는 것을 확인할 수 있었다.

• Journal of Energy Engineering
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• v.16 no.3
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• pp.103-112
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• 2007

Flood issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feedwater line region of intermediate building and lower floor. A calculation for flood level at the main feedwater line isolation compartment is now performing by hand calculation. But, this methodology is quite conservative assumption. The goal of this study was to develop method to analyze flowrate using the RETRAN-3D computer code, and the developed method was applied to flood level analysis following main feedwater line break. As a result of analysis, flood level was low remarkably.

• Nuclear Engineering and Technology
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• v.20 no.4
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• pp.253-264
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• 1988

Our experience of neutron noise analysis in French-type 900 MWe pressurized water reactor (PWR) is presented. Neutron noise analysis is based on the technique of interpreting the signal fluctuations of ex-core detectors caused by core reactivity changes and neutron attenuation due to lateral core motion. It also provides advantages over deterministic dynamic-testing techniques because existing plant instrumentation can be utilized and normal operation of the plant is not disturbed. The data of this paper were obtained in the ULJIN unit 1 reactor during the start-up test period and the statistical descriptors, useful for our purpose, are power spectral density (PSD), coherence function (CF), and phase difference between detectors. It is found that core support barrel (CSB) motions induced by coolant flow forces and pressure pulsations in a reactor vessel were indentified around 8 Hz of frequency.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2010.11a
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• pp.231-234
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• 2010

Analysis of conjugated heat transfer has been conducted for the diffuser exposed to hot combustion gas to design the mechanical durability in high temperature. All the heat transfer means, conduction, convection and radiation have been considered to calculate the total heat flux from hot gas to diffuser surface. The calculation has been implemented by two kinds of methods. One thing is one dimensional method based on empirical equations. The other is CFD(Computational Fluid Dynamics) axisymmetric calculation containing ${\kappa}-{\omega}$ SST(Shear Stress Transport) turbulent model and DO(Discrete Ordinate) radiation model. The derived results of two methods have compared and showed similar values. From this result, the amount of cooling water and the dimension of water cooling channel were decided.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.8
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• pp.767-774
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• 2010

In a passive-type PEMFC stack, axial fans operate to supply both oxidant and coolant to cathode side of the stack. It is possible to make a simple system because the passive-type PEMFC stack does not require additional cooling equipment. However, the performance of a cooling system in which water is used as a coolant is better than that of the air-cooling system. To ensure system reliability, it is essential to make cooling system effective by adopting an optimal stack design. In this study, a numerical investigation has been carried out to identify an optimum cooling strategy. Various channel configurations were applied to the test section. The passive-type PEMFC was tested by varying airflow rate distribution at the cathode side and external heat transfer coefficient of the stack. The best cooling performance was achieved when a channel with thick ribs was used, and the overheating at the center of the stack was reduced when a case in which airflow was concentrated at the middle of the stack was used.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.4
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• pp.580-589
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• 2019

Indonesia has a very short supply of electricity. As a solution to this problem, plans for construction of thermal power plants are increasing. Thermal power plant require the cooling water system to cool the overheated engine and equipment that accompany power generation, and the circulation water pump chamber among the cooling water system are generally designed according to the ANSI (1998) standard. In this study, the design criterion $20^{\circ}$ for the spreading angle of the ANSI (1998) of the layout of the circulating water pump chamber can not be satisfied on the K-coal thermal power plant site condition in Indonesia. Therefore, 3-D numerical model experiment was carried out to obtain a hydraulically stable flow and stable structure. The Flow-3D model was used as numerical model. In order to examine the applicability of the Flow-3D model, the flow study results around the rectangular structure of Rodi (1997) and the numerical analysis results were compared around the rectangular structures. The longitudinal velocity distribution derived from numerical analysis show good agreement. In order to satisfy the design velocity in the circulating water pump chamber, a rectangular baffle favoring velocity reduction was applied. When the approach velocity into the circulating water pump chamber was occurred 1.5 m/s ~ 2.5 m/s, the angle of the separation flow on the baffle was occurred about $15^{\circ}{\sim}20^{\circ}$. By placing the baffle below the separation flow angle downstream, the design velocity of less than 0.5 m/s was satisfied at inlet bay.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1115-1130
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• 2012

In nuclear power plants, the reactor pressure vessel (RPV) upper head control rod drive mechanism (CRDM) penetration nozzles are fabricated using J-groove weld geometry. Recently, the incidences of cracking in Alloy 600 CRDM nozzles and their associated welds have increased significantly. The cracking mechanism has been attributed to primary water stress corrosion cracking (PWSCC), and it has been shown to be driven by welding residual stresses and operational stresses in the weld region. The weld-induced residual stress is the main factor contributing to crack growth. Therefore, an exact estimation of the residual stress is important for ensuring reliable operation. This study presents the residual stress computation performed for an RPV CRDM penetration nozzle in Korea. Based on two and three dimensional finite element analyses, the effect of welding variables on the residual stress variation is estimated for sensitivity analysis.