• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.298-303
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• 1997

실제의 격납건물의 구조는 하부 원통형의 구조를 가지는 영역과 상부 돔 형태와 굴뚝 형태의 구조를 가지는 영역으로 나눌 수 있다. 하부 원통형의 구조만을 고려한다면, 고온의 철제 벽면과 콘크리트 벽면 사이의 gap 크기에 비해서 원통의 반지름이 상대적으로 매우 큰 값을 가지기 때문에 2차원 무한평판으로 가정하는 것이 가능하다. 그러나 돔 및 굴뚝 영역에서는 높이가 높아질수록 돔 단면직경이 감소하고 굴뚝 영역도 유동단면적이 작은 원통의 구조를 가져 2차원 무한평판의 가정에 많은 무리가 따른다. 앞에서 명시한 세 가지의 격납건물 형태에 있어서 ASPWR의 경우는 굴뚝을 포함한 영역까지도 무한평판으로 가정하는 것이 가능하나(돔에서의 열전달 단면적이 하부의 열전달 단면적에 비해 매우 작다는 가정을 한다면) 나머지 AP600과 HWRF의 격납건물에 있어서는 상부까지도 무한평판 가정을 사용하는 것에는 무리가 있다. 본 연구에서는 일반적인 유체해석 코드인 FLUENT V4.3을 이용하여 실제 격납건물 구조에 대한 분석을 시도하여 무한평판 구조에 대한 가정이 과도한 열전달량을 예측하고 있음을 확인하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.534-541
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• 1997

1000MWe급 피동형원자로기 안전계통 성능 및 RELAP5 코드의 적용성 평가를 목적으로 AP600을 참조노형으로 설정된 1000MWe급 대용량 피동형원자로에 대한 냉각재 상실사고를 모의 해석하였다. 대형냉각재상실사고시 발생되는 현상들은 기존 원자로와 큰 차이가 없고, 이들 현상을 모의하기 위한 모델링 요건들이 피동형계통 분석에 동일하게 요구되었으며, 계산된 PCT가 규제기관의 허용치에 충분한 여유도를 갖고 있어 대형냉각재상실사고시 충분한 노심냉각 능력을 갖는 것으로 평가되었다. 또한 안전주입 배관이 파단되는 소형냉각재 상실사고를 해석한 결과 KP1000의 피동안전계통은 ADS의 작동에 의하여 노심을 노출시키지 않고 적절한 사고완화 기능을 수행할 수 있는 것으로 분석되었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.473-478
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• 1997

A severe accident management strategy, in-vessel retention corium through external reactor vessel cooling(ERVC) is being studied worldwide as a means to prevent reactor vessel failure following a core melt accident. An evaluation of feasibility of this ERVC for a large Advanced Light Water Reactor (ALWR) is presented. To account for the coolability of corium and metal in the reactor vessel, a thermal analysis is performed using an existing method. Results show that the peak heat flux along the inner surface of the reactor vessel lower head has a relatively smaller margin than a small capacity reactor such as AP600 in regards with the critical heat flux attainable at the outer surface of the reactor vessel lower head.

• Nuclear Engineering and Technology
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• v.30 no.1
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• pp.26-39
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• 1998

The containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5 is modified to incorporate the passive containment cooling models. The correlations are selected from the existing experimental heat transfer correlations to model the natural and mixed convection in annular space between the containment shell and the shield building. The evaporative heat transfer of the water film on the outer shell of the containment is modeled using the correlations derived from the analogy between the heat and mass transfer. The modified code is applied to the Ap600 containment transient analysis for the model verification and the results are compared to the results of GOTHIC calculation done by Westinghouse. Also, d series of parametric sensitivity studies of heat transfer correlations, water film ratio and delay time of the wet cooling on the containment peak pressure and temperature following LOCA are performed for the containment of 1000MWe passive plant, KP1000.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.423-428
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• 1996

피동형원자로 AP600을 참조발전소로하여 설정된 1000MWe급 대용량 피동형원자로의 계통개념에 대한 안전계통 성능 평가 및 코드의 적용성 평가를 목적으로 RELAP5/MOD3코드를 사용하여 대형냉각재상실사고를 모의 해석하였다. 피동형 안전계통으로 축압기, CMT IRWST를 모델하였으며 가압기에 연결된 1단계부터 3단계까지의 자동감압밸브계통을 모델링 하고 4단계 자동감압밸브계통은 각 루프의 고온관에 연결되어 있는 것으로 모델링 하였다. 피동형 안전계통의 모델이 향상된 RELAP5/MOD3.2와 그 이전의 코드인 RELAP5/MOD3.1의 냉각재상실사고 모의계산결과 원자로내의 압력변화, 노심냉각수 주입유량 및 핵연료 피복재 온도 거동이 거의 유사하게 나타났으며 1000MWe급 대용량 피동형원자로의 안전계통은 냉각재 상실사고시 충분한 노심냉각능력을 가지는 것으로 분석되었다.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.291-298
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• 1997

In advanced nuclear reactors, the passive containment cooling has been suggested to enhance the safety. The passive cooling has two mechanisms, air natural convection and oater cooling with evaporation. To confirm the coolability of PCCS, many works have been performed experimentally and numerically. In this study, the water cooling test was performed to obtain the evaporative heat transfer coefficients in a scaled don segment type PCCS facility which have same configuration with AP600 prototype containment. Air-steam mixture temperature and velocity, relative humidity and well heat flux are measured. The local steam mass flow rates through the vertical plate part of the facility are calculated from the measured data to obtain evaporative heat transfer coefficients. The measured evaporative heat transfer coefficients are compared with an analytical model which use a mass transfer coefficients. From the comparison, the predicted coefficients show good agreement with experimental data however, some discrepancies exist when the effect of wave motion is not considered. Finally, a new correlation on evaporative heat transfer coefficients are developed using the experimental values.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.59-64
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• 1996

Feasibility of soluble boron free operation for the medium-sized commercial reactors was investigated. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with gadolinia burnable poison-high Gd enrichment and axial poison zoning. CASMO and NECTA-C code system checked axial offset and peaking factors as fuels burned up. A core with complex axial burnable poison zoning satisfied design goals - small excess reactivity for 18 month cycle. Therefore, critical bank positioning for three control rod banks was sought with ease. A.O. value and Fq value were kept within the safety limit.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.517-524
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• 1995

가압경수형 피동형 원자로의 대표적인 노형인 미국 westinghouse사의 AP600을 참조발전소로하여 피동형기기 용량 및 캔드형 원자로냉각재 펌프의 관성에 대한 민감도 분석을 수행하였다. 분석결과 축압기 및 노김보충수탱크는 용량을 20% 감소시킨 경우에 대해서도 핵연료 피복재 온도는 설계기준치를 충분한 여유도를 가지고 만족하고 있는 것으로 분석되었으며, 중력과 밀도차이에 의하여 형성되는 자연대류를 이용하는 피동잔열제거계통의 성능은 초기조건보다는 기기의 용량과 위치에 더 큰 영향을 받는 것으로 나타났다. 또한 원자로냉각재 펌프의 관성이 증가함에 따라 DNB 여유도가 증가하며 저관성일 경우 trip system의 지연시간이 중요한 것으로 나타났다.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.32 no.6A
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• pp.600-607
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• 2007

In this paper, we propose a MLKA (Multi-Level Keep-Alive) algorithm based on PRCG (Periodic In this paper, we propose a MLKA (Multi-Level Keep-Alive) algorithm based on PRCG (Periodic Ranging Code Group), in which periodic ranging codes are divided into groups. The AP (Access Point) performs this algorithm when a periodic ranging code arrives at the AP in a WiBro (Wireless Broadband Internet) system. In order to increase the traffic radio bandwidth resource efficiency in this system, we propose the multi-level keep-alive algorithm for finding abnormal AT (Access Terminal). From simulation results, it is verified our proposed MLKA algorithm based on PRCG can provide higher traffic radio resource efficiency compared to message-driven keep-alive algorithms.

• Transactions of the Korean Society of Mechanical Engineers
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• v.12 no.5
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• pp.1189-1196
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• 1988

Thermal entrance lengths of turbulent tube flow for viscoelastic polymer solutions are investigated experimentally in the recirculating flow system with tubes of inside diameters 8.5mm(L/D=710) and 10.3mm(L/D=1158), respectively. In the present system, the hydrodynamic and thermal boundary layers develop simultaneously from the beginning of the test section. To provide the boundary condition of constant heat flux at the wall, the test tubes are heated directly by electricity. The polymer solution used in the current study is 1000 wppm aqueous solution of polyacrylamide(Separan AP-273). The apparent viscosity of the polymer solutions circulating in the flow system are measured by the capillary tube viscometer at regular time intervals. Thermal entrance lengths vary due to the rate of degradation. The entrance lengths of degraded polymer solutions are about 500~600 times the diameter. However, the entrance lengths of fresh polymer solutions are greater than the lengths of the test tubes used in this study suggesting that thermal entrance lengths for viscoelastic polymer solutions are greater than 1100 tube times the diameters. Friction factor is almost insensitive to the degradation, but the heat transfer $j_{H}$-factor is affected seriously by degradation. Based on the present experimental data of fresh solutions a correlation for the heat transfer $j_{H}$-factor is presented.ted.