• Proceedings of the Korean Reliability Society Conference
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• 2000.04a
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• pp.161-168
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• 2000

고리 3,4호기 및 영광 1,2호기의 원자로보호계통 및 공학적안전설비작동계통의 정기점검주기와 허용정지시간에 대하여 계통의 신뢰도와 노신손상빈도를 평가하여 안전성이 저해되지 않는 범위에서 합리적인 정기점검주기와 허용정지시간을 도출하였다. 이를 위하여 원자로보호계통의 17개 원자로정지신호와, ESFAS계통의 11개의 안전설비작동신호에 대해서 고장수목을 작성하였고, 정기점검주기 변화에 따른 신뢰도를 평가하였다. 또한 계통의 신뢰도가 발전소의 안전성에 어떤 영향을 미치는가를 파악하기 위하여 노심손상빈도를 분석하였다. 분석 결과 현행 1개월의 점검주기를 3개월로 연장한다 하더라도 신뢰도는 20%미만 노심손상빈도는 2%정도 저하되는 것으로 나타났다. 이런 정도의 신뢰도와 위험도가 변화는 원자력발전소의 안전성에 거의 영향을 주지 못하기 때문에 점검주기를 연장하는 안이 타당한 것으로 나타났다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.389-396
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• 2018

The Kori-1 Nuclear Power Plant (NPP), WH 2-Loop Pressurized Water Reactor (PWR) operated for approximately 40 years in Korea, was permanently ceased on June 18, 2017. To reduce worker exposure to radiation by reducing the dose rate in the system before starting main decommissioning activities, the permanently ceased Kori-1 NPP will be subjected to full system decontamination. Generally, the range of system decontamination includes Reactor Pressure Vessels (RPV), Pressurizer (PZR), Steam Generators (SG), Chemical & Volume Control System (CVCS), Residual Heat Removal System (RHRS), and Reactor Coolant System (RCS) piping. In order to decontaminate these systems and equipment in an effective manner, it is necessary to evaluate the influence of the flow characteristics in the RCS during the decontamination period. There are various methods of providing circulating flow rate to the system decontamination. In this paper, the flow characteristics in Kori-1 NPP reactor coolant according to RHR pump operation were evaluated. The evaluation results showed that system decontamination using an RHR pump was not effective at decontamination due first to impurities deposited in piping and equipment, and second to the extreme flow unbalance in the RCS caused deposition of impurities.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.444.1-444.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.405.1-405.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.394.1-394
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• 2002

• Nuclear Engineering and Technology
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• v.8 no.2
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• pp.101-116
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• 1976

An attempt is made to analyze the unit nuclear power cost of the Go-ri units 1 and 2 in terms of a set of model data. For the calculational purpose, the power cost is first decomposed into the cost components related to the plant capital, operation and maintenance, working capital requirements, and fuel cycle operation. Then, POWERCO-50 computer code is applied to enumerate the first three components and MITCOST-II is used to evaluate the fuel cycle cost component. The specific numerical results are the fuel cycle cost of Go-ri unit 2 for three alternative fuel cycles presumed, levelized unit power cost of units 1 and 2, and the sensitivity of the power cost to the fluctuation of the model data. Upon comparision of the results with the power cost of the fossil power plants in Korea, it is found that the nuclear power is economically preferred to the fossil power. Nevertheless, the turnkey contract value of Go-ri unit 2 appears to be rather expensive compared with the available data on the construction cost of the PWR plants. Therefore, it is suggested that, in order to make the nuclear power plants more attractive in Korea, the unfavorable contract of such kind must be avoided in the future introduction of the nuclear power plant. Capacity factor is of prime importance to achieving the economic generation of the nuclear electricity from the Go-ri plant. Therefore, it is concluded that more efforts should be directed to make the maximum use of the Go-ri plant.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.413-423
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• 1993

This paper reports an experimental study of direct containment heating (DCH) which would occur if the primary system pressure is still high at the time of vessel breach during a light water reactor core melt accident. The experiments were conducted in 1/30-scale cavity models of Kori unit 1 and 2 and Young Kwang unit 3 and 4 nuclear power plants. One 1/20-scale model of the Kori plant was also used to investigate the scaling effect. The primary variables in the experiments were initial vessel pressure, vessel breach size and cavity geometry. It is observed that higher initial pressure and larger breach size enhance the melt dispersal fraction. Also, the cavity geometry appears to affect the dispersal rate greatly. A simple correlation of melt dispersal fraction is proposed in terms of nondimensional effective period. This correlation shows good agreement with the present experimental data, the KAIST data and the BNL data.

• Nuclear industry
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• v.36 no.1
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• pp.37-39
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• 2016

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1377-1378
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.97-97
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• 1998