• Journal of Energy Engineering
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• v.28 no.4
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• pp.103-110
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• 2019

The domestic innovative power reactor named iPOWER will employ the passive molten corium cooling system(PMCCS) to cool down and stabilize the core melt in the severe accident. The final design concept of the PMCCS is yet to be determined, but the in-vessel retention through external reactor vessel cooling has been also considered as a viable strategy to cope with the severe accident. In this study, the two-phase natural circulation flow established between the reactor vessel and the insulation was simulated using a thermal-hydraulic system code, MARS-KS. The flow path of cooling water was modeled with one-dimensional nodes, and the boundary condition of the heat load from the molten core was defined to estimate the naturally-driven flow rate. The evolution of major thermal-hydraulic parameters were also evaluated, including the temperature and the level of cooling water, the void fraction around the lower head of the reactor vessel, and the heat transfer mode on its external surface.

• Transactions of the KSME C: Technology and Education
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• v.3 no.3
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• pp.209-215
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• 2015

The effects of the mounting location of ICI cables on severe accident mitigation systems, specially IVR-ERVC (In-Vessel Retention by External Reactor Vessel Cooling) and core catcher (Ex-vessel corium retention and cooling system), are investigated. The effects of bottom-mounted ICI strategy on severe accident mitigation are summarized and advantages of top-mounted ICI to improve severe accident mitigation are also highlighted.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.745-750
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• 1998

원자로 압력용기 대형 냉각재상실사고에 기인하는 노심용융사고시 노심용융물과 내벽사이의 간극 및 외벽의 열대류계수의 건전성에 대한 영향을 고찰하기 위하여 건전성평가를 수행하였다. 먼저 유한요소해석을 통해 간극 고려 여부와 외벽의 열대류계수 변화에 따른 원자로 압력용기의 온도 및 응력 분포를 결정하였으며, 결정된 온도 및 응력 분포, Larson-Miller 곡선과 손상 법칙을 이용하여 원자로 압력용기의 손상 정도와 파손 시간을 계산하였다.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2000.11a
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• pp.263-268
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• 2000

원자력 압력용기의 건전성 평가 및 안전성 확보에 대한 관심은 1978년 미국 Rancho Seco 발전소에서 발생한 가압열충격 사고로 인해 크게 부각되기 시작하였다. 가압열충격(Pressurized Thermal Shock： PTS)이란 계통의 압력이 높은 상태이거나 증가중인 상태에서 급속한 냉각과 과도한 냉각이 발생하는 것을 의미한다. 이러한 냉각에 의해 원자로용기 외벽보다 내벽이 빨리 냉각되어 상당한 온도구배가 발생하고 이 온도구배에 따라 용기 내벽에 최대인장 열응력이 발생한다.(중략)

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.1897-1902
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• 2003

An 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APRI400 reactor and insulation system. The behaviors of the boiling-induced two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper exit slot area and configuration. And non-heating experiments have also been performed and discussed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.36-43
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• 2023

Lots of researches have been conducted on in-vessel retention (IVR) to prevent or mitigate severe accident in nuclear power plants. Various methodologies were proposed and the external reactor vessel cooling was selected as a part of promising IVR strategy. In this study, the strategy is strengthened by enhancing the natural circulation performance through the adoption of insulation in the reactor cavity. A thermal analysis was carried out based on an assumed accident scenario and its results were used as boundary conditions for subsequent seven flow analysis cases. By comparing the natural circulation performance, effects of annular gaps and insulation shapes on the mass flow rate and flow velocity were quantified. The improvement in cooling performance can be reflected in actual design via detailed assessment.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.212-212
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.189.2-189
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• 2002

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.1927-1932
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• 2004

A 1/21.6 scaled non-heating experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The behaviors of the air bubble-induced two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the injected air flow rate and distribution. As the injected air flow rates increased, the natural circulation flow rates also increased. Both the longitudinal and the latitudinal distributions of the injected air affected the natural circulation flow rates, especially, the longitudinal effect is more larger.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2141-2145
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• 2004

As part of study on thermal hydraulic behavior in the reactor cavity under external vessel cooling in the APR (Advanced Power Reactor) 1400, one dimensional two phase flow of steady state in the reactor cavity have been analyzed to investigate a coolant circulation mass flow rate in the annulus region between the reactor vessel and the insulation material using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that a two phase natural circulation flow of 300 - 600 kg/s is generated in the annulus region between the reactor vessel and the insulation material when the external vessel cooling has been applied in the APR 1400. An increase in the heat flux of the inner vessel leads to an increase of the coolant mass flow rate. An increase in the coolant outlet area leads to an increase in the coolant circulation mass flow rate, but the coolant inlet area does not effective on the coolant circulation mass flow rate. The change of the lower coolant outlet to a lower position affects the coolant circulation mass flow rate, but the variation trend is not consistent.