• 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 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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• 2004.05a
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• pp.174.2-174.2
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• 2004

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.781-785
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• 2003

This work presents a numerical analysis of two-phase natural circulation flow in reactor cavity under external vessel cooling. Steady, incompressible, three-dimensional Reynolds-averaged Navier-Stokes equations for multiphase flows with zero equation turbulence model are solved to predict the shear key effect on the circulation rate of cooling water and the distribution of void fraction according to the different mass flow of inlet air. Results show that shear key has a positive effect on the circulation rate of cooling water and induce a local increase of void fraction below the shear key, but not remarkably.

• 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

• Transactions of the Korean Society of Mechanical Engineers B
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• v.31 no.5
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• pp.482-490
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• 2007

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests and the simple analysis were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The calculated circulation flow rate was similar to experimental ones within about ${\pm}$15% error bounds and depended on the form loss due to the inlet/outlet area.

• 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.1578-1583
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• 2003

LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the coolability in case of the external reactor vessel cooling (ERVC). All the 4 tests have been performed using Alumina iron thermite melt as a corium simulant. Due to the limited steam venting through the insulation, steam binding occurred inside the annulus in the KSNP case simulation. On the contrary, in the tests which were performed for simulating the APR1400 insulation design, sufficient water ingression and steam venting through the insulation lead to effective cool down of the vessel characterized by nucleate boiling. It could be found from the experimental results that modification of the insulation design allowing sufficient ventilation could increase the positive effects of the external reactor vessel cooling.