• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.309-314
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• 1996

Feasibility study has been accomplished to evaluate the effectiveness of the in-vessel core debris cooling through lower cavity flooding using two dimensional finite difference scheme. The volume of cerium pool and decay power rate generated in corium pool were evaluated as important parameters to the temperature distribution on the reactor vessel lower head through previous works. In this study, the corium volume based on the System 80＋ core structure and time dependent decay power rate are considered for feasibility evaluation. In addition, preliminary plans for the in-vessel core debris cooling through lower cavity flooding as severe accident management strategy, i.e. flooding timing, method and capacity, are suggested based on the result of the numerical study, international tendency related to in-vessel core debris cooling through lower cavity flooding.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.153-153
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.161.2-161
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.194-194
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• 1999

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.469-470
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.717-718
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• 2005

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3086-3091
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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 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 rectangular channel area of the APR1400 reactor vessel. As the water inlet area increased, the natural circulation mass flow rate asymptotically increased, that is, it converged at a specific value. And the circulation mass flow rate also increased as the outlet area, injected air flow rate, and outlet height increased. But the circulation mass flow rate was not changed along with the external water level variation if the water level was higher than the outlet height.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.24 no.11
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• pp.1447-1456
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• 2000

Simulated experiments of high pressure melt ejection(HPME) are performed to measure the released fraction of corium simulant from the French type PWR cavity. The experiments are carried out on a 1/20th linear scaled model of the Ulchin 1&2 cavity. Water or woods metal and nitrogen is used as simulant of molten corium and steam, respectively. Experimental parameters are water mass, annulus area and breach size. It is shown that only breach size effects is very important while the mass and the annulus area do not affect the released fraction. It is found that the liquid film transport is much more dominant mechanism than the entrainment droplet transport, especially in linear scale down simulated HPME experiment.