• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.406-407
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• 2005

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1330-1337
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• 2016

A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.603-616
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• 2009

The past two decades were mainly devoted to model validation and computer code verification against global corium experiments, code application to reactor situations, and investigation of the role of melt properties in steam explosion energetics. Corium data were essentially provided by JRC-Ispra in the FARO and KROTOS facilities and by KAERI in the TROI facility. Verification of code applicability to reactor situations was performed essentially in the frame of the international OECD/SERENA programme. The paper makes a synthesis of the findings made during the above-mentioned period and expresses a personal view of the author with respect to the progress made and expected for the resolution of the steam explosion issue for light water reactors.

• Proceedings of the KSME Conference
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• 2001.06e
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• pp.51-56
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• 2001

Vapor explosion is one of the most important problems encountered in severe accident management of nuclear power plants. In spite of many efforts, a lot of questions still remain for the fundamental understanding of vapor explosion phenomena. Therefore, KAERI launched a real material experiment called TROI using 20 kg of UO2 and ZrO2 to investigate the vapor explosion phenomena. In addition, a small-scale experiment with molten-tin/water system was performed to quantify the characteristics of vapor explosion and to understand the phenomenology of vapor explosion. A number of instruments were used to measure the physical change occurring during the vapor explosion. In this experiment, the vapor explosion generated by molten fuel water interaction is visualized using high speed camera and the pressure behavior accompanying the explosion is investigated.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.407-410
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• 2002

A series of steam explosion experiments using real core materials of $ZrO_2$ and corium(a mixture of $ZrO_{2}\;and\;UO_{2}$) has been performed to evaluate the risk of steam explosion load in nuclear power plants. Surprisingly, spontaneous steam explosions are observed far both materials, which have been thought to be inexplosive so far. The dynamic pressure and morphology of the debris clearly indicate the evidence of an explosion. The experimental results also indicate that $ZrO_2$ is more explosive than corium.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.617-648
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• 2009

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.344-357
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• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.25 no.12
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• pp.1887-1897
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• 2001

Korea Atomic Energy Research Institute (KAERI) launched an intermediate scale steam explosion experiment named "Test for Real Corium Interaction with water (TROI)" using reactor material to investigate whether the molten reactor material would lead to energetic steam explosion when interacted wish cold water at low pressure. The melt-water interaction experiment is performed in a pressure vessel with the multi-dimensional fuel and water pool geometry. The novel concept of cold crucible technology, where powder of the reactor material in a water-cooled cafe is heated by high frequency induction, is firstly implemented for the generation of molten fuel. In this paper, the lest facility and cold crucible technology are introduced and the results or the first series of tests were discussed. The 5 kg of molten ZrO$_2$jet was poured into the 67cm deep water pool at 30 ∼ 95 $\^{C}$. Either spontaneous steam explosions or quenching was observed. The morphology of debris and pressure wave profiles clearly indicate the differences between the two cases.