• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.261-272
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• 2012

Molten Core Concrete Interaction (MCCI) is a complex process characterized by concrete ablation and volatile generation; Thermal and solutal convection in a bubble-agitated melt; Physico-chemical evolution of the corium pool with a wide solidification range (of the order of 1000 K). Twelve experiments have been carried out in the VULCANO facility with prototypic corium and sustained heating. The dry oxidic corium tests have contributed to show that silica-rich concrete experience an anisotropic ablation. This unexpected ablation pattern is quite reproducible and can be recalculated, provided an empirical anisotropy factor is assumed. Dry tests with oxide and metal liquid phases have also yielded unexpected results: a larger than expected steel oxidation and unexpected topology of the metallic phase (at the bottom of the cavity and also on the vertical concrete walls). Finally, VULCANO has proved its interest for the study of mitigation solutions such as the COMET bottom flooding core catcher.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.268-271
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• 2008

If a reactor vessel is failed to retain a molten corium in a postulated severe accident, the molten corium is released outside the reactor vessel into a reactor cavity. The molten corium would attack the concrete wall and basemat of the reactor cavity, which may lead to inevitable concrete decompositions and possible radiological releases. In the OECD/MCCI project, a series of tests were performed to secure the data for cooling the molten corium spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction). Also, a MCCI (Molten Core Concrete Interaction) analysis code, CORQUENCH was upgraded at Argonne National Laboratory with embedding the new models developed for the tests. This paper deals with analyses of MCCI at plant scale under the conditions of top flooding using the upgraded CORQUENCH code. The modeling approach is briefly summarized first, followed by presentation of a validation calculation that illustrates the predicative capability of the modeling tool. With this background in place, the model is then used to carry out a parametric set of scoping calculations that define approximate coolability envelopes for the LCS (Limestone Common Sand) concrete that has been evaluated in the OECD/MCCI project.

• 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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• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.777-782
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• 1998

To terminate the progression of the sever accident and mitigate the accident consequences, corium coaling has been suggested as one of most important design features considered in the swore accident mitigation. Till now, some kinds of cooling methodologies have been identified and, specially the corium cooling at the reactor cavity has been considered as one of the most promising cooling methodologies. Moreover, several design requirements related to the cerium cooling at the reactor cavity have been also suggested and applied to the design of the next generation reactor. In this study technical description are briefly described for the important issues related to the cerium cooling at the reactor cavity, i.e. cavity area, cavity flooding system, etc., and simple evaluation for those items have been performed considering present technical levels the experiment and analytical works..

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.437-442
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• 1998

Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling heat been performed. However, there is no available correlation adequate for severe accident analysis. In this study, boiling heat transfer correlations have been developed, and their applicable ranges heat been enlarged and their prediction accuracy has been enhanced.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1508-1515
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• 2022

In the present study, the cavity module of the MELCOR code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous MELCOR Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the MELCOR code to investigate the impact of water ingression and melt eruption models which were hitherto absent in MELCOR code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of MELCOR code has led to the reduction of the maximum ablation depth in radial and axial directions by ~38% and ~32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the MELCOR code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.225.2-225
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• 2001

• The Korea Journal of Herbology
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• v.22 no.4
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• pp.29-34
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• 2007

Objective : Kyenegum(Galli Stomachichum Corium) has been popularly used long as the digestive. The purpose of this study was to investigate the enzymatic characteristic of Kyenegum crude enzyme. Methods : To evaluate of the enzymatic characteristic of Kyenegum, we examined the activity of Kyenegum crude enzyme from optimum solvent, optimum temperature and pH of crude Kyenegum extract. Futhermore, we examined the effects of NaCI and acidity of crude Kyenegum extract. Results : The Kyenegum was composed with crude protein about 20%, crude lipid 2%. The optimum Kyenegum dry condition, optimum extract solvent, optimum temperature and optimum pH were $4{\sim}6$ hours at $60^{\circ}C$, commercial apple vinegar, $50^{\circ}C$ and 2.0. Conclusion : The result suggests that the Kyenegum crude enzyme extract very strong enzyme in temperature, NaCl and acidity, respectively.