• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.1-10
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• 2006

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.842-848
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• 2018

Motivated by reducing the uncertainties in quantification of debris bed coolability, this paper reports an experimental study on two-phase flow resistances and interfacial drag in packed porous beds. The experiments are performed on the DEBECO-LT (DEbris BEd COolability-Low Temperature) test facility which is constructed to investigate the adiabatic single and two phase flow in porous beds. The pressure drops are measured when air-water two phase flow passes through the porous beds packed with different size particles, and the effects of interfacial drag are studied especially. The results show that, for two phase flow through the beds packed with small size particles such as 1.5 mm and 2 mm spheres, the contribution of interfacial drag to the pressure drops is weak and ignorable, while the significant effects are conducted on the pressure drops of the beds with bigger size particles like 3 mm and 6 mm spheres, where the interfacial drag in beds with larger particles will result in a descent-ascent tendency in the pressure drop curves along with the fluid velocity, and the effect of interfacial drag should be considered in the debris coolability analysis models for beds with bigger size particles.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4551-4559
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• 2022

The aim of this research is to analyze the characteristics of a core melt discharged from the reactor vessel and the spreading behavior the core melt in the reactor cavity of the SMART. First, a severe accident sequence under conservative conditions is simulated by the MELCOR code to obtain the conditions for an analysis of the spreading behavior and coolability of the ex-vessel melt. Second, the spreading behavior and coolability of the ex-vessel melt are analyzed by the MELTSPREAD code. The level, temperature, and pressure of the water in the cavity as well as the temperature, mass, composition, and discharge velocity of the melt were utilized to construct the ex-vessel analysis. The melt spread only to part of the cavity, and that the height of the corium in a static state was less than 25 cm. The characteristics of a small modular reactor on the spreading behavior and coolability of melt were analyzed. In the SMART, the amount of melt discharged into the cavity is relatively small and the area of the cavity is sufficiently large when compared to a high-power pressurized water reactor. It was found that the coolability of an ex-vessel core melt can be sufficiently secured.

• 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.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.669-674
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• 1996

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction(MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass, The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment.

• Nuclear Engineering and Technology
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• v.17 no.1
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• pp.53-67
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• 1985

The severe core meltdown accident, which is not included as a design basis accident, has high consequence and low probability of occurrence and turns out to be a major risk factor in the overall risk assessment. The physical mechanisms of containment failure in core meltdown accidents are identified as steam explosion, debris bed coolability, hydrogen burning, steam spike and concrete interaction. The state of technology review is made for each subtopic about the previous and current researches for better understanding of the phenomenon.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.636-643
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• 2024

The coolability of the debris bed with a simulant of solidified corium is experimentally studied, focusing on the effects of the structure of the axial stratified debris bed on the dryout heat flux (DHF). DHF was obtained for the four structures with different particle sizes for the axial stratified debris bed under top flooding. The experimental results show that the dryout position of the axial stratified debris bed is formed at the stratified interface indicated by the temperature rise, and the DHF of the axial stratified bed is much lower than that of the homogeneous bed packed with the upper small particles. To predict the dryout heat flux of the stratified debris beds, by considering the properties of the mixed area, a one-dimensional dryout heat flux model of the porous medium is derived from a water and vapor momentum equation for porous medium, two-phase permeability modifications, interfacial drag, and the correlation between capillary pressure and liquid saturation and verified with the experimental data. The modified model can give reasonable results under different structures.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.206-206
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• 2001

• 한국전산유체공학회:학술대회논문집
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• 2011.05a
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• pp.302-310
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• 2011

A fluid in an enclosure can be heated by electric heating, chemical reaction, or fission heat. In order to remove the volumetric heat of the fluid, the walls surrounding the enclosure must be cooled. In this case, a natural convection occurs in the pool of the fluid, and it has a dominant role in heat transfer to the surrounding walls. It can augment the heat transfer rates tens to hundreds times larger than conductive heat transfer. The heat transfer by a natural convection in a regular shape such as a square cavity or semi-circular pool has been studied experimentally and numerically for many years. A pool of an inverted triangular shape with 10 degree inclined bottom walls has a good cooling performance because of enhanced boiling critical heat flux (CHF) compared to horizontal downward surface. The coolability of the pool is determined by comparing the thermal load from the pool and the maximum heat flux removable by cooling mechanism such as radiative or boiling heat transfer on the pool boundaries. In order to evaluate the pool coolability, it is important to correctly expect the thermal load by a natural convection heat transfer of the pool. In this study, turbulence models with modifications for buoyancy effect were validated for unsteady natural convections by volumetric heating. And natural convection in the triangular pool was evaluated by using the models.