• Nuclear Engineering and Technology
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• 제50권6호
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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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• 제38권1호
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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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• 제56권2호
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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.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.117-129
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• 2022

The downward discharge of the molten fuel to the lower structure of the fuel assembly could increase of the pressure drop and degrade of coolability of the assembly. To analyze the phenomena, experiments for the generation of the debris bed were conducted as LOF-DT series. Based on the debris bed in the LOF-DT, pressure drop experiment was conducted with intact and blocked component. Parametric study on the pressure drop was conducted by CFD. The LOF-DT experiments were conducted for the position and porosity of the debris bed. 85% of the debris were sedimented in the lower reflector, and 15% were in the nose piece, approximately. Porosity of the debris bed were about 0.7 and 0.85 in the lower reflector and nose piece, respectively. Pressure drop increased significantly with debris bed, especially in the lower reflector. More than 120 time of the pressure drop increased in the lower reflector, while only 10% increased in the nose piece. According to the parametric study, mass of the debris was the most important for pressure drop. The lower discharge phenomena could have a significant effect to the total pressure drop of the fuel assembly, approximately 10.8 times for the base case.

• Nuclear Engineering and Technology
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• 제17권1호
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• pp.53-67
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• 1985

중대 노심사고는 아직 Design Basis Accident에 포함되지 않고 있으나, 극히 적은 사고 확률을 가지는 반면 사고 후 영향이 큼으로해서 원자력발전소의 전반적 위험 평가에 중요한 요인중의 하나가 되고 있다. 중대 노심사고시 격납용기 손상에 관련된 물리현상들은 Steam Explosion, Debris Bed Coolability, Hydrogen Burning, Steam Spike와 Core-Concrete Interaction 등이며, 각각의 현상에 대한 좀 더 나은 이해를 위해 현재 이루어지고 있는 연구들에 대한 개략적 설명을 시도 하였다.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.457-469
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• 2016

During the late phase of severe accidents of light water reactors, a porous debris bed is expected to develop on the bottom of the flooded reactor cavity after breakup of the melt in water. The geometrical configuration, i.e., internal and external characteristics, of the debris bed is significant for the adequate assessment of the coolability of the relocated corium. The internal structure of a debris bed was investigated experimentally using the DAVINCI (Debris bed research Apparatus for Validation of the bubble-Induced Natural Convection effect Issue) test facility. Particle sedimentation under the influence of a two-phase natural convection flow due to the decay heat in the debris bed was simulated by dropping various sizes of particles into a water vessel with air bubble injection from the bottom. Settled particles were collected and sieved to obtain the particle mass, size distribution in the radial and axial positions, and the bed porosity and permeability. The experimental results showed that the center part of the particle bed tended to have larger particles than the peripheral area. For the axial distribution, the lower layer had a higher fraction of larger particles. As the sedimentation progressed, the size distribution in the upper layers can shift to larger sizes because of the higher vapor generation rate and stronger flow intensity.

• Nuclear Engineering and Technology
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• 제17권1호
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• pp.8-15
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• 1985

Debris bed내·외로부터 깨스유량을 갖는 debris/water 열적상호작용 해석모델이 중대사고 분석을 위해 제시되었다. 제시된 모델은 증기 소비, debris bed에서 수소 생성, 유입깨스 및 화학반응열에 대한 인자들을 포함하고 있으며, 금속-물반응 및 debris/concrete 작용으로 인한 깨스 생성을 평가하기 위해 MARCH code에 도입시켰다. 그 결과 수소원은 격납용기 과도압력에 큰 영향을 미치나 debris bed로 대류깨스 냉각과 콘크리트로 전도 열손실은 debris bed 냉각성에 조그마한 영향을 주는 것으로 나타났다. 하지만 debris 인자의 재가열과 재용융은 콘크리트와 상호작용에 의해 상당히 지연될 수 있다.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4003-4013
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• 2021

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1, consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [2]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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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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• 제41권5호
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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.