• Nuclear Engineering and Technology
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• 제45권3호
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• pp.335-346
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• 2013

The Critical Heat Flux (CHF) of water with dispersed alumina nanoparticles was measured for the geometry and flow conditions relevant to the In-Vessel Retention (IVR) situation which can occur during core melting sequences in certain advanced Light Water Reactors (LWRs). CHF measurements were conducted in a flow boiling loop featuring a test section designed to be thermal-hydraulically similar to the vessel/insulation gap in the Westinghouse AP1000 plant. The effects of orientation angle, pressure, mass flux, fluid type, boiling time, surface material, and surface state were investigated. Results for water-based nanofluids with alumina nanoparticles (0.001% by volume) on stainless steel surface indicate an average 70% CHF enhancement with a range of 17% to 108% depending on the specific flow conditions expected for IVR. Experiments also indicate that only about thirty minutes of boiling time (which drives nanoparticle deposition) are needed to obtain substantial CHF enhancement with nanofluids.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1405-1413
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• 2017

To investigate the heat loads imposed on a reactor vessel through the natural convection of core melts in severe accidents, mass transfer experiments were performed based on the heat transfer/mass transfer analogy, using two- (2-D) and three-dimensional (3-D) facilities of various heights. The modified Rayleigh numbers ranged from $10^{12}$ to $10^{15}$, with a fixed Prandtl number of 2,014. The measured Nusselt numbers showed a trend similar to those of existing studies, but the absolute values showed discrepancies owing to the high Prandtl number of this system. The measured angle-dependent Nusselt numbers were analyzed for 2-D and 3-D geometries, and a multiplier was developed that enables the extrapolation of 2-D data into 3-D data. The definition of $Ra^{\prime}_H$ was specified for 2-D geometries, so that results could be extrapolated for 3-D geometries; also, heat transfer correlations were developed.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.906-914
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• 2016

We investigated the heat load imposed on the lower head of a reactor vessel by the natural convection of the oxide pool in a severe accident. Mass transfer experiments using a $CuSO_4-H_2SO_4$ electroplating system were performed based on the analogy between heat and mass transfer. The $Ra^{\prime}_H$ of $10^{14}$ order was achieved with a facility height of only 0.1 m. Three different volumetric heat sources were compared; two had identical configurations to those previously reported, and the other was designed by the authors. The measured Nu's of the lower head were about 30% lower than those previously reported. The measured angular heat flux ratios were similar to those reported in existing studies except for the peaks appearing near the top. The volumetric heat sources did not affect the Nu of the lower head but affected the Nu of the top plate by obstructing the rising flow from the bottom.

• Nuclear Engineering and Technology
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• 제41권5호
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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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• 제55권1호
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• pp.180-189
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• 2023

We measured the heat load to a reactor vessel with and without the in-vessel debris bed under an IVR-ERVC condition. Mass transfer methodology was adopted based on heat and mass transfer analogy to achieve high Ra'_H of order ~10¹⁵ with compact test rigs. We postulated the in-vessel debris bed has a flat top and particulate debris was simulated as an identical diameter spheres. We conducted experiments varying the height of the debris bed and the results showed that Nusselt numbers decreased in both uppermost and curved surfaces with the increasing bed height. Once the debris bed is formed, it acts as an obstacle to the natural convective flow, which reduces the buoyancy. The reduction of driving force results in the impaired heat transfer in both upward and downward heat transfers.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.194-194
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• 1998

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.707-712
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• 1997

This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regardt to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard power plant (KSNPP) reactor The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.210-210
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• 1999

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2499-2510
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• 2020

We performed 3-D (3-dimensional) IVR (In-Vessel Retention) natural convection experiments simulating the oxide layer in the three-layer configuration, varying the aspect ratio (H/R). Mass transfer experiment was conducted based on the analogy to achieve high Ra_H's of 1.99 × 10¹²-6.90 × 10¹³ with compact facilities. Comparisons with 2-D (2-dimensional) experiments revealed different local heat transfer characteristics on upper and lower boundaries of the oxide layer depending on the H/R. For the 3-D shallow oxide layer, the multi-cell flow patterns appeared and the number of cells was considerably increased with the H/R decreases, which differs with the 2-D experiments that the number of cells was independent on H/R. Thus, the enhancement of the downward heat transfer and the mitigation of the focusing effect were more noticeable in the 3-D experiments.