• 열처리공학회지
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• 제18권2호
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• pp.112-118
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• 2005

Study was focused on the effect of the cooling rate on the embrittlement behavior of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment. Claddings were oxidized at given temperature and given time followed by various water quenching in the range of $0.6^{\circ}C$ and $100^{\circ}C$ per second. Cladding failed after water quenching above the threshold oxidation. Threshold oxidation was decreased as the cooling rate increased, which is due to the matensite structure formed during fast cooling rate.

• 에너지공학
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• 제10권4호
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• pp.342-348
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• 2001

A new design concept for a decay heat removal system in a liquid metal reactor is proposed. The new design utilizes a thermosyphon to enhance the heat removal capacity and its heat transfer characteristics are analyzed against the current PSDRS (Passive Safety Decay heat Removal System) in the KAL IMER (Korea Advanced LIquid MEtal Reactor) design. The preliminary analysis results show that the new design with a thermosyphon yields substantial increase of 20∼40% in the decay heat removal capacity compared to the current design that do not have the thermosyphon. The new design reduces the temperature rise in the cooling air of the system and helps the surrounding structure in maintaining its mechanical integrity for long term operation at an accident. Also the analysis revealed the characteristics of the interactions among various heat transfer modes in the new design.

• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.161-162
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• 2004

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.932-940
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• 2016

Pool boiling heat transfer of water saturated at atmospheric pressure was investigated experimentally on Cu surfaces with high-temperature, thermally-conductive, microporous coatings (HTCMC). The coatings were created by sintering Cu powders on Cu surfaces in a nitrogen gas environment. A parametric study of the effects of particle size and coating thickness was conducted using three average particle sizes (APSs) of $10{\mu}m$, $25{\mu}m$, and $67{\mu}m$ and various coating thicknesses. It was found that nucleate boiling heat transfer (NBHT) and critical heat flux (CHF) were enhanced significantly for sintered microporous coatings. This is believed to have resulted from the random porous structures that appear to include reentrant type cavities. The maximum NBHT coefficient was measured to be approximately $400kW/m^2k$ with APS $67{\mu}m$ and $296{\mu}m$ coating thicknesses. This value is approximately eight times higher than that of a plain Cu surface. The maximum CHF observed was $2.1MW/m^2$ at APS $67{\mu}m$ and $428{\mu}m$ coating thicknesses, which is approximately double the CHF of a plain Cu surface. The enhancement of NBHT and CHF appeared to increase as the particle size increased in the tested range. However, two larger particle sizes ($25{\mu}m$ and $67{\mu}m$) showed a similar level of enhancement.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.929-944
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• 2009

The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Light Water Reactor (LWR) lower plenum during hypothetical severe accident progression. The ECM uses heat transfer characteristic velocities to describe turbulent natural convection of a melt pool. The simple approach of the ECM method allows implementing different models of the characteristic velocity in a mushy zone for non-eutectic mixtures. The Phase-change ECM (PECM) was examined using three models of the characteristic velocities in a mushy zone and its performance was compared. The PECM was validated using a dual-tier approach, namely validations against existing experimental data (the SIMECO experiment) and validations against results obtained from Computational Fluid Dynamics (CFD) simulations. The results predicted by the PECM implementing the linear dependency of mushy-zone characteristic velocity on fluid fraction are well agreed with the experimental correlation and CFD simulation results. The PECM was applied to simulation of melt pool formation heat transfer in a Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lower plenum. The study suggests that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1343-1354
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• 2022

Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the in-vessel retention (IVR) phase of a severe accident. Deionized light water was used as the working fluid. Low carbon steel shielding balls with diameters ranging from 6.4 to 12.7 mm were used, allowing for the development of an empirical correlation of CHF as a function of shielding ball diameter. Previously published data is used to develop a more comprehensive empirical correlation accounting for the impacts of both shielding ball diameter and heating surface height. Tests using borosilicate shielding balls demonstrated that the dependence of CHF on shielding ball thermal conductivity is insignificant. The deposition of iron oxide particles transported from shielding balls to the heating surface is verified to increase CHF non-trivially. The results presented in this paper improve the state of the knowledge base permitting quantitative prediction of CHF in the CANDU end shield, refining our ability to assess the feasibility of IVR. The findings clarify the mechanisms governing CHF in this scenario, permitting identification of potential future research directions.

• 에너지공학
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• 제24권1호
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• pp.78-87
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• 2015

중대사고시 금속용융물층의 열속 집중 현상(Focusig effect)에 대해 상부와 측면벽의 냉각 조건과 높이를 변화시키면서 실험과 수치해석을 수행하였다. 상사성(Analogy) 원리를 이용해 열전달 실험 대신 물질전달 실험을 수행하였으며 황산-황산구리 수용액의 전기도금계를 물질전달계로 채택하였다. $Ra_H$ 수 $8.49{\times}10^7{\sim}5.43{\times}10^9$ 범위에서 상부와 측면벽의 냉각 조건을 세 가지로, 높이를 네 가지로 변화시키면서 열전달을 측정하였다. 상부만 냉각인 경우의 실험결과를 동일한 조건인 Rayleigh-Benard 자연대류 상관식과 비교한 바 Dropkin과 Somerscales, Globe와 Dropkin의 상관식과 매우 일치하였다. 측면벽만 냉각인 경우, 상부와 측면벽 모두 냉각인 경우, 상부만 냉각인 경우 순으로 열전달이 감소하였고, 냉각 조건을 고정한 상태에서 높이를 감소시킬수록 측면 열전달이 향상되었다.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.