• 대한기계학회논문집
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• 제10권1호
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• pp.7-14
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• 1986

본 연구에서는 가압경수형 원자로심을 모의하는 3*3배열로 된 모의연료 봉다발의 실험장치를 이용하여 재완수과정의 유동특성과 열전달특성을 파악하였으며, 재완수과정중 연료봉의 온도거동을 예측하는 REFLUX코드를 최근 개발된 연구자료를 토대로 수정하여 본실험결과와 비교하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.257.1-257
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.243-244
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• 2004

• Nuclear Engineering and Technology
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• 제12권3호
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• pp.153-162
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• 1980

냉각재상실사고의 재관수 단계중 연료봉 피복재의 온도거동 및 열전달 기구를 파악하는 것은 비상노심냉각계통 및 원자로의 안전성해석에 중요하다. 냉각재유동채널의 방위가 rewetting과정에 미치는 영향을 연구하기 위하여 수직 및 수경 유동채널을 이용한 실험을 수행하였으며, 노심이 수평압력관으로 구성되어 있는 CANDU원자로에 관한 실험을 중점적으로 수행하여 그 결과를 수직채널의 결과와 비교 하였다. 또한 rewetting현상을 육안관찰가기 위해 환상형 테스트부 및 외부에서 가열되는 석영관을 사용하였다. 실험결과로써 수평채널에서의 rewetting 속도는 유동의 층상 현상에 크게 영향을 받으나 그 평균값은 수직채널리 경우와 큰차이없음을 알 수 있었다.

• 대한기계학회논문집
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• 제8권1호
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• pp.48-56
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• 1984

In order to estimate the fuel rod temperature during the reflooding phase of the PWR LOCA, it is essential to obtain a better understanding of the rewetting mechanism. This paper presents the results of analytical and experimental investigations aimed at analyzing the rewetting phenomena in a heated tube. A two-dimensional solution of the rewetting for a tube is described and used to yield the correlation of the rewetting heat transfer coefficient as the function of flooding rate and inlet subcooling. This correlation prediction is in good agreement with the experimental data.

• 대한기계학회논문집
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• 제9권4호
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• pp.463-470
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• 1985

본 논문에서는 가열관을 이용한 실험과 고온표면위에 놓인 물방울의 증발실험 을 통하여 재수착온도에 미치는 영향인자들을 분석하였으며, 이러한 분석결과를 토대 로 재수착온도상관식을 제시하였다.

• Journal of Mechanical Science and Technology
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• 제18권6호
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• pp.990-1001
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• 2004

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a Pressurized Water Reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3154-3165
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• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.532-548
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• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• Journal of Advanced Marine Engineering and Technology
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• 제10권2호
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• pp.156-164
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• 1986

In the case of boiling on high temperature wall, vapor film covers fully or parcially the surface. This phenomenon, film boiling or transition boiling, is very important in the surface heat treatment of metal, design of cryogenic heat exchanger and emergency cooling of nuclear reactor. Mainly supposed hydraulic-thermal accidents in nuclear reactor are LCCA (Loss of Coolant Accident) and PCM (Power-Cooling Mismatch). Recently, world-wide studies on reflooding of high temperature rod bundles after the occurrence of the above accidents focus attention on wall temperature history and required time in transient cooling process, wall superheat at rewet point, heat flux-wall superheat relationship beyond the transition boiling region, and two-phase flow state near the surface. It is considered that the further systematical study in this field will be in need in spite of the previous results in ref. (2), (3), (4). The paper is the study about the fast transient cooling process following the wall temperature excursion under the CHF (Critical Heat Flux) condition in a forced convective subcooled boiling system. The test section is a vertically arranged concentric annulus of 800 mm long and 10 mm hydraulic diameter. The inner tube, SUS 304 of 400 mm long, 8 mm I.D, and 7 mm O.D., is heated uniformly by the low voltage AC power. The wall temperature measurements were performed at the axial distance from the inlet of the heating tube, z=390 mm. 6 chromel- alumel thermocouples of 76 .mu.m were press fitted to the inner surface of the heating tube periphery. To investigate the heat transfer characteristics during the fast transient cooling process, the outer surface (fluid side) temperature and the surface heat flux are computed from the measured inner surface temperature history by means of a numerical method for inverse problems of transient heat conduction. Present cooling (boiling) curve is sufficiently compared with the previous results.