• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3930-3943
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• 2021

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper (Kawada and Suzuki, 2020) [1]. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, based on the analysis of the experimental data, the behaviors related to the stub motion were evaluated and quantified by the author from scratch. A simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.

• Nuclear Engineering and Technology
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• 제56권9호
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• pp.3864-3871
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• 2024

In case of a hypothetical severe accident in a Sodium-cooled Fast Reactor (SFR), coolability of the debris bed in the post-accident phase plays a vital role in mitigating the accident and ensuring the structural integrity of the reactor vessel. Few numerical studies are reported in literature, in which the boiling heat transfer in debris bed is expressed as equivalent heat conduction using similarity law between heat conduction and two-phase heat transfer. However, these studies assumed steady state mass conservation for the boiling zone and neglected the gravity force. Hence, a detailed study has been carried out for various particle sizes and porosities of SFR debris to investigate the influence of above considerations. The effect of gravity on debris bed coolability is studied using steady state model of Lipinski, which showed that gravity has a non-negligible effect, for particle size of 0.3 mm and porosity of 0.5. However, the gravitation force was found to have a negligible effect in dryout heat flux estimation for the bottom cooled configuration. A transient numerical model is developed for simulating the boiling phenomena in debris beds and validated with the published experimental results. The assumption of steady state mass conservation is verified by carrying out transient analysis, which indicated early prediction of the dryout inception. For time dependent heat generation case, the unsteady mass conservation predicted higher DHF compared to constant heat generation.

• 한국전산유체공학회지
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• 제20권4호
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• pp.14-20
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• 2015

Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.

• Nuclear Engineering and Technology
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• 제33권2호
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• pp.241-253
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• 2001

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

• 한국가시화정보학회지
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• 제15권3호
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• pp.20-26
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• 2017

The two-fluid equations are widely used to simulate two-phase flows in a nuclear reactor. For the two-fluid momentum equation, the wall and interfacial drag terms play an important role in predicting a two-phase flow behavior. Since the bubble density is much smaller than the water density, the bubble accelerates faster than the liquid in a nozzle. As a result, the bubble phase becomes faster than the liquid phase in the nozzle. In contrast, the opposite phenomena occur in the diffuser. The purpose of our study is to experimentally show these behaviors in an area-varying channel such as nozzle and diffuser. Experiments were made of turbulent bubbly flows in an area-varying horizontal channel. The velocities of the bubble and liquid phases were measured by the PIV technique. It was shown that the two-phase velocities were no longer close to each other in the area-varying regions. The bubble was faster than the liquid in the nozzle; in contrast, the bubble was slower than the liquid in the diffuser. Code simulations were also performed using the MARS code. By replacing the original wall drag model in the MARS code with Kim (1)'s wall drag partition model, we obtained the simulation results being consistent with experimental observations.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.88-103
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• 1999

The purposes of this study are to understand the severe accident phenomena, to establish the simulation method for the experimental test, and to assess the current models in MELCOR for future improvement. This paper presents the results of the PHEBUS FPT1 post test analysis using MELCOR computer code, version 1.8.4. The entire PHEBUS facility has been modeled; the core, the primary circuit including the steam generator, and the containment vessel. Both the thermal hydraulic and the fission product behavior have been investigated. The code simulation results of the thermal hydraulic behavior show good agreement with the experimental data, The fission product release and transport are calculated using the CORSOR models in MELCOR code and the results will be compared with the experiment when the experimental data are available.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(4)
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• pp.51-56
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• 1996

In a NNP (Nuclear Power Plant) severe accident, radionuclides are dispersed into the air. The official regulatory institute, KINS (Korea Institute of Nuclear Safety), has been authorized and developing Computerized technical Advisory system for the Radiological Emergency preparedness (CARE). In this paper, in line with the CARE system, we presented the result of a modularized intermediate-level emergency dose assessment computer code. The RADCON (RADiological CONsequence analysis) version 3.0, which is operable on PC, is developed for simulating emergency situation by considering continuous washout phenomena, and provide a function of effective emergency planning. The source files are coded by using C language in order to increase the compatibility with the other computer system and modularized to adjust the functions and characteristics of each module fer easy understanding and further modification.

• Nuclear Engineering and Technology
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• 제14권4호
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• pp.186-195
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• 1982

SYSRAN code를 사용하여 고리 1호기의 중기배관파열사고를 분석하였다. SYSRAN code는 중성자출력과 열선속계산은 각각 점근사 중성자 운동방정식과 집중정수 모형을 이용하고 냉각수 계통 과도현상에 대해서는 전 계통을 균일한 압력으로 취급하여 질량 및 에너지 평형방정식을 이용하여 계산한다. 사고 결과를 심각하게 만드는 노심상태로 부냉각재 온도계수가 커지는 노심말기와 증기발생기의 유체함량이 가장 많은 고온 정지상태를 호기조건으로 하여, 격납용기외부의 가장 큰 배관면적인 1.4f $t^2$ 크기의 증기배관이 파열되었을때 Moody critical flow model에 따라 증기가 방출된다고 가정하여 분석하였다. 그 결과 노심의 최대 열선속은 사고후 60초에 정상상대의 38%로서 FSAR의 26%에 비해 높은 값을 나타냈으나 모든 과도현상의 경향은 FSAR의 결과와 잘 일치하였다. 민감도 조사결과 이 사고는 냉각재밀도 계수와 노심 하부공간혼합인자에 가장 민감한 것으로 나타났다. B bank중 한 개의 RCCA가 완전인출 상태에서 노심에 삽입되지 않았다고 가정했을 경우의 FSAR 분석결과인 $F_{$\Delta$H}$를 3.66으로 Fz를 1.55로 하여 DNBR을 계산해 본 결과, 최소 DNBR은 1.62가 되어 핵연료의 손상은 예상되지 않았다. 점근사중성자 운동방정식, 집중 정수모형 및 질량과 에너지평형 방정식을 이용한 계통 과도 현상모델은 발전소 전 계통의 과도 현상의 경향을 연구하는데 적합한 것으로 밝혀졌다.구하는데 적합한 것으로 밝혀졌다.

• Journal of Preventive Medicine and Public Health
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• 제23권1호
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• pp.98-105
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• 1990

교통사고와 기상요소와의 관련성을 보기 위하여 대구지방에서 1988년도 1월 1일부터 12월 31일까지 1년간 교통사고를 일으킨 총 운전자중 약 1/10인 2,562명을 대상으로 하여 대구측후소에서 측정한 기상요소를 중심으로 그 관련성을 관찰하였다. 사고양상은 사고당시 인적피해와 물적피해로 구분하고 인적피해와 물적피해가 중복된 경우는 인적피해로 간주하였다. 인적피해는 다시 사망, 중상, 경상 등으로 구분하였으며 두가지 이상에 해당될 때는 피해가 큰 쪽으로 결정하였다. 계절별로 보면 겨울철이 1일평균 사고건수도 적고 인적피해 사고의 비율도 가장 낮았으며 계절에 따른 사고양상은 통계학적으로도 유익한 차이가 있었다. 기온에 다른 시간당 사고건수는 $25.1-30.0^{\circ}C$에서 가장 많았고 인적피해 사고의 비율은 41-50%일 때 가장 높았으나 습도에 따른 사고양상은 통계학적으로 유의하지 않았다. 풍속에 따른 시간당 사고건수는 풍속이 강할수록 많아지는 경향을 보였고 인적피해 사고의 비율은 6.1-7.0m/sec일 때 가장 높은 비율을 보였으나 풍속에 따른 사고양상은 통계학적으로 유의하지 않았다. 강수에 따른 시간당 사고건수는 5.1-10.0mm일 때 5.4건으로 가장 많았고 인적피해 사고이 비율은 강수량이 많을수록 높아지는 경향을 보였으며 강수량과 사고양상 간에는 통계학적으로도 유의한 관련성이 있었다. 시정에 따른 시간당 사고건수는 시정 6 Km미만일 때 가장 많았고 인적피해사고의 비율은 6 Km미만일 때 가장 낮았으며 시정과 사고양상 사이에는 통계학적으로 유의한 관련성이 있었다. 적설이 있을 때 시간당 사고건수는 적설이 없을 때보다 많았고 인적피해사고의 비율은 비교적 적었으나 적설유무와 사고양상과는 통계학적으로 유의하지 않았다. 이상의 결과로 보아 자동차 사고는 기온, 습도, 풍속 등 몇가지 기상조건에 따라 서로 그 발생양상에 크고 작은 차이를 보이고 있는데 계절, 강수량 빛 시정에 따른 사고발생 분포의 차이는 통계적으로도 유의했다.

• Nuclear Engineering and Technology
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• 제37권3호
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• pp.265-272
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• 2005

Severe accidents in nuclear power plants can cause hydrogen-generating chemical reactions, which create the danger of hydrogen combustion and thus threaten containment integrity. For containment analyses, a three-dimensional mechanistic code, GOTHIC-3D has been applied near source compartments to predict whether or not highly reactive gas mixtures can form during an accident with the hydrogen mitigation system working. To assess the code applicability to hydrogen combustion analysis, this paper presents the numerical calculation results of GOTHIC-3D for various hydrogen combustion experiments, including FLAME, LSVCTF, and SNU-2D. In this study, a technical base for the modeling oflarge- and small-scale facilities was introduced through sensitivity studies on cell size and bum modeling parameters. Use of a turbulent bum option of the eddy dissipation concept enabled scale-free applications. Lowering the bum parameter values for the flame thickness and the bum temperature limit resulted in a larger flame velocity. When applied to hydrogen combustion analysis, this study revealed that the GOTHIC-3D code is generally able to predict the combustion phenomena with its default bum modeling parameters for large-scale facilities. However, the code needs further modifications of its bum modeling parameters to be applied to either small-scale facilities or extremely fast transients.