• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.169.2-169
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.163.2-163
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.188-188
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.181-181
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• 2003

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.941-952
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• 2017

Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2499-2508
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• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1134-1145
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• 2021

A design parameter study is presented for the closed-loop type passive containment cooling system (PCCS) which is equipped with two heat exchangers: one installed at the inside of the containment and the other submerged in the water pool at the outside of the containment. A GOTHIC code model for PCCS performance analyses was set up and the design parameters such as the heat exchanger sizes, locations, and water pool tank volumes were analyzed to investigate the feasibility of installing this type of PCCS in PWRs like OPR-1000 being operated in Korea. We identified the size of the circulation loop and heat exchangers as major design parameters affecting the performance of PCCS. The analyses showed that the heat exchangers in the inside of the containment would be more influential on the heat removal capability of PCCS than that installed in the water pool at the outside of the containment. Hence, it was recommended to down-size the heat exchangers in the water pool to optimize PCCS without compromising its performance. Based on the parametric study, it was demonstrated that a closed-loop type PCCS could be designed sufficiently compact for installation in the available space within the containment of PWRs like OPR-1000.

• 대한전기학회논문지
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• 제37권8호
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• pp.579-587
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• 1988

일반적인 한글 자동 인식 방법은 세선화 과정을 통한 문자의 골격 추추르 기본자소의 분리 및 인식과정으로 이루어진다. 그러나 이 방법은 복잡한 세선화 과정이 필요하고 잡음에 민감하여 전처리 과정에서는 많은 처리가 필요하며 인식과정에서는 복잡성을 피할 수 없다. 본 연구에서는 고딕체 한글의 기본자소들이 윤곽선의 방향 성분들로서 표현이 가능함을 보이고, 복잡한 세선화 과정이 필요없는 윤곽선 방향 성분들의 추출 방법을 보이며, 추출된 윤곽선 방향 성분들로 한글 문자를 자동인식하는 방법을 제안하였다. 구성된 시스템은 전처리 과정이 매우 간단하며, 잡음에 민감하지도 않고 한글 문자의 윤곽선 방향 성분들을 매우 빠르게 추출하였다. 패턴이 인식 과정도 문자열 패턴매칭 방법으로 대치되어 매우 빠르고 정확하게 한글 문자를 인식해 내었다. 인식율은 92%정도 되었다.

• Nuclear Engineering and Technology
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• 제36권1호
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• pp.24-35
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• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management