• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.496-501
• /
• 1998

최근 국내의 PWR 발전소에서는 유체유발진동에 의한 핵연료의 Fretting Wear가 많이 발생하였다. 이는 Baffle Jetting이나 그 밖의 요인도 있을 수 있으나 핵연료의 장주기화, 높은 열적여유도등의 설계요건을 만족하기 위한 노심 내의 유동조건 변화에 기인한다. 특히 고리 2호기에서 발생한 핵연료 손상 중 15%정도가 유체유발진동으로 추정되고 있다. 따라서 본 연구는 손상 핵연료의 노심내 위치, 부위, 유동조건 등으로 부터 유체유발진동의 주요 손상 원인을 규명하는데 있다. 이를 위해 핵연료 집합체에서 발생할 수 있는 유체유발진동 메카니즘의 특징과 유동조건, 손상 핵연료의 노심내 위치, 파손 부위, 집합체와 지지격자의 기하학적 형태를 고려한 유동 방향 등을 연관 분석 결과 파손을 일으키는 주요원인을 단일 집합체 내에서 발생되는 Vortex Shedding과 인접한 집합체 사이에서 발생되는 Fluidelastic Instability의 중복효과로 규명하였다 또한 최근 핵연료 설계에 도입된 Mixing Vane의 효과가 과도하여 핵연료 손상을 일으키는 가설을 정립하였다.

• Nuclear Engineering and Technology
• /
• v.23 no.3
• /
• pp.340-351
• /
• 1991

Experimental studies were conducted on a l/5.03 scale reactor flow model of the Yong-gwang Nuclear Units 3 and 4. The purpose of the flow model test was to estimate the hydraulic effect in the reactor vessel due to the relative size difference between the ABB-CE's System 80 and the YGN 3&4 reactors. The flow model was designed according to the principle of similarity. Obtained from the test were the core inlet flow distribution, the core exit pressure deviations, and the segmental and overall pressure losses across the flow path from the reactor vessel inlet to outlet nozzle. These data will be used to provide input data for the core thermal margin analysis and to verify the analytical hydraulic design method.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.38 no.3
• /
• pp.271-277
• /
• 2014

Internal structures, especially those located in the upstream of a reactor core, may have a significant influence on the core inlet flow rate distribution depending on both their shapes and the relative distance between the internal structures and the core inlet. In this study, to examine the effect of the reactor internal structure geometry treatment method on the prediction accuracy for the scale-down APR+ flow distribution, simulations with real geometry modeling were conducted using ANSYS CFX R.14, a commercial computational fluid dynamics software, and the predicted results were compared with those of the porous medium assumption. It was concluded that the core inlet flow distribution could be predicted more accurately by considering the real geometry of the internal structures located in the upstream of the core inlet. Therefore, if sufficient computational resources are available, an exact representation of these internal structures, for example, lower support structure bottom plate and ICI nozzle support plate, is needed for the accurate simulation of the reactor internal flow.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.35 no.4
• /
• pp.425-429
• /
• 2011

In the unlikely of nuclear reactor meltdown, the leaked core melt or corium must be contained in a device called core-catcher so that the corium can be cooled and stabilized. The ex-vessel behavior of corium involves complex physical and chemical mechanisms of flow propagation, heat transfer, and reactions with sacrificial substrates. In this study, the detailed characteristics of corium flow and heat transfer were investigated by using a commercial CFD code for VULCANO VE-U7 test reported in the literature. The volume-of-fluid (VOF) model was used to predict the interfacial surface formation of corium and the surrounding air, and the discrete ordinate model was adopted to calculate radiation between corium and the surroundings. It was found that cooling via radiation through the top surface of corium had a dominant effect on the temperature and viscosity profiles at the front of the corium flow.

• Nuclear Engineering and Technology
• /
• v.25 no.2
• /
• pp.292-299
• /
• 1993

원자로 노심내 핵 연료봉의 정확한 온도 분포를 구하기 위해서는 핵 연료봉 주위의 난류 유동장에 대한 해석이 필요하다. 난류 유동장의 특성을 파악하기 위해 실험적 연구와 해석적 방법에 의한 연구가 함께 수행되고 있다. 본 기고문은 현재까지 보고된 난류 유동장의 특성을 알아보고 해석적 방법의 연구 동향과 문제점을 분석하여 이 분야에 대한 연구 활동에 도움이 되고자 한다.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.37 no.9
• /
• pp.855-862
• /
• 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.

• 한국전산유체공학회:학술대회논문집
• /
• 2004.10a
• /
• pp.151-154
• /
• 2004

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. A flow simulated test facility has been developed for the verification of structural integrity of those experimental facilities prior to loading In the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The half-core structure assembly is composed of plenum, grid plate, core channel with flow tubes, chimney and dummy pool. The flow channels are to be filled with flow orifices to simulate similar flow characteristics to the HANARO. This paper describes an analysis of the flow distribution of the cote channel and compares with the test results. As results, the analysis showed similar flow characteristics compared with those in the test results.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.05a
• /
• pp.557-562
• /
• 1997

노심용융물의 노내 자연 냉각 현상은 TMI-2 사고 이래로 실험과 해석 분야에서 많은 연구가 이루어지고 있으나, 아직까지는 이에 대한 명확한 규명이 이루어지지 않은 상태이다. 원자로 용기 냉각 Mechanism 중에서 노심용융물이 원자로 용기 하부 반구내로 재배치되어 하부 반구 내벽과 접촉할 때 용융물과 하부 반구 내벽 사이에 생길 수 있는 작은 간극으로 냉각수가 침투되어 노심용융물의 냉각이 이루어질 수 있다는 가정이 유력하게 제기되고 있다 본 논문에서는 노심용융물과 원자로 용기 하부 반구 사이의 간극을 통한 노심용융물의 냉각 특성을 규명하는 SONATA-IV실험 연구와 연계하여 이상 유동이 존재하는 고온 표면에서의 미세한 간극을 측정할 수 있는 방법의 검토 및 시편을 이용한 실험을 통하여 가장 적합한 간극측정기법을 도출하였다 간극 측정 기법으로는 중성자 래디오그라피, X 선 후방산란 단층기법 그리고 초음파 펄스 반사 탐상법을 검토하였으며, 시편 측정 실험결과 실시간 간극 측정방법으로는 초음파 펄스 반사 탐상법이 가장 적합한 것으로 나타났다.

• Journal of Energy Engineering
• /
• v.28 no.4
• /
• pp.103-110
• /
• 2019

The domestic innovative power reactor named iPOWER will employ the passive molten corium cooling system(PMCCS) to cool down and stabilize the core melt in the severe accident. The final design concept of the PMCCS is yet to be determined, but the in-vessel retention through external reactor vessel cooling has been also considered as a viable strategy to cope with the severe accident. In this study, the two-phase natural circulation flow established between the reactor vessel and the insulation was simulated using a thermal-hydraulic system code, MARS-KS. The flow path of cooling water was modeled with one-dimensional nodes, and the boundary condition of the heat load from the molten core was defined to estimate the naturally-driven flow rate. The evolution of major thermal-hydraulic parameters were also evaluated, including the temperature and the level of cooling water, the void fraction around the lower head of the reactor vessel, and the heat transfer mode on its external surface.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.39 no.11
• /
• pp.905-911
• /
• 2015

In this study, to examine the effect of a hole size change(smaller hole diameter) in the outer region of the lower-support-structure-bottom plate(LSSBP) on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.15. The predicted results were compared with those of the original LSSBP. Through these comparisons, it was concluded that a more uniform distribution of the mass flow rate at the core-inlet plane could be obtained by reducing the hole size in the outer region of the LSSBP. Therefore, from the nuclear regulatory perspective, design change of the hole pattern in the outer region of the LSSBP may be desirable in terms of improving both the mechanical integrity of the fuel assembly and the core thermal margin.