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http://dx.doi.org/10.6112/kscfe.2011.16.3.095

ASSESSMENT of CORE BYPASS FLOW IN A PRISMATIC VERY HIGH TEMPERATURE REACTOR BY USING MULTI-BLOCK EXPERIMENT and CFD ANALYSIS  

Yoon, S.J. (서울대학교 대학원 원자핵공학과)
Lee, J.H. (서울대학교 대학원 원자핵공학과)
Kim, M.H. (한국원자력연구원 수소생산원자로기술개발부)
Park, G.C. (서울대학교 원자핵공학과)
Publication Information
Journal of computational fluids engineering / v.16, no.3, 2011 , pp. 95-103 More about this Journal
Abstract
In the block type VHTR core, there are inevitable gaps among core blocks for the installation and refueling of the fuel blocks. These gaps are called bypass gap and the bypass flow is defined as a coolant flows through the bypass gap. Distribution of core bypass flow varies according to the reactor operation since the graphite core blocks are deformed by the fast neutron irradiation and thermal expansion. Furthermore, the cross-flow through an interfacial gap between the stacked blocks causes flow mixing between the coolant holes and bypass gap, so that complicated flow distribution occurs in the core. Since the bypass flow affects core thermal margin and reactor efficiency, accurate prediction and evaluation of the core bypass flow are very important. In this regard, experimental and computational studies were carried out to evaluate the core bypass flow distribution. A multi-block experimental apparatus was constructed to measure flow and pressure distribution. Multi-block effect such as cross flow phenomenon was investigated in the experiment. The experimental data were used to validate a CFD model foranalysis of bypass flow characteristics in detail.
Keywords
Very High Temperature Reactor; Block-type Core; Core Bypass Flow; Bypass Gap; Crossflow Gap; Multi-block Experiment; CFD;
Citations & Related Records
Times Cited By KSCI : 1  (Citation Analysis)
연도 인용수 순위
1 2009, 윤수종 외3인, "단위-셀 실험과 전산유체해석을 통한 블록형 초고온가스로의 노심우회유량 평가," KSCFE, Vol.14, No.2, pp.59-67.
2 2009, ANSYS, ANSYS CFX-Solver Thoery Guide, ANSYS Inc., Canonsburg, PA.
3 2004, Fluent Inc., GAMBIT 2.2 User's Guide, Fluent Inc., Lebanon, NH.
4 1993, Menter, F.R., "Zonal Two Equation k-w Turbulence Models For Aerodynamic Flow," AIAA 93-2906.
5 1997, Bardina, J.E., et al., "Turbulence Modeling Validation," AIAA 97-2121.
6 1996, Idelchik, I.E., "Handbook of Hydraulic Resistance 3rd edition," Begell house Inc., New York, NY.
7 2008, KAERI, "PMR 200MWth의 GAMMA+ 코드 입력자료 작성 및 정상 상태 열수력 해석," NHDD-KA-08-RDCA-002, KAERI.
8 1994, Burchell, T.D. et al., "The Effect of Neutron Irradiation on the Structure and Properties of Carbon-Carbon Composite Materials," Journal of Nuclear Materials, Vol.191-194, pp.295-299.
9 2004, Independent Technology Review Group, "Design Features and Technology Uncertainties for the Next Generation Nuclear Plant," INEEL/EXT-04-01816, INL.
10 2005, MacDonald, P.E., "Next Generation Nuclear Plant Research and Development Program Plan," INEEL/EXT-05-02581, INL.