CFD ASSESSMENT OF THE LOCAL HOT CORE TEMPERATURE IN A PEBBLE-BED TYPE VERY HIGH TEMPERATURE REACTOR

被引:0
作者
Kim, Min-Hwan [1 ]
Lim, Hong-Sik [1 ]
Lee, Won Jae [1 ]
机构
[1] Korea Atom Energy Res Inst, Taejon 305353, South Korea
来源
ICONE16: PROCEEDING OF THE 16TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING - 2008, VOL 2 | 2008年
关键词
D O I
暂无
中图分类号
X [环境科学、安全科学];
学科分类号
08 ; 0830 ;
摘要
Assessment of the local hot core temperature during normal operation in a pebble-bed type of Very High Temperature Reactor (VHTR) has been carried out by using the Computational Fluid Dynamic (CFD) method for which the boundary conditions were obtained from the results of a macroscopic analysis of the core using a system thermal analysis code, GAMMA. Three pebble arrangements are selected, which are Simple Cubic (SC), Body-Centered Cubic (BCC), and Face-Centered Cubic (FCC). Results showed that the SC arrangement having the lowest porosity gives the highest fuel temperature of 1237 degrees C but still below the normal operational fuel limit of 1250 degrees C. Comparison of the CFD results with an empirical correlation was made for the pressure drop and the Nusselt number but there were large differences between them. The benchmark calculation of a pressure drop for packed particles in a square channel indicated that the correlation for the full core used in the system code is not appropriate for the prediction of a local thermal fluid behavior.
引用
收藏
页码:281 / 287
页数:7
相关论文
共 7 条
[1]  
*ANSYS INC, 2005, CFX REL 10 0 REF MAN
[2]   CFD modelling and experimental validation of pressure drop and flow profile in a novel structured catalytic reactor packing [J].
Calis, HPA ;
Nijenhuis, J ;
Paikert, BC ;
Dautzenberg, FM ;
van den Bleek, CM .
CHEMICAL ENGINEERING SCIENCE, 2001, 56 (04) :1713-1720
[3]  
FRENCH H, 1980, HEAT TRANSFER FLUID, P382
[4]  
*IAEA, 2002, 2 IAEA CFP5 TECDOC, P14
[5]  
In W.K., 2005, T KOR NUCL SOC AUT M
[6]   GAMMA multidimensional multicomponent mixture analysis to predict air ingress phenomena in an HTGR [J].
Lim, HS ;
No, HC .
NUCLEAR SCIENCE AND ENGINEERING, 2006, 152 (01) :87-97
[7]  
RICHARDS M, 2007, T 15 INT C NUCL ENG