Thermal-hydraulic characteristics of reacting zone for TWR bundles based on CFD method

被引:0
作者
Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu 610041, China [1 ]
机构
[1] Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China
来源
Lu, C. | 1600年 / Atomic Energy Press卷 / 47期
关键词
Bundle; CFD; Temperature; Travelling wave reactor;
D O I
10.7538/yzk.2013.47.12.2244
中图分类号
学科分类号
摘要
Thermal-hydraulic characteristics of reacting zone for TWR (travelling wave reactor) bundles were analysed by CFD method. The calculation results of 7, 19 and 37 fuel pin bundles show the similar characteristics. The hot coolant seems to congregate into the centre as flowing to the downstream area. The high temperature coolant always distributes in the inner area while the temperature shows distinct gradation in the outer area. The temperature difference is more than 100 °C for the bundle whose diameter is about 26 cm. The major temperature gradations mainly locate in the outermost fuel rods of two circles while other circles show much smaller temperature gradients. This conclusion is estimated to be true for more fuel pin bundles such as 217 fuel pin bundles. The fuel assembly structure of the existing TWR design should be optimized in future.
引用
收藏
页码:2244 / 2248
页数:4
相关论文
共 11 条
[1]  
Yan M.Y., Sekimoto H., Design research of small long life CANDLE fast reactor, Ann Nucl Energy, 35, pp. 18-36, (2008)
[2]  
Yan M.Y., Sekimoto H., Safety analysis of small long life CANDLE fast reactor, Ann Nucl Energy, 35, pp. 813-828, (2008)
[3]  
Weaver K., Ahlfeld C., Gilleland J., Et al., Extending the nuclear fuel cycle with traveling-wave reactors, Proceedings of Global, (2009)
[4]  
Gajapathy R., Velusamy K., Selvaraj P., CFD investigation of helical wire-wrapped 7-pin fuel bundle and the challenges in modeling full scale 217 pin bundle, Nuclear Engineering and Design, 237, pp. 2332-2342, (2007)
[5]  
Liu Y., Yu H., Numerical simulation of flow and temperature field of fuel subassembly for China Experimental Fast Reactor, Atomic Energy Science and Technology, 41, SUPPL., pp. 230-234, (2007)
[6]  
Gajapathy R., Velusamy K., Selvaraj P., A comparative CFD investigation of helical wire-wrapped 7, 19 and 37 fuel pin bundles and its extendibility to 217 pin bundle, Nuclear Engineering and Design, 239, pp. 2279-2292, (2009)
[7]  
Bieder U., Ducros F., Fauchet G., Et al., CFD investigations of a full scale helical wire-wrapped 61-pin fuel bundle by using the code TRIO_U, Annual Meeting on Nuclear Technology 2009, (2009)
[8]  
Bieder U., Barthel V., Ducros F., Et al., CFD calculations of wire wrapped fuel bundles: Modelling and validation strategies, CFD for Nuclear Reactor Safety Applications Workshop 2010 (CFD4NRS-3), (2010)
[9]  
Hamman K.D., Berry R.A., A CFD simulation process for fast reactor fuel assemblies, Nuclear Engineering and Design, 240, pp. 2304-2312, (2010)
[10]  
Gou J., Shanga Z., Ishiwararib Y., CFD analysis of heat transfer in subchannels of a super fast reactor, Nuclear Engineering and Design, 240, pp. 1819-1829, (2010)