Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor

被引：23

作者：

Wu, Y. W. ^{[1
]}

Li, Xin ^{[1
]}

Yu, Xiaolei ^{[1
]}

Qiu, S. Z. ^{[1
]}

Su, G. H. ^{[1
]}

Tian, W. X. ^{[1
]}

机构：

[1] Xi An Jiao Tong Univ, Sch Nucl Sci & Technol, State Key Lab Multiphase Flow Power Engn, Xian 710049, Peoples R China

来源：

PROGRESS IN NUCLEAR ENERGY | 2013年 / 68卷

关键词：

Sodium-cooled fast reactor; Subchannel code; Thermal hydraulics; FLOW PRESSURE-DROP; SPACED ROD BUNDLES; HEAT-TRANSFER; MATRA-LMR; CODE; METALS; ELEMENTS; MODEL;

D O I：

10.1016/j.pnucene.2013.05.001

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

Reasonable mathematical and physical models as well as auxiliary models have been established to develop a subchannel analysis code for one fuel assembly of the Sodium-cooled Fast Reactor (SFR). The conduction model of mixed fuel UO2-PuO2 was adopted in the sub-channel analysis. The comparison of the flow velocity distribution in the fuel assembly was performed between the Chiu-Rohsenow-Todreas (CRT) and Novendstern models. Heat transfer correlations for liquid metals were compared with each other and one was selected as the optimization correlation. The validation of the code was performed with Oak Ridge National Laboratory (ORNL) 19 pin tests. The temperature profiles at the end of the heated length for low and high power cases were compared between experimental results and other codes. And then, based on the subchannel code, thermal-hydraulic characteristics of the Chinese Experimental Fast Reactor (CEFR) were investigated. Axial and radial coolant temperature profiles for different subchannels were presented. In addition, the mass flow rate with mixing effects were carefully studied. The effect of the wire was investigated and the optimization ratio of the pitch to diameter was provided according to current simulated conditions. (C) 2013 Elsevier Ltd. All rights reserved.

引用

页码：65 / 78

页数：14

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