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

Y. W. Wu, Xin Li, Xiaolei Yu, S. Z. Qiu, G. H. Su, W. X. Tian

研究成果: Article査読

14 被引用数 (Scopus)

抄録

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.

本文言語English
ページ(範囲)65-78
ページ数14
ジャーナルProgress in Nuclear Energy
68
DOI
出版ステータスPublished - 2013
外部発表はい

ASJC Scopus subject areas

  • 原子力エネルギーおよび原子力工学
  • 安全性、リスク、信頼性、品質管理
  • エネルギー工学および電力技術
  • 廃棄物管理と処理

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