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Transient behavior of multi-dimensional core cooling by D-DHX in sodium-cooled fast reactors

ナトリウム冷却高速炉における浸漬型DHXによる多次元的炉心冷却過渡挙動

江連 俊樹 ; 秋元 雄太 ; 小野島 貴光; 栗原 成計 ; 田中 正暁  

Ezure, Toshiki; Akimoto, Yuta; Onojima, Takamitsu; Kurihara, Akikazu; Tanaka, Masaaki

浸漬直接熱交換器による炉心崩壊熱除去中の熱流動挙動を把握するため、ナトリウム試験装置PLANDTL-2を用いた試験研究を実施した。PLANDTL-2の模擬炉心は55体の六角流路管で構成され、複数列の燃料集合体を有する炉心領域の冷却挙動を炉心径方向の熱流動挙動を含めて把握することが可能である。本研究では、浸漬直接熱交換器を用いた炉心崩壊熱除去について、炉心スクラム後を対象とする試験を実施し、模擬炉心領域を含む系統全体の温度の時間変化を計測した。その結果、模擬炉心領域において過度に歪んだ径方向温度分布が生じないことを確認した。

In order to grasp the thermal-hydraulic behaviors during decay heat removal by dipped-direct heat exchangers (D-DHXs) in a sodium-cooled fast reactor, an experimental study was performed using a sodium experimental facility. The simulated core of PLANDTL-2 was formed by 55 hexagonal-shaped flow channel tubes, which allows to examine the cooling behavior of the simulated core region having multiple rows of fuel assemblies including the thermal hydraulic behavior to the radial direction. In this study, transient core cooling behavior in the situation after the scram with the decay heat removal using a D-DHX was examined. The time evolution of the temperature was measured in the whole system including the simulated core region. As the results, it was confirmed there was not excessively skewed temperature distribution in the radial direction in the core region.

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