高速炉シビアアクシデント時の炉内自然対流崩壊熱除去特性の把握

Study on cooling process of decay heat removal systems in a reactor vessel of sodium-cooled fast reactor under severe accident

辻 光世 

Tsuji, Mitsuyo

高速炉シビアアクシデント時の自然対流除熱特性を把握するために、縮尺水流動試験装置を用いてコアキャッチャにデブリが一様に堆積した条件を模擬し、浸漬型DHXを起動した場合の自然対流による炉内流動場をPIV計測した。DHXからの低温流体が炉壁を伝い下降して下部プレナムへ流入しコアキャッチャ上で加熱され炉心中央部へと上昇する流速場や、低温流体の下降流と高温流体の上昇流の相互作用によるうずの形成を確認した。また、既往知見との比較により、同様の流動場や同等の最大流速を確認した。

To elucidate the core cooling systems of a sodium-cooled fast reactor under severe accident, PIV experiments were carried out by using scaled water experimental facility simulating the condition of uniformly-accumulated debris on the core catcher, measured the natural convection flow field in starting up the submerged type DHX. The authors confirmed the specific flow patterns that low temperature fluid flowed down along the core wall and flowed into lower plenum, and the fluid was heated on the core catcher and elevated toward the center of reactor core, then some vortexes were formed by the mutual interaction due to the down-flow of low temperature fluid and up-flow of elevated temperature fluid. Furthermore, the equivalent flow field and the maximum flow velocity were observed compared with the existing knowledge.