高温ガス炉の冷却材流量喪失時の炉容器冷却システムに関する解析研究

Analytical study on reactor vessel cooling system during loss of coolant flow in HTGR

武田 哲明; 栃尾 大輔 ; 稲葉 良知  ; 一宮 浩一*; 西尾 仁志*

Takeda, Tetsuaki; Tochio, Daisuke; Inaba, Yoshitomo; Ichimiya, Koichi*; Nishio, Hitoshi*

汎用の熱流動数値解析コードSTAR-CDを用いて、高温工学試験研究炉(HTTR)の一次冷却材流量部分喪失試験時の解析を行い、実測値と比較するとともに、圧力容器と炉容器冷却パネル間の熱放射が圧力容器周りの構造物の温度変化に及ぼす影響を調べた。解析の結果、一次冷却材の流量が喪失した場合の圧力容器内各部温度の解析値は実測値と定量的に一致した。また、遮へい体コンクリート温度評価では、圧力容器壁面から放射熱量の評価が重要であることがわかった。

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The test of loss of coolant flow as one of safety demonstration tests is carried out by tripping of the helium gas circulators. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the vessel cooling system (VCS) during the loss of coolant flow. The temperature distribution of the RPV and surrounding concrete structure were obtained using a commercially available analysis code STAR-CD. The effect of thermal radiation from the RPV was evaluated using the analytical and experimental results. It was found that it is important to take account of the effect of thermal radiation in the transient analysis to evaluate the temperature change of the concrete accurately.