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高温ガス炉の圧力容器周りの熱流動解析

Thermal hydraulic analysis on reactor pressure vessel in HTGR

武田 哲明; 西尾 仁志*; 一宮 浩一*

Takeda, Tetsuaki; Nishio, Hitoshi*; Ichimiya, Koichi*

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。熱放射を考慮した解析の結果、圧力容器側の温度は低下し、炉容器冷却システム(VCS)側のコンクリート温度は上昇するものの、構造物全体ではVCSによる冷却が十分なされていることを確認した。

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 the RPV and surrounding concrete structure were cooled enough by the VCS in the case of loss of coolant flow.

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