タンク型ナトリウム冷却高速炉における崩壊熱除去特性に係る安全評価

Safety evaluation of decay heat removal characteristics in a pool-type sodium-cooled fast reactor

時崎 美奈子*; 谷 明洋*; 安藤 将人*; 小野田 雄一 

Tokizaki, Minako*; Tani, Akihiro*; Ando, Masato*; Onoda, Yuichi

タンク型ナトリウム冷却高速炉(600MW[e]級出力)を対象に、運転時の異常な過渡変化(AOO)及び設計基準事故(DBA)の範疇を対象とした崩壊熱除去特性評価を行い、安全性の判断基準に照らして炉心燃料及び原子炉冷却材バウンダリの健全性を確保できる見通しを得た。また、崩壊熱除去機能に係るロバスト性確認の観点から、浸漬型DRACS単独運転時の崩壊熱除去特性評価を行い、自然循環による炉心冷却の成立見通しを得た。

Decay heat removal characteristics were evaluated for the Pool-Type Sodium-Cooled Fast Reactor (600 MWe class power) under the categories of Anticipated Operational Occurrences (AOO) and Design Basis Accidents (DBA), and it is expected that the integrity of the core fuel and coolant boundary can be ensured in light of the safety criteria. In addition, from the viewpoint of robustness confirmation of decay heat removal function, decay heat removal characteristics of immersion type DRACS during independent operation were evaluated, and the prospect of core cooling by natural circulation was obtained.