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Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

日本におけるナトリウム冷却高速炉の熱流動評価手法と実験研究の進展

上出 英樹; 大島 宏之; 堺 公明; 田中 正暁

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

第4世代原子炉システム国際フォーラムにおいて構築されている安全設計基準では、福島第一原子力発電所事故の教訓や革新技術と関連する研究開発成果を取り入れ、第4世代炉としてのナトリウム冷却高速炉の機器、構造、システムの安全設計にかかる基準を具体化している。この活動には多くの熱流動現象の評価が必要とされる。本論文はこの中から4つの現象として集合体熱流動、自然循環崩壊熱除去、炉心崩壊事故、高サイクル熱疲労を取り上げ、熱流動評価手法の進展を解析コードの検証、実験研究と合わせて示す。これらの手法は高速炉で生じ得る様々な熱流動現象の評価に適用できる包括的な評価手法に統合化され、人的資源の発展や熱流動知見の集約にも役立てられることが期待できる。

In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

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パーセンタイル:64.68

分野:Nuclear Science & Technology

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