核燃料施設の事故影響評価手法に関する調査（Ⅱ）: 溶液沸騰事象での放射性物質の移行割合に関する基礎的データと試解析

摘要

溶液沸騰事象に関するデータの調査として，事故解析ハrnンドブックに示された溶液沸騰時のARFの推奨値の根拠rnとなった実験を調査した。その結果，沸騰を伴う実験でrnは，フィルタに吸着したPu量をARF値としているが，rnこの実験体系では上昇空気流量により吸着量が影響を受けrnるだけでなく，沸騰液面とエアロゾルを回収するフィルタrnが近接しているため，Pu量の測定値には，重力沈降で落rn下しエアロゾル化しない比較的大きな液摘も含まれる可能rn性があり，これのようなデータに基づくハンドフ、ソクのrnARFの推奨値は過大であることがわかった。%A special committee on "Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)" was organized by the Atomic Energy Society of Japan under the entrustment of Japan Atomic Energy Agency for research on the state-of-the-art consequence analysis method for Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objective of this research is to obtain the basic useful information related to the establishment of the quantitative performance requirement and to risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on accidents with more severe consequences than design basis, such as events of criticality, explosion, fire, and boiling of radioactive solution postulated in NFFs resulting in the release of radioactive materials into the environment. The research results are summarized in this technical report about basic experimental data related to key physical and chemical phenomena postulated in a boiling event of a radioactive solution storage tank caused by the loss of the cooling function.