Development of Level 2 PSA Methodology for Sodium-cooled Fast Reactors (6) Development of Technical Basis in Ex-Vessel Accident Sequences
In order to construct and quantify containment event trees in a probabilistic safety assessment (Level 2 PSA), technical basis on important phenomena should be compiled as available information. This paper describes the development of the technical basis for evaluating ex-vessel accident sequences of sodium-cooled fast reactors, which are discussed in the case that a core damage accident might affect to the outside of a reactor vessel. When considering the characteristics of the Japan Sodium-cooled Fast Reactor (JSFR), important behaviors to be investigated are sodium or its vapor leak and fire, sodium-concrete reaction, hydrogen generation and combustion, debris-bed cooling, debris-concrete interaction, etc. The paper summarizes the dominant factors in these behaviors which are identified from parametric analyses with the CONTAIN/LMR code and/or with related individual computational models in the code. The dominant factors are sodium/vapor leak rate, hydrogen generation rate in the sodium-concrete reaction, and morphologies of debris-bed such as debris particle diameter and debris-bed porosity. The description in the paper goes to the compilation of existing available information for selecting a reliable value of a dominant factor and for doing appropriate evaluation of phenomena. The analyses indicate further the relation between dominant factors and accident consequences; I.e. Pressure loading value to a containment vessel and its timing, so that the analytical information can be utilized to quantify the containment event trees.
Level 2 PSA Containment vessel FBR Event trees Sodium leak and fire CONTAIN/LMR
Shuji Ohno Hiroshi Seino Shinya Miyahara
Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki, 311-1393, Japan Japan Atomic Energy Agency, Shiraki, Tsuruga, Fukui 919-1279, Japan；Research Institute of Nuclear En
国际会议
上海
英文
900-911
2010-10-10（万方平台首次上网日期，不代表论文的发表时间）