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A PRELIMINARY EVALUATION OF UNPROTECTED LOSS-OF-FLOW ACCIDENT FOR A PROTOTYPE FAST-BREEDER REACTOR
TOHRU SUZUKI,Yoshiharu Tobita,KENICHI KAWADA,HIROTAKA TAGAMI,JOJI SOGABE,KENICHI MATSUBA,KEI ITO,HIROYUKI OHSHIMA 한국원자력학회 2015 Nuclear Engineering and Technology Vol.47 No.3
In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of in-vessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.
CHARACTERISTICS OF SELF-LEVELING BEHAVIOR OF DEBRIS BEDS IN A SERIES OF EXPERIMENTS
Cheng, Songbai,Yamano, Hidemasa,Suzuki, TYohru,Tobita, Yoshiharu,Nakamura, Yuya,Zhang, Bin,Matsumoto, Tatsuya,Morita, Koji Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.3
During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.
Characteristics of Self-Leveling Behavior of Debris Beds in a Series of Experiments
Songbai Cheng,Hidemasa Yamano,Tohru Suzuki,Yoshiharu Tobita,Yuya Nakamura,Bin Zhang,Tatsuya Matsumoto,Koju Morita 한국원자력학회 2013 Nuclear Engineering and Technology Vol.45 No.3
During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA)and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.