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Kimura, Nobuyuki; ; ; ; Kamide, Hideki; Tokuhiro, Akira; Hishida, Koichi
JNC TN9400 2000-057, 60 Pages, 2000/05
ln experimental study for the thermohydraulics of fast reactor, a simple experiment with fine measurement has been desired for understanding of phenomena and for verification of computer code rather than mockup experiments of large scale. For such purposes quality of experimental data must be improved. ln the velocity measurement, instantaneous velocity profile will have great advances for the understanding of phenomena and for the verification of computer code. ln this report two methods of the velocity profile measurement are discussed; one is ultrasound Doppler velocimetry (UDV) and the other is particle image velocimetry (PIV). These methods were applied to water experiments. The UDV was applied to pipe flow, planer jet, and the inter-wrapper flow which is seen in the gap region between subassemblies of fast reactor core. Cross check with laser Doppler velocimetly showed proper measurement of the UDV. Problems including the application to sodium experiments are also discussed. The PIV was also applied to the inter-wrapper flow. For the application to complex flow geometry, noise reduction method was developed to improve the measurement accuracy.
Ono, Ayako; Kurihara, Akikazu; Tanaka, Masaaki; Ohshima, Hiroyuki; Miyake, Yasuhiro*; Ito, Masami*; Nakane, Shigeru*
no journal, ,
Thermal-hydraulic phenomena driven by natural circulation in a reactor vessel was investigated by using scaled model water experiments simulating a reactor vessel in order to enforce of safety and optimize design and operation of decay heat removal systems under normal operation and severe accident conditions. Through the flow visualization tests, the behavior of cold fluid from the dipped-type heat exchanger and cooling process of the core and debris in the vessel were revealed.
Tsuji, Mitsuyo
no journal, ,
To elucidate the core cooling systems of a sodium-cooled fast reactor under severe accident, PIV experiments were carried out by using scaled water experimental facility simulating the condition of uniformly-accumulated debris on the core catcher, measured the natural convection flow field in starting up the submerged type DHX. The authors confirmed the specific flow patterns that low temperature fluid flowed down along the core wall and flowed into lower plenum, and the fluid was heated on the core catcher and elevated toward the center of reactor core, then some vortexes were formed by the mutual interaction due to the down-flow of low temperature fluid and up-flow of elevated temperature fluid. Furthermore, the equivalent flow field and the maximum flow velocity were observed compared with the existing knowledge.
Tsuji, Mitsuyo; Ono, Ayako; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*
no journal, ,
Thermal-hydraulic phenomena driven by natural circulation in a reactor vessel was investigated by using scaled model water experiments simulating a reactor vessel in order to enforce of safety and optimize design and operation of decay heat removal systems under normal operation and severe accident conditions. This paper reports PIV measurement results of natural convection flow field in reactor vessel simulating the condition of uniformly-accumulated debris on the core catcher and operating the submerged type DHX.
Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Nakane, Shigeru*; Ishida, Katsuji*
no journal, ,
Thermal-hydraulic phenomena caused by the natural circulation in a reactor vessel were investigated using scaled model water experiments simulating the reactor vessel in order to enhance safety and optimize the design and operation of decay heat removal systems under normal operation and severe accident conditions. This report shows the measurement results of temperature and PIV of natural convection flow field in the reactor vessel simulating an operation of dipped type direct heat exchanger. In addition, it is shown that the effect of dispersal condition of molten fuel on natural convection flow field in the reactor vessel.
Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu
no journal, ,
Thermal-hydraulic phenomena driven by natural circulation in a reactor vessel was investigated by using scaled model water experiments simulating a reactor vessel in order to enforce of safety and optimize design and operation of decay heat removal systems under normal operation and severe accident conditions. This report shows temperature and PIV measurement results of natural convection flow field in the reactor vessel under operations of dipped-type DHX and penetrated-type DHX. From the experimental result, it was conformed that the effect of the operation of the penetrated-type DHX on the natural convection behavior in the reactor vessel including the cooling of core and debris on core catcher.