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Iguchi, Tadashi; Shibamoto, Yasuteru; Asaka, Hideaki; Nakamura, Hideo
Proceedings of 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), 16 Pages, 2003/10
Thermal-hydraulic and neutronic dynamics are always interrelated in BWR core. This is called thermal-hydraulic and neutronic (T/N) coupling. Channel stability experiments with T/N coupling under non-nuclear condition are very limited. This is mainly due to the difficulties in the real-time simulation of neutron dynamics and in the fast-response void fraction measurement under high-pressure and temperature conditions. Authors have developed techniques to solve the above difficulties, and have succeeded in experimentally simulating T/N coupling under non-nuclear conditions with the THYNC facility. Using THYNC facility, T/N coupling effect on channel stability was investigated. Experiments were performed under Pressure=2-7MPa, Subcooling=10-40K, and Mass flux=270-660kg/m
s. THYNC results indicated T/N coupling lowered the channel stability threshold. The reduction of channel stability threshold due to T/N coupling was small within 10% at 7MPa in the present THYNC experiment, although the experimental condition was set to be more severe than that supposed in a reactor.
; Tanaka, Masaaki; Kimura, Nobuyuki; Kamide, Hideki
Proceedings of 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), CDROM, A0512 Pages, 2003/00
Research of high cycle thermal fatigue is significant for the safety of a fast reactor. In Japan Nuclear Cycle Development Institute, several experiments and the improvement of the analysis methods have been carried out to construct an evaluation rule of high cycle thermal fatigue. In this study, fluid temperature and flow velocity distributions were measured and direct numerical simulation (DNS) code was applied to understand mixing phenomena. From the experimental results, a prominent frequency of temperature fluctuation was observed just down stream from the branch pipe where temperature fluctuation intensity was also large. The velocity measurement showed that the vortices were generated in the wake region behind the branch pipe jet. This vortex was correlated with prominent frequency of the temperature fluctuation. The temperature and velocity field and the two vortices in the wake region were reproduced by DNS.
Kimura, Nobuyuki; Hayashi, Kenji; Igarashi, Minoru; Kamide, Hideki; Ito, Masami*; Sekine, Tadashi*
Proceedings of 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), CD-RO, K0102 Pages, 2003/00
An innovative sodium cooled fast reactor has been investigated on the feasibility study of FBR cycle system in JNC. A compact reactor vessel (R/V) and a column type upper inner structure (UIS) with a radial slit for an arm of a fuel-handling machine (FHM) are adopted. Dipped plates (D/P) are set in the R/V below the free surface to prevent gas entrainment at free surface. We performed an 1/10th scaled model water experiment for the upper plenum of the R/V. Gas entrainment at the free surface was not observed in the experiment under the same velocity condition as the designed reactor. However, the free surface rose in front of the UIS slit due to upward flow through the gap between the D/P and the R/V wall. The upward flow will cause free surface vortex and also the gas entrainment. Three vortex cavitations were observed near the hot leg (H/L) inlet. The vortex cavitations were broken out under the same cavitation factor condition as the reactor. A vertical rib was set on the R/V wall