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Ohira, Hiroaki; Honda, Kei; Sotsu, Masutake
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 12 Pages, 2011/09
In order to evaluate the upper plenum thermal-hydraulics of the Monju reactor vessel, we have performed detail calculations under the 40% rated power operational condition using high resolution mesh models by a commercial FVM code, FrontFlow/Red. In this study, we applied a high resolution meshes around the flow holes (FHs) on the inner barrel. We mainly made clear that the thermal-hydraulics did not change largely since the flow rates through the FHs were small enough to the total coolant flow rate but were affected largely incase without FHs on the honeycomb structure.
Kamide, Hideki; Kobayashi, Jun; Ono, Ayako; Kimura, Nobuyuki; Watanabe, Osamu*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 16 Pages, 2011/09
Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan sodium cooled fast reactor (JSFR). Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and start-up transient of the DHRS loop with parameters of pressure loss coefficients in the loops. Influences of the pressure loss coefficient in the primary loop and the DHRS loop were limited on the core temperature and also heat removal of PRACS due to recovery of natural circulation head via the increase of temperature difference in each loop.
Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 13 Pages, 2011/09
In the design of Japan Sodium-cooled Fast Reactor (JSFR), the flow-induced vibration (FIV) has been studied for the large-diameter hot-leg pipe with a short-elbow. The FIV will be a phenomenon caused by the pressure fluctuation in the pipe. In this study, for the purpose of clarification of FIV mechanism, the velocity and pressure fluctuations in the elbow pipe were measured. It was found that the pressure fluctuation on the wall with elbow was closely related with the movement of separation region formed near the elbow outlet.
Ezure, Toshiki; Kimura, Nobuyuki; Kobayashi, Jun; Kamide, Hideki
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 13 Pages, 2011/09
In order to clarify the influence of kinematic viscosity (
) on the occurrence of vortex cavitation, a water experiment was carried out in a cylindrical tank with a suction pipe. The occurrences of vortex cavitation were measured under several fluid temperature conditions between 10
C and 80C ( : 1.3
10
to 3.710m
/s). The velocity fields around vortex were also measured by Particle Image Velocimetry. The influence of was observed under relatively high conditions. However, that influence diminished with the decrease of or the increase of suction velocity. And also, normalized circulation 
was found as an index to estimate such influences of or suction velocity on the vortex cavitation.
Uchibori, Akihiro; Watanabe, Akira*; Ohshima, Hiroyuki
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 11 Pages, 2011/09
A SERAPHIM program has been developed to calculate multicomponent multiphase flow involving sodium-water chemical reaction under a tube failure accident in a steam generator of sodium-cooled commercial fast reactors. In this study, the experiment on supersonic chlorine jet into Na-NaCl mixture was analyzed to validate applicability of the numerical models to reactive multiphase flows. The program uses a multi-fluid model and a HSMAC method modified for compressible multiphase flows. A mass generation rate by chemical reaction is evaluated from the assumption that progress of chemical reaction is limited by a mass flow rate of a reactant gas toward a liquid surface. Numerical results showed that the injected gas disappeared at a certain height. The calculated plume length showed good agreement with the experimental data against three different experimental conditions. The proposed numerical models were found to be applicable to multiphase flow with chemical reaction.
Doda, Norihiro; Ohshima, Hiroyuki; Kamide, Hideki; Watanabe, Osamu*; Okubo, Yoshiyuki*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 13 Pages, 2011/09
Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed, in which fully natural circulation system is adopted as the decay heat removal system. A new evaluation method of core hot spot which can be applied to natural circulation decay heat removal has been developed. The new method consists of three-step thermal hydraulics analyses in order to consider the effects of physical phenomena particular to natural circulation, such as inter-fuel-assembly heat transfer and flow redistribution in the core due to buoyancy force. From the viewpoint of calculation cost reduction, we have also developed a simplified model substituting for the third step analysis (subchannel analysis). The new method was applied to the evaluations of a loss-of-external-power event and of a sodium leakage accident in a secondary loop of a large scale reactor.
Ninokata, Hisashi*; Kamide, Hideki
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 20 Pages, 2011/09
Key issues in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. Design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., fully natural circulation decay heat removal and recriticality free core, have been investigated in order to achieve higher level of reactor safety. Preliminary evaluations are on-going. Here, progress of design study is introduced.
Nakamura, Hideo; T
th, I.*; Sandervag, O.*; Umminger, K.*; Dreier, J.*; Prior, R.*; Alonso, J. R.*; Muellner, N.*; D'Auria, F.*; M
hleisen, A.*; et al.
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 15 Pages, 2011/09
The working group on Analysis and Management of Accidents (WGAMA) of the Committee on the Safety of Nuclear Installations (CSNI) of OECD-NEA had a task on the effectiveness of CET indication in accident management (AM) of light water reactors (LWR). The task collected and reviewed the design basis of CET application for AM procedures through a survey of the CET use in the NEA member countries, and reviewed pertinent experimental results from such test facilities as LOFT, ROSA/LSTF, PKL and PSB-VVER focusing on the time delay in CET from core temperature rise. Scaling issues were discussed considering extrapolation of experimental results to LWR. This paper summarizes major outcomes of the task and indicates possible future work.