Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Shibamoto, Yasuteru; Kukita, Yutaka*; Nakamura, Hideo
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 15 Pages, 2005/10
no abstracts in English
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Iyoku, Tatsuo
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 12 Pages, 2005/10
Safety demonstration tests using the HTTR are in progress to verify the inherent safety features, to improve the safety design and the technologies for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely becomes a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The SIRIUS code was developed to analyze reactor transient during the tests with reactor dynamics. This paper describes the validation of the SIRIUS code with the measured values of one and two gas circulators tripping test at 30% (9 MW). It was confirmed that the SIRIUS code was able to analyze the reactor transient within 10% during the tests. The result of this study and the way of resolving problems can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) as one of the Generation IV reactors.
Watanabe, Tadashi
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 8 Pages, 2005/10
no abstracts in English
Ohshima, Hiroyuki; Imai, Yasutomo*
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 14 Pages, 2005/10
A numerical simulation system is being developed at JNC in order to offer methodologies to clarify thermal-hydraulic phenomena in a fuel subassembly of sodium-cooled fast reactors under various operating conditions. This paper describes the validation study of SPIRAL that is one component code of the numerical simulation system and that plays the role to simulate detailed local flow and temperature fields in a wire-wrapped fuel pin bundle. Fundamental validity related to solving mass, momentum and energy conservation equations and applicability of turbulence models were confirmed by simulating several basic problems. As the typical examples, two kinds of simulations using mainly high Re number models, backward facing step flow and flow in 4-fuel-pin bundle in a rectangular duct, are selected and the characteristics of flow field prediction by the models as well as the validity of the component code are mentioned.
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Tamai, Hidesada; Takase, Kazuyuki; Akimoto, Hajime
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 15 Pages, 2005/10
no abstracts in English
Kunugi, Tomoaki*; Ezure, Toshiki; Sakai, Takaaki; Ito, Kei
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 13 Pages, 2005/10
In order to design a compact fast reactor, it is necessary to clarify a criterion of a gas entrainment (GE) from the coolant free surface of the vessel. The surface velocity is considered as a key parameter of the GE. However, there is no quantitative criterion regarding the onset condition of the GE. In this study, two-dimensional waterfall simulation and a three-dimensional dimple simulation for the GE have been performed by means of the MARS method. The final goal of this study is to establish the evaluation procedure and clarify a criterion of these phenomena. Resulting two-dimensional waterfall simulations, the time-averaged wave shape is very similar to the experimental one. The surface wave velocity, the maximum surface head and the maximum surface slope are well correlated with inflow energy, i.e., Froude number. Moreover, the three-dimensional unsteady vortex-dimple motion can be captured by the MARS method and three flow modes in the vortex-dimple evolution can be observed.
Monji, Hideaki*; Akimoto, Toshinori*; Miwa, Daisuke*; Kamide, Hideki
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 10 Pages, 2005/09
In this study, the unsteady flow structure of the free surface vortex was investigated under periodical flow fluctuating conditions in a cylindrical vessel. The gas core length of the free surface vortex and the velocity field in the vessel were measured by image processing. The experimental results showed that the gas core length under the fluctuating flow condition was shorter than that in the steady state having the same flow rate as the maximum rate of the fluctuating flow when the period of the flow fluctuation was short.