Begell House Inc.
Multiphase Science and Technology
MST
0276-1459
18
4
2006
PARALLEL-CHANNEL INSTABILITY IN NATURAL CIRCULATION SYSTEM
305-333
10.1615/MultScienTechn.v18.i4.10
Mio
Hirayama
Kansai University, Department of Mechanical Engineering, Yamate-cho 3-3-35, Suita, Osaka 564-8680, Japan
Hisashi
Umekawa
Department of Mechanical Engineering, Kansai University, 3-3-35 Yamate-cho, Suita-shi, Osaka 564-8680, Japan
Mamoru
Ozawa
Department of Safety Science, Kansai University, 7-1 Hakubai-cho, Takatsuki-shi, Osaka 569-1098, Japan
Recently, regional flow instability in two-phase flow system, i.e. flow fluctuations occurred only in the core region of nuclear reactor or multi-channel region without fluctuation of the total inlet flow, is attracting much interest. This flow instability is a significant problem for the systems from the viewpoint of safety and economy. Many investigations on the flow instabilities have been conducted so far, and respective stability criteria have been proposed, while these investigations focused normally on the one-channel system without flow maldistribution, and the resulting stability criteria are hardly applied to the regional flow instability problems.
Thus in the present investigation, we conducted analytical investigation focusing not only on the whole system flow instability but also on the regional flow instability. The stability analysis was performed with lumped-parameter model of parallel boiling channels. Finally the stability criteria for the two types of the flow instabilities and the stability maps were obtained.
EXPERIMENTAL STUDY ON STABILITY OF STARTUP IN NATURAL CIRCULATION BWRS WITH AND WITHOUT NUCLEAR COUPLING
335-358
10.1615/MultScienTechn.v18.i4.20
Mamoru
Ishii
Therma-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907, USA
Selim
Kuran
Global Nuclear Fuel - Americas
Xiaodong
Sun
Nuclear Engineering Program, Department of Mechanical and Aerospace Engineering, The Ohio State University, 201 W 19th Avenue, Columbus, OH 43210, USA
Ling
Cheng
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907, USA
Yiban
Xu
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907, USA
Ho Jun
Yoon
Thermal-Hydraulics and Reactor Safety Laboratory, School of Nuclear Engineering, Purdue University, 400 Central Drive, West Lafayette, IN 47907, USA
Shripad
Revankar
Purdue University
Several experiments have been performed in Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility to study the startup transient in a natural circulation BWR design. The strategy for the experimental study has been developed from pressure scaling based on single- and two-phase flow scaling laws. In the experiments, the differences of the stored energy of the solid structures and the core inlet subcooling between the facility and a natural circulation BWR design have been taken into account. A novel core inlet plate has been designed and installed to study the effect of the core inlet flow-loss coefficient on the startup instabilities, namely geysering and flashing-induced loop type oscillations. The experiments for the startup transient have been performed with and without considering the void-reactivity feedback. Based on a detailed analysis for the fuel heat conduction dynamics and neutron kinetics, the scaling criteria for the void-reactivity feedback simulation have been derived. By means of the local void fraction measurements with conductivity probes installed in the core section of the PUMA reactor pressure vessel, the void-reactivity feedback has been studied through the real-time solution of the point kinetic model equations. The presence of the void-reactivity coupling has been found to have a major destabilizing effect especially for the flashing-induced loop type oscillations.
AN ANALYSIS OF INTERACTING INSTABILITY MODES
359-385
10.1615/MultScienTechn.v18.i4.30
J.
Yin
Center for Multiphase Flow, Rensselaer Polytechnic Institute, Troy, NY, USA
Richard T.
Lahey, Jr.
Center for Multiphase Research, Rensselaer Polytechnic Institute, Troy, NY 12180-3590, USA
Michael Z.
Podowski
Center for Multiphase Flow, Department of Nuclear Engineering and Engineering Physics, Rensselaer Polytechnic Institute, Troy, New York, USA, 12180-3590
Michael K.
Jensen
Center for Multiphase Flow, Rensselaer Polytechnic Institute, Troy, NY, USA; University of Wisconsin-Milwaukee, Mechanical Engineering Department Milwaukee, Wisconsin 53201
It is well known that phase change systems may experience various system instabilities, such as: Ledinegg type excursive instabilities, density-wave oscillations (DWO) and pressure-drop oscillations (PDO). Moreover, there can be interaction between these instability modes.
The experimentally observed interaction of DWO and PDO instability modes is, for the first time, predicted in this paper. This interaction is particularly important in low pressure phase change systems, such as those which have been proposed by NASA to support their future missions associated with the human exploration and development of space (e.g., the manned mission to Mars).
HEAT TRANSFER AND FLOW CHARACTERISTICS OF A NON-UNIFORMLY HEATED TUBE UNDER LOW PRESSURE AND LOW MASS FLUX CONDITION
387-412
10.1615/MultScienTechn.v18.i4.40
Hisashi
Umekawa
Department of Mechanical Engineering, Kansai University, 3-3-35 Yamate-cho, Suita-shi, Osaka 564-8680, Japan
Mio
Hirayama
Kansai University, Department of Mechanical Engineering, Yamate-cho 3-3-35, Suita, Osaka 564-8680, Japan
T.
Kitajima
Dept. of Mechanical Engineering, Kansai University, Osaka, Japan
Mamoru
Ozawa
Department of Safety Science, Kansai University, 7-1 Hakubai-cho, Takatsuki-shi, Osaka 569-1098, Japan
Kaichiro
Mishima
Institute of Nuclear Safety System, Inc., 64 Sata, Mihama, Mikata, Fukui 919-1205, Japan
Yasushi
Saito
Institute for Integrated Radiation and Nuclear Science, Kyoto University, 2,
Asashiro-Nishi, Kumatori-cho, Osaka 590-0494, Japan
In an actual boiling channel, the circumferential heat flux is not uniform. Thus, the understanding of the heat transfer characteristics of non-uniformly heated tube becomes an important design factor for conventional boilers, especially for the compact water-tube boiler with tube-nested combustor. The small compact boiler is operated at low-pressure and low-mass flux condition compared with the large scale boiler, so that the non-uniformity in heat flux may strongly affect the heat transfer characteristics. In this investigation, non-uniform heat flux distribution along the circumferential direction was generated by using Joule heating of SUS304 eccentric tubes. The heated length of test-section was 900mm, inner diameter was 20mm, and outer diameter was 24mm. The eccentricity of the tube was s = 0, 0.5, 1.0 and 1.5mm. Corresponding maximum/minimum heat fluxes rations were 1.0, 1.7, 3.0 and 7.0, respectively. Experiments were conducted for upward and downward flows, with system pressures 0.3 and 0.4MPa, mass flux range 10-100kg/m2s and inlet temperatures 30 and 80deg.C. The eccentricity resulted in an increase in the CHF of the upward flow, while the effect was hardly observed in the downward flow. The effect of eccentricity on the CHF was well interpreted by the liquid film redistribution. In this paper, the local heat transfer coefficient along the circumferential direction was also presented.