A Large-Scale Numerical Simulation on Bubbly and Liquid Film Flows in Narrow Fuel Channels-Reference-Cited by-同舟云学术

A Large-Scale Numerical Simulation on Bubbly and Liquid Film Flows in Narrow Fuel Channels

Published:2005-01-01 Issue: Volume: Page:
ISSN:
Container-title:Fluids Engineering
language:
Short-container-title:

Author:

Takase K.¹,Yoshida H.¹,Ose Y.²,Akimoto H.¹

Affiliation:

1. Japan Atomic Energy Research Institute

2. Yamato Systems Engineer Company, Ltd.

Abstract

In order to predict the water-vapor two-phase flow structure in a fuel bundle of an advanced light-water reactor, large-scale numerical simulations were carried out using a newly developed two-phase flow analysis method and a highly parallel-vector supercomputer. Conventional analysis methods such as subchannel codes need composition equations based on many experimental data. Therefore, it is difficult to obtain highly prediction accuracy on the thermal design of the advanced light-water reactor core if the experimental data are insufficient. Then, a new analysis method using the large-scale direct numerical simulation of water-vapor two-phase flow was proposed. The coalescence and fragmentation of small bubbles were investigated numerically and the bubbly flow dynamics in narrow fuel channels were clarified. Moreover, the liquid film flow inside a tight-lattice fuel bundle which is used to the advanced light-water reactor core was analyzed and the water and vapor distributions around fuel rods and a spacer were estimated quantitatively.

Publisher

ASMEDC

Link

http://asmedigitalcollection.asme.org/IMECE/proceedings-pdf/doi/10.1115/IMECE2005-80772/4531253/947_1.pdf

Reference11 articles.

1. Kelly, J. E., Kao, S. P., Kazimi, M. S., 1981, “THERMIT- 2: A two-fluid model for light water reactor subchannel transient analysis,” MIT-EL-81-014.

2. Sugawara S. , MiyamotoY., 1990, “FIDAS: Detailed subchannel analysis code based on the three-fluid and threefield model,” Journal of Nuclear Engineering and Design, 129, pp. 146–161.

3. Thurgood, M. J., 1983, “COBRA/TRAC — A thermalhydraulic code for transient analysis of nuclear reactor vessels and primary coolant systems, equation and constitutive models,” NURREG/CR-3046, PNL-4385, Vol. 1, R4.

4. Iwamura, T. et al., 2002, “Development of reduced-moderation water reactor (RMWR) for sustainable energy supply,” Proceedings of the 13th Pacific Basin Nuclear Conference (PBNC 2002), Shenzhen, China, pp. 1631–1637.

5. Iwamura, T., 2002, “Core and system design of reduced-moderation water reactor with passive safety features,” Proceedings of International Congress of Advances in Nuclear Power Plants (ICAPP 2002), No. 1030, Hollywood, Florida, USA.