Validation of ANSYS Model of Experimental Test Rig Simulating the Flow Inversion in RRs
Archives of Current Research International,
Page 19-28
DOI:
10.9734/acri/2022/v22i430283
Abstract
The experimental setup was built to simulate the flow inversion in natural circulation loops in research reactors (RRs). In an effort to recognize the buildup of natural circulation in RRs, pool type upward flow after the pump coasts down due to power loss, by Abdel-Latif et al. [1], was investigated. The setup consists of two vertically stacked pipes that simulate the two branches, one of which contains a test section that is composed of electrically heated, corresponding channels that simulate the core. The second one, represents the portion of the coming back pipe that is involved in the growing of core natural circulation. Several experimental tests under various conditions as the branch’s initial temperature are performed. The channel’s coolant and surface temperatures were monitored. In this study, the thermal-hydraulic (TH) behaviour of the setup is complemented by theoretical analysis using the ANSYS Fluent 17.2 code. The ANSYS Fluent model is validated against the measured values. Typically, the setup is nodlized and a code input is being prepared. The results show that ANSYS Fluent 17.2 qualitatively predicts the thermal hydraulic behaviour and associated flow inversion phenomenon of such facilities. There is a difference between the predicted and measured values, especially for the channel’s surface temperature.
Keywords:
- Research reactores
- parallel channel
- ansys fluent
- natural circulation
- flow inversion
How to Cite
References
Takeda T, Kawamura H, Seki M. Natural Circulation in parallel vertical channels with different heat input, Nuclear Engineering and Design. 1987;104:133-143.
Chang HOH, Stan BE. Critical Heat Flux for Low Flow Boiling in Vertical Uniformly Heated Thin Rectangular Channels. International Journal of Heat and Mass Transfer. 1993;36(2):325-335.
Yang BW, Ouyang W. Dynamics and Developing of Natural Circulation Cooling from Vertical Up flow and down Flow Conditions. The 4th international Topical Meeting on Nuclear thermal hydraulics, operation and safety Columbia University; 1994.
Abe N, Yokobori S, Nagasaka H, Tsunoyama S. Two-Phase Flow Natural Circulation Characteristics Inside BWR Vessels. Nuclear Engineering and Design. 1994;146(1-3):253-265.
Khattab MS, Mina AR. Core Conversion from Rod to Plate Type Fuel Element in Research Reactors. Arab Journal of Nuclear Sciences and Applications. 1999; 30:247-255.
Housiadas C. Simulation of loss-of-flow transients in research reactors. Annals of Nuclear Energy. 2000;27:1683-1693.
Wang Y, Vafai K. Experimental Investigation of the Transient Characteristics on a Flat-Plate Heat Pipe during Startup and Shut down Operations. Journal of Heat Transfer. 2000;122(3): 525-535.
Meng1 H, zhu1 H, Wang2 T. Study on Flow Instability in Natural Cycle Parallel Channel Based on RELAP5. Experimental Thermal, IOP Conf. Series: Earth and Environmental Science. 2021;770:012-038. DOI:10.1088/1755-1315/770/1/012038.
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