Two-Phase Flow Heat Transfer in Nuclear Reactor SystemsView this Special Issue
Editorial | Open Access
Boštjan Končar, Eckhard Krepper, Dominique Bestion, Chul-Hwa Song, Yassin A. Hassan, "Two-Phase Flow Heat Transfer in Nuclear Reactor Systems", Science and Technology of Nuclear Installations, vol. 2013, Article ID 587839, 2 pages, 2013. https://doi.org/10.1155/2013/587839
Two-Phase Flow Heat Transfer in Nuclear Reactor Systems
Heat transfer and phase change phenomena in two-phase flows are often encountered in nuclear reactor systems and are therefore of paramount importance for their optimal design and safe operation. The complex phenomena observed especially during transient operation of nuclear reactor systems necessitate extensive theoretical and experimental investigations.
This special issue brings seven research articles of high quality. Though small in number, they cover a wide range of topics, presenting high complexity and diversity of heat transfer phenomena in two-phase flow. In the last decades a vast amount of research has been devoted to theoretical work and computational simulations, yet the experimental work remains indispensable for understanding of two-phase flow phenomena and for model validation purposes. This is reflected also in this issue, where only one article is purely experimental, while three of them deal with theoretical modelling and the remaining three with numerical simulations.
The experimental investigation of the critical heat flux (CHF) phenomena by means of photographic study is presented in the paper of J. Park et al. They have used a high-speed camera system to observe the transient boiling characteristics on a thin horizontal cylinder submerged in a pool of water or highly wetting liquid. Experiments show that the initial boiling process is strongly affected by the properties and wettability of the liquid. The authors have stressed the importance of the local scale observation leading to better understanding of the transient CHF phenomena.
In the article of G. Espinosa-Paredes et al. a theoretical work concerning the derivation of transport equations for two-phase flow is presented. The author proposes a novel approach based on derivation of nonlocal volume averaged equations which contain new terms related to nonlocal transport effects. These non-local terms act as coupling elements between the phenomena occurring in at least two different spatial scales.
Uncertainty in modelling of bubble departure diameter at boiling was studied by M. Matkovic and B. Koncar. In this article the propagation of input uncertainties for the simplified model of bubble departure size is evaluated. A methodology for estimating the prediction capability of a given correlation is provided taking into account its range of applicability.
Aqueous nanofluids have a great potential for cooling applications, hence they have been studied in the article of P. N. Alekseev et al. as a possible coolant in pressurized water reactor (PWR). The theoretical study presents how a stable formation of nanoparticles in water solution can be established. Formation of fractal nanoparticles with a higher thermal conductivity than water can enhance the heat transfer of water used as a coolant in PWR. Apart from solid particles, also alternative formation of gaseous nanoparticles in density fluctuations of water is discussed.
The article of R. Rzehak and E. Krepper provides a comprehensive overview of the state-of-the-art in the field of CFD modelling of subcooled flow boiling. The efficient predictive capability of current models requires calibration of model parameters over a wide range of measured data and operating conditions. The results presented in the article confirmed the great potential of the existing modelling approach for the 3D simulation of subcooled flow boiling in industrial applications but also highlight the need for specific model improvements to achieve highly accurate predictions.
Two articles deal with one-dimensional analyses of two-phase flows. In the article of O. Costa et al., a rapid depressurization in vertical heated pipe is simulated with the in-house 1D computer code WAHA, which was developed specifically for simulations of two-phase water hammer phenomena. The WAHA results were confronted with the simulations of the well-known system code RELAP5 on the same experimental data.
The thermal-hydraulic system code TRACE was used for validation of typical transient in Boiling Water Reactor (BWR) as elaborated by V. H. Sanchez et al. A validation of BWR relevant models on the BFBT benchmark experiments was upgraded by development of an integral plant model of the German BWR. The critical power tests with the 2D plant model demonstrated that TRACE predictions are in good agreement with the recorded plant data.
We believe that this special issue provides interesting information on the recent progress in nuclear thermal-hydraulics research.
Yassin A. Hassan
Copyright © 2013 Boštjan Končar et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.