Multiphase Science and Technology article. The full article is published by Begell House Inc. and available at: http://www.begellhouse.com/ .

Critical heat flux in subcooled flow boiling- an assessment of current understanding and future directions for research

Effects of flow obstacles on the critical heat flux in a vertical tube cooled with upward flow of R-134a

Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

After a short review of existing nuclear reactor pressure vessels and core geometry this chapter summarizes the main ideas on how to perform a thermohydraulic analysis of boiling flows in nuclear reactors. Remarkable is that all known threefluid flow patterns and heat transfer mechanisms create the cooling mechanism of the core. To demonstrate the achievable accuracy using gross discretization I simulate 333 three-dimensional tests on bundles with 1, 16, 25, and 36 heated rods, seven different bundles with 64 heated rods from different laboratories: 273 3D experiments on six bundles for critical heat flux, 54 3D experiments on seven bundles for void fraction, two 3D experiments on a bundle for transients and four 1D experiments on a subchannel for transients. The mass flow rates and the pressure in these tests varied from 3 to 2000 kg/(m2 s) and from 1 to 200 bar, respectively. The subcooling was less than 140 K and the thermal power varied from some tenths of a kilowatt to 7 MW. Comparison with the results of other authors is made and a discussion is provided.

Core thermal hydraulics

The emphasis in this chapter is on averaged turbulence modeling in rod-bundles on an intermediate scale that is finer than the sub channel scale but much larger than the scale required for direct numerical simulation. Thirteen benchmarks based on analytical solutions and experimental data are presented and compared with the prediction of the IVA computer code. For the first time distribution of the averaged turbulence structure in boiling bundles is presented. It is demonstrated that k-eps models with two-way coupling possess substantial potential for increasing the accuracy of the description of multi-phase flow problems in bundles especially the effect of the space grids.

Set of benchmarks for verification of k-eps models in system computer codes

https://asmedigitalcollection.asme.org/IMECE/proceedings-pdf/doi/10.1115/IMECE2000-1518/6791680/169_1_imece2000-1518.pdf

K. F. Rudzinski

Papers

Some Aspects of Critical-Heat-Flux Enhancement in Tubes