Core Thermal Hydraulics

Abstract

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 three-fluid 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.

Keywords

Mass Flow Rate Void Fraction Fuel Assembly Critical Heat Flux Annular Flow 
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  1. 1.ErlangenGermany

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