Abstract
The present paper provides a discussion of the thermal-hydraulics requirements in fusion reactor components, with particular reference to the removal of high heat flux from plasma facing components and the critical heat flux (CHF) limit. Available experimental data on CHF of subcooled flow boiling in water, in the ranges of interest of fusion reactors thermal-hydraulic conditions, i.e. high inlet subcooling and velocity, and small channel diameter and length, are analyzed to discuss the influence of these parameters on CHF. The reference data-set (1887 experimental points) covers a wide range of operating conditions in the frame of present interest (0.1 < p < 8.4 MPa; 0.3 < D < 25.4 mm; 0.25 < L < 61 cm; 900 < G < 90000 kg/m 2·s; 0.3 < T in < 242.7°C). The aim of the research was to identify a new correlation based on a structure representing the relation of heat balance and using a non-linear regression analysis of the available data-set. A preliminary correlation (DINCE-92), based on 544 data points, had been developed providing a sensible improvement in predictions with respect to available predictive tools. Now, a new correlation (DINCE-93), based on the same structure of the above one and characterized by a very good statistics using a total of 1887 experimental points (88% of predictions are within ±20%) and by an R.M.S. error of 14.2%, has been identified and analyzed.
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