The present paper reports the results of an experimental investigation about the occurrence of the critical heat flux (CHF) in subcooled flow boiling of water, carried out to ascertain the influence of thermal hydraulic parameters on CHF under conditions typical of thermonuclear fusion divertor thermal hydraulic design. Helically coiled wires were used as turbulence promoters to enhance the CHF with respect to the smooth channel. Geometric characteristics of stainless steel 304 Type test sections were : 6.0 and 8.0 mm i.d., 0.25 mm wall thickness, 0.1 and 0.15 m heated length, horizontal and vertical (upflow) position. Test sections were uniformly heated using d.c. current. A maximum CHF of about 30 MW m-2 was reached with smooth tubes under the following conditions : T(in) = 30-degrees-C, p = 4.6 MPa, u = 10 in s-1, D = 8.0 mm, L = 0.1 m. Helically coiled wires (d = 1.0 mm, pitch = 20.0 mm) allowed an increase of the CHF up to 50%, with reference to smooth channels, coupled with a moderate increase of pressure drop (down to 25%). Pressure revealed a negative effect on the efficiency of turbulence promoters. No observable influence of the channel orientation was detected.