Abstract The critical heat flux (CHF) limits relevant to the design of plasma facing components in tokamak fusion reactors are considered. Highly subcoolcd water in unobstructed pipe flow are investigated using experiments and computational models. The experiments employ water flowing through a 9.5 mm bore in a 19 mm x 19 mm copper monoblock. Single-sided heating of the block is achieved by passing an electric current through a 51 mm long plasma sprayed thin layer (0.4 mm) of tungsten overlaying a thin film (0.1 mm) of plasma sprayed ceramic on an outer wall. In the analysis, the heat transfer coefficient on the coolant-side wall relics on extrapolation of existing nucleate boiling correlations but is validated using outer wall temperature measurements and a heat conduction model. The hydraulic boundary conditions for 15 experimental bench mark points are: pressure between 2.2 and 3.0 MPa, coolant mass flux between 2.6 and 15 Mg/m2s. and equilibrium exit quality between −0.44 and −0.49. The critical heat flux ranges between 13 and 28 MW/m2. A correlation is formulated in which the data is fit as a relation between Stanton and Peclet numbers. The experimental results are combined with a CHF data base from several sources to enhance the generality of the proposed CHF correlation. The CHF data base parameter ranges are as follows: Peclet numbers between 7 x 104 to 3.2 x 106, coolant channel diameter between 5 and 25 mm, pressure between 1 and 7 MPa, and equilibrium quality between −0.49 and −0.07. The proposed correlation bounds the CHF data base as a lower limit and, thus, is an appropriate conservative limit for design applications.
Hechanova et al. (Sun,) studied this question.