A framework for critical heat flux prediction in high heat flux, high subcooling components
- Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering
The critical heat flux (CHF) limits relevant to the design of plasma facing components in tokamak fusion reactors are considered. Highly subcooled 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-sized 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 relies on extrapolation of existing nucleate boiling correlations but is validated using outer wall temperature measurements and a heat conduction model. 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 {times}10{sup 4} to 3.2 {times} 10{sup 6}, coolant channel diameter between 5 and 25 mm, pressure between 1 and 7 MPa, and equilibrium quality between {minus}0.49 and {minus}0.07. The proposed correlation bounds the CHF data base as a lower limit and, thus, is an appropriate conservative limit for design applications.
- OSTI ID:
- 458409
- Report Number(s):
- CONF-951135--; ISBN 0-7918-1751-2
- Country of Publication:
- United States
- Language:
- English
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