Critical heat flux in natural convection cooled TRIGA reactors with hexagonal bundle

A three-rod bundle Critical Heat Flux (CHF) study at low flow, low pressure, and natural convection condition has been conducted, simulating TRIGA reactors with the hexagonally configured core. The test section is a custom-made trefoil shape tube with three identical fuel pin heater rods located symmetrically inside. The full scale fuel rod is electrically heated with a chopped-cosine axial power profile. CHF experiments were carried out with the following conditions: inlet water subcooling from 30 K to 95 K; pressure from 110 kPa to 230 kPa; mass flux up to 150 kg/m{sup 2}s. About 50 CHF data points were collected and compared with a few existing CHF correlations whose application ranges are close to the testing conditions. Some tests were performed with the forced convection to identify the potential difference between the CHF under the natural convection and forced convection. The relevance of the CHF to test parameters is investigated. (authors)