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An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant [electronic resource].

Published
Rockville, Md. : U.S. Nuclear Regulatory Commission, 1995.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
Physical Description
pages 532-553 : digital, PDF file
Additional Creators
U.S. Nuclear Regulatory Commission and United States. Department of Energy. Office of Scientific and Technical Information
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Summary
In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, η, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70°C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm², respectively, which gave an η of 1.8.
Report Numbers
E 1.99:nureg/cp--0142-vol.1
E 1.99: conf-950904--vol.1
conf-950904--vol.1
nureg/cp--0142-vol.1
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Note
Published through SciTech Connect.
09/01/1995.
"nureg/cp--0142-vol.1"
" conf-950904--vol.1"
"TI95017077"
7. international topical meeting on nuclear reactor thermal-hydraulics (Nureth-7), Saratoga Springs, NY (United States), 10-15 Sep 1995.
Marshall, T.D.; McDonald, J.M.; Watson, R.D.
View MARC record | catkey: 14742812