Critical Heat Flux Tests of a Simulated Power Burst Facility Rod Bundle

Technical Report · Tue Aug 01 00:00:00 EDT 1978

DOI:https://doi.org/10.2172/6638854· OSTI ID:6638854

Ambrosek, Richard G. ^[1]; Wadkins, Robert P. ^[1]; Young, Michael W. ^[1]

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1. Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States)

Critical heat flux (CHF) tests were performed at low pressure in a close-packed rod bundle. The test bundle was electrically heated with geometrical configurations the same as the Power Burst Facility nuclear core. Existing low pressure CHF correlations, namely Bernath's and Lund's, did not correlate the test data well. The Bernath correlation overpredicts CHF in some cases by a factor of five when compared with measured values. Lund's correlation overpredicts CHF at measured CHF values above 1.5 MW/m², and underpredicts CHF at measured CHF values below 1.5 MW/m². These CHF tests provided the first close-packed rod bundle data with a sufficient data base to develop a correlation. The study examined CHF with absolute coolant system pressures of 117 to 255 kPa, mass velocities of 1992 to 4830 kg/s . m², and subcooling of up to 53°C with a rod spacing of 1.02 mm.

• Research Organization: Idaho National Engineering Laboratory (INEL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE; USNRC
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 6638854
• Report Number(s): TREE--1170; NUREG/CR-0260
• Country of Publication: United States
• Language: English

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