Prediction of low flow critical heat flux in rod bundles
Creators
- 1. Heat Transfer Research Facility, Dept. of Chemical Eng., Columbia Univ., NY, NY
Description
An accurate but simple low flow CHF correlation is developed using bundle average conditions from an extensive rod bundle CHF data base for use in safety analysis codes such as RELAP, RETRAN, TRAC, and RELAP-UK. The low flow rate range is defined as mass velocities below 0.55 M.lbs/hr/sq. ft. (746 Kg/m2.s). The Columbia low flow CHF correlation predicts the source data set of 818 CHF points from 52 fuel assemblies simulating both PWR and BWR cores with an average ratio of 0.994, standard deviation of 7.06%, and a RMS error of 7.09%. A comparison of this correlation with the existing correlations used in computer codes such as RELAP, RETRAN, TRAC, and RELAP-UK showed that the Columbia correlation represents significant improvement in predicting CHF for low flow conditions
Additional details
Publishing Information
- Publisher
- American Society of Mechanical Engineers.
- Imprint Place
- New York, NY (USA)
- Imprint Title
- Proceedings of the ASME winter annual meeting
- Journal Page Range
- p. 8.
Conference
- Title
- ASME winter annual meeting.
- Dates
- 9-13 Dec 1984.
- Place
- New Orleans, LA (USA).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 17069419
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BWR TYPE REACTORS; COMPUTER CODES; CRITICAL HEAT FLUX; FLOW RATE; FORECASTING; FUEL ASSEMBLIES; FUEL ELEMENT CLUSTERS; PWR TYPE REACTORS; R CODES; T CODES
- Descriptors DEC
- ENRICHED URANIUM REACTORS; HEAT FLUX; POWER REACTORS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS