Void fraction under high pressure, low flow conditions in rod bundle geometry
Creators
- 1. Oak Ridge National Lab., TN (USA). Engineering Technology Div.
- 2. Cornell Univ., Ithaca, NY (USA). Lab. of Nuclear Engineering
Description
Void-fraction data are reported from a series of high pressure, low heat and mass flux experiments. Testing was performed in a heated vertical rod bundle with internal dimensions similar to a PWR fuel bundle. The results are of interest in analyses of small break loss of coolant accidents. The experiments showed that, at a given pressure, void-fraction data could be fitted to a drift-flux equation with a constant drift-velocity. The drift-velocity was observed to decrease with increasing pressure and was independent of void fraction: a characteristic normally associated with churn-turbulent flow. However, relevant drift-flux correlations found in the literature gave relatively poor predictions of void fractions. The best predictions were obtained from an empirical correlation based on dimensional analyses. (orig.)
Additional details
Publishing Information
- Journal Title
- Nucl. Eng. Des.
- Journal Volume
- 75
- Journal Issue
- 1
- Series
- CODEN: NEDEA.;Nucl. Eng. Des.
- Journal Page Range
- 99-108
- ISSN
- 0029-5493
INIS
- Country of Publication
- Netherlands
- Country of Input or Organization
- Netherlands
- INIS RN
- 15010224
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
- Descriptors DEI
- FLUID FLOW; FUEL ELEMENT CLUSTERS; HIGH PRESSURE; MASS TRANSFER; PWR TYPE REACTORS; REACTOR CORES; VOIDS
- Descriptors DEC
- FUEL ASSEMBLIES; REACTOR COMPONENTS; REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Contract/Grant/Project number
- Contract DOE 40-551-75; DOE 40-552-75; W-7405-eng-26