Outlet sampling measurement of mass flux, enthalpy and void fraction in rod bundles

Thesis/Dissertation · Mon Jan 01 04:00:00 EST 1979

OSTI ID:5337821

Sreepada, S R

The thermal-hydraulic performance of nuclear reactor cores is based on semi-empirical correlations and the local thermal-hydraulic conditions of the coolant, inferred analytically (using computer codes such as COBRA) from the rod bundle averaged conditions. The experimental data on local conditions of the coolant, such as mass flux, enthalpy and void fraction are limited and do not cover a wide range of thermodynamic variables. The improvements in the experimental isokinetic sampling technique for the measurement of enthalpy and mass flux are presented. Experiments were carried out on a 16 rod bundle prototypical of a boiling water reactor. Measurements were carried out on two subchannels. The experimental data are presented. Measurements were compared with the predictions of the computer code COBRA. The areas of disagreement between the measurements and the code predictions are presented along with the suggested code improvements. A dissolved radio-active salt technique for the measurement of subchannel void fractions is developed. The details of the technique and experimental void fraction measurements are presented. Future improvements of the method are suggested.

• Research Organization: Columbia Univ., New York (USA)
• OSTI ID: 5337821
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
BWR TYPE REACTORS
C CODES
COMPUTER CALCULATIONS
COMPUTER CODES
DATA
ENERGY TRANSFER
ENTHALPY
EXPERIMENTAL DATA
FLUID MECHANICS
FUEL ASSEMBLIES
FUEL ELEMENT CLUSTERS
HEAT TRANSFER
HYDRAULICS
INFORMATION
MECHANICS
NUMERICAL DATA
PHYSICAL PROPERTIES
REACTOR COMPONENTS
REACTOR CORES
REACTORS
ROD BUNDLES
THERMODYNAMIC PROPERTIES
VOID FRACTION
WATER COOLED REACTORS
WATER MODERATED REACTORS