Heat transfer and pressure drop in an annular channel with downflow

Conference · Wed Jan 01 04:00:00 EST 1992

OSTI ID:5348855

Dolan, F X; Crowley, C J ^[1]; Qureshi, Z H ^[2]

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1. Creare, Inc., Hanover, NH (United States)
2. Westinghouse Savannah River Co., Aiken, SC (United States)

The onset of a flow instability (OFI) determines the minimum flow rate for cooling in the flow channels of a nuclear fuel assembly. A test facility was constructed with full-scale models (length and diameter) of annular flow channels incorporating many instruments to measure heat transfer and pressure drop with downflow in the annulus. Tests were performed both with and without axial centering ribs at prototypical values of pressure, flow rate and uniform wall heat flux. The axial ribs have the effect of subdividing the annulus into quadrants, so the problem becomes one of parallel channel flow, unlike previous experiments in tubes (upflow and downflow). Other tests were performed to determine the effects if any of asymmetric and non-uniform circumferential wall heating, operating pressure level and dissolved gas concentration. Data from the tests are compared with models for channel heat transfer and pressure drop profiles in several regimes of wall heating from single-phase forced convection through partially and fully developed nucleate boiling. Minimum stable flow rates were experimentally determined as a function of wall heat flux and heat distribution and compared with the model for the transition to fully developed boiling which is a key criterion in determining the OFI condition in the channel. The heat transfer results in the channel without ribs are in excellent agreement with predictions from a computer model of the flow in the annulus and with empirical correlations developed from similar tests. The test results with centering ribs show that geometrical variations between the channels can lead to differences in subchannel behavior which can make the effect of the ribs and the geometry an important factor when assessing the power level at which the fuel assembly (and the reactor) can be operated to prevent overheating in the event of a loss-of-coolant-accident (LOCA).

• Research Organization: Westinghouse Savannah River Co., Aiken, SC (United States)
• Sponsoring Organization: DOE; USDOE, Washington, DC (United States)
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 5348855
• Report Number(s): WSRC-MS-92-123; CONF-920804--10; ON: DE92014363
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
ANNULAR SPACE
BOILING
CONFIGURATION
ENERGY TRANSFER
FLOW RATE
FLUID FLOW
FLUID MECHANICS
HEAT FLUX
HEAT TRANSFER
HEATING
HYDRAULICS
LOSS OF COOLANT
MECHANICS
NATIONAL ORGANIZATIONS
NUCLEATE BOILING
PHASE TRANSFORMATIONS
PRESSURE DROP
PRODUCTION REACTORS
REACTOR ACCIDENTS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
SAVANNAH RIVER PLANT
SPACE
TESTING
US AEC
US DOE
US ERDA
US ORGANIZATIONS