Heat transfer and pressure drop in an annular channel with downflow

Conference · Mon Jun 01 04:00:00 EDT 1992

OSTI ID:10147152

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: USDOE, Washington, DC (United States)
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 10147152
• Report Number(s): WSRC-MS--92-123; CONF-920804--10; ON: DE92014363
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220200
220600
220900
ANNULAR SPACE
COMPONENTS AND ACCESSORIES
FLOW RATE
FLUID FLOW
HEAT FLUX
HEAT TRANSFER
HEATING
HYDRAULICS
LOSS OF COOLANT
NUCLEATE BOILING
PRESSURE DROP
PRODUCTION REACTORS
REACTOR CHANNELS
REACTOR SAFETY
RESEARCH
TEST
TRAINING
PRODUCTION
IRRADIATION
MATERIALS TESTING REACTORS
SAVANNAH RIVER PLANT
TESTING