A phenomenological model of thermal-hydraulics of convective boiling during the quenching of hot rod bundles
After completion of the thermal-hydraulic model developed in a companion paper, the authors performed developmental assessment calculation of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The overall interfacial drag model predicted reasonable drag coefficients for both the nucleate boiling and the inverted annular flow (IAF) regimes. The predicted pressure drops agreed reasonably well with the measured data of two transient experiments, CCTF Run 14 and a Lehigh reflood test. The thermal-hydraulic model for post-CHF convective heat transfer predicted the rewetting velocities reasonably well for both experiments. The predicted average slope of the wall temperature traces for these tests showed reasonable agreement with the measured data, indicating that the transient-calculated precursory cooling rates agreed with measured data. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. The interfacial heat-transfer model tended to slightly underpredict the vapor temperatures. The maximum difference between calculated and measured vapor temperatures was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall temperatures were in reasonable agreement with measured data with a maximum relative error of less than 13%.
- Research Organization:
- Los Alamos National Lab., NM (USA)
- Sponsoring Organization:
- DOE; USDOE, Washington, DC (USA)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 5885253
- Report Number(s):
- LA-UR-91-252; CONF-910739--9-Pt.2; ON: DE91007470
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
220100 -- Nuclear Reactor Technology-- Theory & Calculation
220900 -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400* -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
BOILING
CRITICAL HEAT FLUX
ENERGY TRANSFER
FLUID MECHANICS
HEAT FLUX
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
NUCLEATE BOILING
PHASE TRANSFORMATIONS
PRESSURE DROP
QUENCHING
REACTOR ACCIDENTS
ROD BUNDLES
STEADY-STATE CONDITIONS
TESTING
TRANSIENTS
VALIDATION
WALLS