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|Title:||PREDICTION OF DRYOUT AND POST DRYOUT BEHAVIOR OF CLAD SURFACE DURING FLOW OF WATER OVER A VERTICAL HEATED ROD|
|Keywords:||MECHANICAL INDUSTRIAL ENGINEERING|
VERTICAL HEATED ROD
|Abstract:||Critical flux is an important parameter in the designing and operation- of nuclear reactor and many other devices like heat exchanger. Critical heat flux occurs in the processes by two mechanisms - - Departure from nucleate boiling - Annular film dry out. Annular film dryout is the expected mode of critical heat flux in the boiling water reactor. CHF is traditionally . been evaluated using look up tables or empirical correlations. A model of annular two phase flow is used to calculate dryout on the assumption that dry out occurs when the liquid flow rate in the film on the solid surfaces becomes equal to zero. The theoretical model is based on fundamental mass conservation principle. Through numerically solving the equations, liquid film thickness, CHF, critical quality and Clad surface temperature at the critical point is obtained. Flow nucleate boiling has an extremely high heat transfer coefficient. The high heat transfer flux is limited by a maximum value. Above this. maximum heat flux, benign nucleate boiling is transformed to a film boiling of poor heat transfer. The transition of boiling mechanism, characterized by sudden rise . of surface temperature due to drop of heat transfer coefficient, is called boiling crisis. The maximum heat flux just before boiling crisis is called critical heat flux (CHF) and can occur in various flow patterns. Boiling crisis occuring in bubbly flow is called DNB; and boiling crisis occuring in annular flow is called dryout. Here in this report the methodology for the prediction of dryout is discussed and a computer code is developed for the same. Clad surface temperature is another important parameter, which is fixed by the metallurgical limits of the clad metal. Accurate prediction of the CHF will yield correct prediction of the clad surface temperature at the CHF value, which will be beneficial for the safe operation of the nuclear reactor.|
|Appears in Collections:||MASTERS' DISSERTATIONS (MIED)|
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