Tesi etd-11092011-235820 |
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Tipo di tesi
Tesi di laurea vecchio ordinamento
Autore
PESETTI, ALESSIO
URN
etd-11092011-235820
Titolo
Review, implementation and validation of physical models in the FLICA code for application to LWR and SFR
Dipartimento
INGEGNERIA
Corso di studi
INGEGNERIA NUCLEARE
Relatori
relatore Prof. Ambrosini, Walter
correlatore Ing. Forgione, Nicola
relatore Dott. Bucci, Matteo
correlatore Prof. Oriolo, Francesco
correlatore Ing. Forgione, Nicola
relatore Dott. Bucci, Matteo
correlatore Prof. Oriolo, Francesco
Parole chiave
- boiling heat transfer
- CHF
- correlation
- critical heat flux
- DFFB
- DNB
- dryout
- friction multiplier
- IAFB
- post critical heat flux
- sodium pressure drop
- TB
Data inizio appello
28/11/2011
Consultabilità
Non consultabile
Data di rilascio
28/11/2051
Riassunto
This work has been performed during a stage of nine months, carried out at the Commissariat à l’ ́énergie atomique et aux énergies alternatives (CEA), in Saclay, France.
The work aims at improving the capabilities of the FLICA-OVAP code. FLICA-OVAP
is a three-dimensional code for two-phase thermal-hydraulic analysis. It is oriented to
several reactor types; it can be employed for safety analysis and design of light water
reactors, liquid metal reactors and gas cooled reactors.
The work consists in improving the FLICA code correlation database, by the review, the
implementation and the validation of physical models to enable the FLICA-OVAP code
to analyses properly heat transfer phenomena in the post critical heat flux region for light
water reactors and the two-phase pressure drops in sodium fast reactors. These two subjects are treated separately in Part I and Part II of this work, respectively.
In the first part, an extensive literature review has been performed, in order to understand
the physics of the phenomenon and to identify the more appropriate correlations. Then,
selected correlations have been implemented and validated against an extensive experimental database consisting of 219 tests, performed with boiling water flowing upward in
a vertical tube heated by an uniform and constant heat flux.
In the second part, correlations to estimate the two-phase flow multipliers and the slip
ratio in sodium boiling channels have been reviewed, implemented and validated against
experimental and analytical data, provided respectively by the Kottowski and Savatteri
database, and an analytical model developed on the CAST3M platform. The analysis is
performed for two-phase sodium up-flow in a vertical tube heated by a uniform and constant heat flux.
The results obtained in both first and second part of this thesis are in good agreement with
reference data. It confirms the suitability of the correlations implemented.
The work aims at improving the capabilities of the FLICA-OVAP code. FLICA-OVAP
is a three-dimensional code for two-phase thermal-hydraulic analysis. It is oriented to
several reactor types; it can be employed for safety analysis and design of light water
reactors, liquid metal reactors and gas cooled reactors.
The work consists in improving the FLICA code correlation database, by the review, the
implementation and the validation of physical models to enable the FLICA-OVAP code
to analyses properly heat transfer phenomena in the post critical heat flux region for light
water reactors and the two-phase pressure drops in sodium fast reactors. These two subjects are treated separately in Part I and Part II of this work, respectively.
In the first part, an extensive literature review has been performed, in order to understand
the physics of the phenomenon and to identify the more appropriate correlations. Then,
selected correlations have been implemented and validated against an extensive experimental database consisting of 219 tests, performed with boiling water flowing upward in
a vertical tube heated by an uniform and constant heat flux.
In the second part, correlations to estimate the two-phase flow multipliers and the slip
ratio in sodium boiling channels have been reviewed, implemented and validated against
experimental and analytical data, provided respectively by the Kottowski and Savatteri
database, and an analytical model developed on the CAST3M platform. The analysis is
performed for two-phase sodium up-flow in a vertical tube heated by a uniform and constant heat flux.
The results obtained in both first and second part of this thesis are in good agreement with
reference data. It confirms the suitability of the correlations implemented.
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