ETD system

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Tesi etd-02252016-204610


Thesis type
Tesi di dottorato di ricerca
Author
DEL FRATE, LICIA
URN
etd-02252016-204610
Title
Assessment of CFD methods for single and two-phase flows in nuclear reactors
Settore scientifico disciplinare
ING-IND/19
Corso di studi
INGEGNERIA "L. DA VINCI"
Commissione
tutor Prof. D'Auria, Francesco
relatore Dott. Galassi, Giorgio
relatore Dott. Moretti, Fabio
Parole chiave
  • CFD
  • nuclear
Data inizio appello
02/03/2016;
Consultabilità
parziale
Data di rilascio
02/03/2019
Riassunto analitico
The present work deals with the application of single-phase and two-phase Computational Fluid Dynamics (CFD) techniques to the nuclear technology, and in particular to the modelling of thermal-fluid-dynamic phenomena relevant for the Nuclear Reactor Safety (NRS) that cannot be successfully investigated and predicted by the system thermal-hydraulic computer codes, since they require a fully three-dimensional description. Boiling and the related momentum, heat and mass transfer phenomena are among the most important phenomena that affect both the nuclear reactor safety and operation, and they need a fully three-dimensional description.<br>The CFD tools able to deal with two-phase flows are still under development. For instance, large effort is being spent by the international community on the development and validation of the multiphase CFD code NEPTUNE_CFD. The present research aims to contribute to the assessment of NEPTUNE_CFD code testing its ability to simulate fundamental cases for which a theoretical calculation is possible or experimental data are available, in single-phase or two-phase field, with the main focus on the flow boiling processes. <br>The starting phase of the activity is focused on the analysis of two single-phase cases. The first case is a turbulent jet impinging on a flat heated surface. This test case has been chosen because, despite its relatively simple geometry, it has complex flow characteristics. In the nuclear field the jet impingement with heat transfer occurs in PTS scenarios, when cold water is injected into the cold leg of PWRs to mitigate the effects of a LOCA. The second case deals with the accuracy assessment of the friction pressure loss estimation based on Darcy formula combined with an equivalent hydraulic diameter and a friction factor valid for circular pipes when applied to a square rod bundle.<br>The following activity phase is focused on the validation of the boiling models of the NEPTUNE_CFD code. The test cases analysed are: the PSBT experiment, the Bartolomei experiment, the Boiling sub-channel test. The PSBT experiment includes PWR sub-channel and bundle tests aimed at the evaluation of the void fraction distribution in the test section. The Bartolomei experiment provides the averaged void fraction along a vertical channel with uniform heat release over the length (prototypical of BWRs and PWRs). The last section reports the activity done for the FONESYS Benchmark. The main goal of the Benchmark is the comparison of different system codes performance in modelling the so called “boiling channel exercise”. The present work reports a simplified simulation of this exercise in steady-state configuration performed with NEPTUNE_CFD. As in the case of the Bartolomei Loop the obtained results are evaluated against RELAP5 simulations.<br>All the cases are simulated with NEPTUNE_CFD and other CFD or system codes following the Best Practice Guidelines (BPG) set by NEA-CSNI in order to achieve the widest validation of the presented calculation results. The simulations are performed on different meshes and a numerical error analysis is carried out in order to compare the differences in turbulence models, boiling models or near wall treatment.
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