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Electronic theses and dissertations repository


Tesi etd-10222012-104242

Thesis type
Tesi di laurea specialistica
Thermal Hydraulic Analysis of Postulated Accidents in a HLM Cooled Fast Reactor
Corso di studi
relatore Ambrosini, Walter
relatore Forgione, Nicola
relatore Bandini, Giacomino
Parole chiave
  • simmer
Data inizio appello
Data di rilascio
Riassunto analitico
This thesis work, carried out at the Dipartimento di Ingegneria Civile ed Industriale (DICI) of the University of Pisa, concerns the thermo-hydraulic analysis of postulated accidents in a HLM cooled fast reactor, i.e. the MYRRHA-FASTEF reactor.
The first part of the work describes the general features and the historical background of the SIMMER-III code, such as the code assessment, and also a state of the art concerning the applications of the code to both separate effect facilities and full scale reactors. In this context, the SIMMER-III code can be adopted in analyses of sodium-cooled fast reactor, lead-cooled and LBE-cooled fast reactors; with some limitations and integration by additional models, it can be also applied to molten salt reactors and light water reactors.
The second part focuses on the description of the MYRRHA-FASTEF system and on its modelling by SIMMER-III, highlighting the adopted modelling of the different components. The reactor was simulated by a 2-Dimensional axial-symmetric geometry. The results of steady-state and transient calculations are then reported.
The steady-state analysis was performed in order to assess the correctness of the code and of the adopted model; so, the obtained results in relation to the major variables were compared with the design values. In particular, the most relevant results obtained for temperature trends and profile, both in the core and in the PHX, and the velocity and mass flow rate trends are reported. Significant thermal stratification is predicted by SIMMER-III in the upper plenum of the vessel which is responsible for temperature oscillation at the PHX inlet.
Finally, transient analyses were performed. Selected design basis condition transients and design extended condition transients were addressed in order to assure a sufficient safety level of the reactor, following postulated accidents or in unprotected transients. The sensitivity to the MOX density on fuel redistribution in the primary circuit has been also investigated. After fuel release, a certain amount of fuel particles is transported by the LBE coolant and, depending on the fuel porosity and the type of circulation, it tends to settle down or to float at the LBE free surface.