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Tesi etd-09222009-103041


Thesis type
Tesi di laurea vecchio ordinamento
Author
ROZZIA, DAVIDE
URN
etd-09222009-103041
Title
Modelling of pellet-clad interaction during power ramp in water reactor fuel
Struttura
INGEGNERIA
Corso di studi
INGEGNERIA NUCLEARE
Supervisors
relatore Dott. Del Nevo, Alessandro
relatore Prof. Oriolo, Francesco
relatore Prof. D'Auria, Francesco
Parole chiave
  • Transuranus code
  • PCI
  • SCC
  • TU
Data inizio appello
12/10/2009;
Consultabilità
Parziale
Data di rilascio
12/10/2049
Riassunto analitico
The comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident conditions are part of the defense in depth concept. The research in the field involves the availability of experimental data and the development of advanced computational tools suitable for a reliable, best estimate simulation of the fuel behavior.
In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for models development and codes validation. This database includes the data sets of carried out at Studsvik in the framework of the Inter-Ramp BWR and Super-Ramp PWR Projects.
The phenomenon addressed in this activity is the pellet cladding interaction (PCI). In particular, special attention is given to the cladding failure during power ramps (stress corrosion cracking), caused by the combined effects of the clad stress occurring in the region of the pellets ends and the presence of aggressive fission products (e.g. iodine).
The first part of this work is focused in the descriptions of the phenomena occurring in fuel and cladding during normal irradiation and during power ramps in light water reactors.
In the second part the code TRANSURANUS version “v1m1j08” is assessed against the databases Inter-Ramp and Super-Ramp in order to verify the capability of the code in predicting the failures due to stress corrosion cracking and the associated phenomena prior and after ramps.
The results presented include the complete set of simulations of all rods irradiated in the Studvisk R2 reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of the boundary conditions implemented and the choice of the different code options on results.
Finally conclusions are presented with the aim to compare the results obtained from the two simulations.
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