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Digital archive of theses discussed at the University of Pisa


Thesis etd-04212011-175113

Thesis type
Tesi di dottorato di ricerca
email address
Thesis title
Methodology for the Analysis of Fuel Behavior during LOCA and RIA
Academic discipline
Course of study
tutor Prof. D'Auria, Francesco Saverio
relatore Dott. Vitanza, Carlo
relatore Dott. Del Nevo, Alessandro
relatore Prof. Oriolo, Francesco
  • Licensing
  • LOCA
  • Nuclear Fuel
  • RIA
Graduation session start date
Loss of coolant accidents (LOCA) mean those postulated accidents that result from the loss of reactor coolant at a rate in excess of the capability of the reactor coolant makeup system from breaks in the reactor coolant pressure boundary, up to and including a break equivalent in size to the double-ended rupture of the largest pipe of the reactor coolant system (DEGB LB-LOCA).

A peculiarity of the Atucha-2 design is the positive void reactivity coefficient. This is a characteristic in common to other heavy water moderated reactors that utilize natural uranium as fuel. This implies that after a LB-LOCA event, the fission power peak at the very beginning of the transient is controlled by the void formation in the core channels, and then it is determined by the pressure wave propagation from the break. Indeed, the moderator is still liquid and flashes delayed with respect to the coolant, thus the LOCA event is also a RIA (reactivity insertion accident) event.

Licensing requirements vary by country in terms of their scope, range of applicability and numerical values and in general imply the use of complex system thermal hydraulic computer.

Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes, as for burst temperature, burst strain and flow blockage calculations. This may imply the use of a dedicated fuel rod thermo-mechanical computer code such as TRANSURANUS, which can be coupled with thermal-hydraulic system codes to be used for the safety analysis.

This thesis consists in the development and application of a methodology for the analysis of the 2A LB-LOCA scenario in Atucha-2 nuclear power plant (NPP), focusing on the procedure adopted for the use of the fuel rod thermo-mechanical code and its application for the safety analysis of Atucha-II NPP (Chapter 15 FSAR). The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient.

The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions (BIC) (e.g. pin power axial profiles) are provided by core physics and three dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method.

Validation activities are performed to enhance the TRANSURANUS code capabilities and to improve the reliability of the code results. They relies on the two main sources of data, namely, specific data on Atucha-1 and/or 2, experiments or independent calculations and other data which are representative of the Atucha-2 fuel, in particular for the analysis of the normal operation and power ramp during normal operation and severe transients like LOCA and RIA.