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Tesi etd-01152016-163055


Thesis type
Tesi di laurea magistrale
Author
FACCHINI, ALBERTO
URN
etd-01152016-163055
Title
Analysis of the European Sodium Fast Reactor core under an unprotected transient overpower
Struttura
INGEGNERIA CIVILE E INDUSTRIALE
Corso di studi
INGEGNERIA NUCLEARE
Commissione
relatore Dott. Giusti, Valerio
relatore Dott. Ciolini, Riccardo
relatore Dott. D'Agata, Elio
Parole chiave
  • nuclear engineer
  • sodium reactor
  • power pin analysis
  • neutronic analysis
  • MCNP
Data inizio appello
22/02/2016;
Consultabilità
completa
Riassunto analitico
This work is based on the European Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). CP-ESFR was initiated as part of the EURATOM contribution to the GIF and as an attempt to create a common European framework to support the sodium fast reactor technology. It was launched within the 7th EURATOM Framework programme and 24 European partners are assigned the objective to establish the technical basis of a European sodium fast reactor plant with improved safety performance, resource efficiency and cost efficacy.
The design and safety analysis of current and future nuclear reactors, during normal operation and under accidental condition, requires continuous improvement of computational accuracies. To this purpose, multi-physics approaches including detailed, coupled neutronic and thermal hydraulic assessments are being increasingly developed and used. In this work, a further step in the neutronic analysis of the European Sodium Fast Reactor (ESFR) is attempted.
This thesis work was developed with the collaboration of the Joint Research Centre – Institute for Energy and Transport (JRC – IET) and it was a follow up of previous studies.
In particular this study is focused on the research of the power peaking fuel pin at end of cycle (EoC) after the insertion of a specific value of reactivity.
The Monte Carlo code MCNP6 has been used for all the calculations here presented. This code differs from its predecessors being the first which integrates all the features of MCNP5 and MCNPX providing, therefore, the capability to perform burn up calculation with the depletion code CINDER90.
The MCNP6 input file of the reactor, developed at JRC – IET, represents the whole core at the beginning of life (BoL) condition and as such it could not be directly used to achieve the aforementioned goal of this study. Thus, it has been necessary to work on it in order to search, first, for the position of the control rods corresponding to the critical reactor (the original input file was not representing such a reactor state), hence, by using the new burnup card, to get the material composition of the fuel at the EoC, the breeding ratio and some other safety parameters before to start the fuel assembly and fuel pin analysis to identify the power peaking fuel pin.
The insertion of reactivity was done in three different ways:
1. Partial withdrawn of all the control rods.
2. Complete extraction of a selected control rod.
3. Complete extraction of two selected control rods.
This choice was done to make a comparison between a global insertion of reactivity and a local insertion of reactivity.
Finally, the detailed power profile within the hottest fuel pin will be used as input in TRANSURANUS or a finite element code (e.g. Ansys or Abaqus) to assess the mechanical safety margin of the cladding during accidental conditions.
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