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Tesi etd-11122017-182529


Tipo di tesi
Tesi di laurea magistrale
Autore
ANDRELLO, CONCETTINA
URN
etd-11122017-182529
Titolo
Post irradiation examination and characterization of nuclear fuel CERMET samples
Struttura
INGEGNERIA CIVILE E INDUSTRIALE
Corso di studi
INGEGNERIA NUCLEARE
Commissione
relatore Dott.ssa Lo Frano, Rosa
relatore Dott. Freis, Daniel
Parole chiave
  • Nuclear waste management
  • Nuclear transmutation
  • Nuclear fuel Image Analysis
  • CERMET fuel
  • FUTURIX-FTA
  • Post Irradiation Examination
Data inizio appello
27/11/2017;
Consultabilità
parziale
Data di rilascio
27/11/2020
Riassunto analitico
The main challenge of new reactors is the reduction of the radiological impact of nuclear waste on the environment through a sustainable fuel cycle.
Currently, most of the high-level nuclear waste comes from Light Water Reactors (LWR), and a possible strategy is to close their fuel cycle through recycling of uranium and plutonium and transmutation of the minor actinides, such as americium and curium, in dedicated fast reactors. Under the assumption of sufficiently small losses during the fuel reprocessing, it would be possible to reach a complete closure of the fuel cycle, resulting in a high reduction of the actinide content of the high-level waste (HLW) and consequently in the reduction of the long-term radiotoxicity. Only short-lived fission products remain to be stored in a repository, which would decay within a few hundred years. Such a strategy would support the feasibility of licensing and public acceptance of a final repository significantly.
Dedicated fuels are required in order to reach efficient transmutation in nuclear reactors. In this context, one of the possible strategies to transmute Minor Actinides (MAs) and long-lived fission products is to incorporate them in a matrix of neutron physically inert material. Dedicated transmutation fuels are typically uranium free, in order to maximize transmutation efficiency.
A substantial input to the development of advanced transmutation fuels and the understanding of their behaviour under irradiation has been given by an experimental test called FUTURIX-FTA (Fuel for Transmutation of transURanium elements in phenIX/Fortes Teneursen Actinide), which has been conducted in the Phénix reactor (Marcoule, France) during its last operation cycles.
The FUTURIX-FTA experiment was performed in the framework of an international collaboration between CEA, DOE, JRC Karlsruhe (the former Institute for Transuranium Elements) and JAEA, and it involved a number of different fuels dedicated to high minor actinide content incineration. Eight innovative fuel forms were tested, and these were metallic, nitride, CERCER and CERMET fuels, the latter two being based on ceramic fuels particles embedded in ceramic or metallic matrix, respectively. The fuels have been prepared by different laboratories (INL, LANL, CEA and JRC-Karlsruhe). In particular, two different CERMET fuel, called FX-5 and FX-6, were fabricated at JRC-Karlsruhe where the Post Irradiation Examination corresponding to the experimental work of this thesis was performed.
This thesis aims to pave the way for the understanding on the performance under irradiation of CERMET fuel for transmutation, focusing on the main factors that affect the fuel behaviour under irradiation. The main safety aspects, which are required for the implementation of new fuel in the reactor have been investigated. In order to reach a high level of confidence with the behaviour of these nuclear fuels, experimental together analytical approaches have been used.
The experimental results have been obtained by using fully qualified apparatuses in JRC-Karlsruhe hot cells facility. The apparatuses because of their high reliability allow high accuracy and replicability of the results.
The work was divided into two parts: (I) experimental measurements of thermo-physical properties (density and thermal diffusivity/conductivity) and ceramography examinations; (II) study and analysis of experimental results. In order to help and improve the interpretation of the measurement results the use of Image Analysis (IA) technique and the development of a Finite Element Method (FEM) Modell were required.
For the experimental part, standard procedures in preparation of irradiated samples and investigation measurements of irradiated fuel were adopted. In particular, the work was focused on an in depth examination of density, thermal properties and macro and microstructure of the fuels which are the main factors influencing the performance of nuclear fuel under irradiation.
Considering the most important safety margins, which must be fulfilled during nuclear reactor operation, this work was intended to investigate following aspects:
• Thermo-physical properties after irradiation;
• Maximum temperature reached inside the pellet;
• Mechanical and geometrical integrity of both matrix and particles;
• Swelling behaviour of both matrix and particles;
• Fission gases released.
The analysis of experimental results led to the following findings. The inert matrix maintained mechanical and dimensional stability, showed absence of damage and cracking.
As expected because the high thermal conductivity of the Mo, the temperature reached inside the pellets was “cold” in comparison with standard ceramic fuel, then showing a flat temperature profile.
Decreasing in density has been measured and it is consistent with the swelling showed by the pellet. In addition, the comparison between the geometrical dimension of fresh and irradiated fuel showed the occurrence of a non-isotropic swelling of pellets, with an axial elongation higher than radial elongation. This special behaviour was attributed to the elliptical shape of ceramic particles.
The measurements of the ceramic particles lead to the conclusion that swelling of fuel pellets was only caused by swelling of ceramic particles, which increasing in volume due to the high gases released during the irradiation.
In addition, the maximum measured diameter of the pellet was lower than the inner clad diameter, leading to the absence of gap closure and it implies not pellet cladding interaction occurred during irradiation.
Finally, the fission gases released (measured by gas puncturing technique) have been connected to the ceramic particles closed to the external surface.
From these results, a good behaviour of CERMET fuel under irradiation have been detected and some of safety requirements are completely satisfied.
Moreover, further examinations have been planned, in order to reach a comprehensive understanding of the CERMET behaviour under irradiation.
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