NuFuel & MMSNF 2015

First Workshop on Research into Nuclear Fuel in Europe
and Materials Modeling and Simulation for Nuclear Fuels Workshop
Karlsruhe, Germany, November 16th to 18th, 2015

Updated: Tue 08 Dec 2015, 14:27

Poster 4.2: Fuel chemistry of irradiated FBR mixed oxide fuel – Comparison between thermodynamic calculations and post-irradiation experimental results.

Jean-Christophe Dumas1, C. Guéneau2, T-N. Pham Thi1, N. Dupin3, B. Sundman4, J. Lamontagne1
  • 1: CEA, DEN, DEC, Centre de Cadarache, F–13108 Saint Paul Lez Durance Cedex, France
  • 2: CEA, DEN/DANS/DPC – Centre de Saclay – 91191 Gif-sur-Yvette Cedex, France
  • 3: Calcul Thermo – 63670 Orcet, France
  • 4: CEA, INSTN, Centre de Saclay - 91191 Gif-sur-Yvette Cedex, France


The mixed oxide fuel used in the Sodium-cooled Fast Reactors (SFR) undergoes high temperature level in the fuel pin with a large temperature gradient between the center and the periphery of the pellet. These extreme conditions are the cause of many thermal, mechanical and chemical transformations in the fuel. Moreover, the behavior of the fission products (FP) generated by the irradiation strongly depends on their chemical nature as well as conditions in the fuel (temperature, oxygen potential, …).

Experimental observations show the formation of a layer located between the fuel and the cladding material for a burn-up of 6 to 8 ha%. This layer, called JOG for “Joint Oxyde-Gaine” [1] is due to the release, migration and association of caesium, iodine, and tellurium mainly with oxygen and molybdenum created in the fuel pellet. At high burn ups, these fission products (FP) compounds can react with the cladding components (Fe,Cr,Ni) corroding the inner part of the cladding material. The fuel and cladding chemical interaction is nowadays recognized as one of the major factors determining integrity and lifetime of pins in the SFR [2].

For all these reasons, it is important to model the physico-chemical behavior of the fuel pin under irradiation conditions typical of FBR mixed oxide fuel. The modelling of mixed oxide fuel pins has been recently improved by CEA with the development of the GERMINAL V2 fuel performance code [3] where the nuclear FP database of the neutronics module has been chained to a thermodynamic code, ANGE, advanced version of the SOLGASMIX software. In addition, in conjunction with the OECD/NEA project [4/5], thermodynamic modelling of fission product systems such as Cs-Te, Cs-I and Cs-Mo-O … have been performed with the CALPHAD method and incorporated into the TAF-ID database.

Thermodynamic calculations on irradiated MOX fuels using ANGE code + TBASE on one side and by using Thermo-Calc and Open Calphad software [6] + TAF-ID database on the other side will be presented and compared to post-irradiation observations performed on (U0.78Pu0.22)O1.975 fuel samples irradiated to 13.6 ha% into the PHÉNIX reactor.

Next steps for the improvement of the thermodynamic description of the irradiated MOX fuel as well as the coupling between thermodynamics and the fuel performance code GERMINAL will be proposed and discussed.

  1. M. Tourasse, M. Boidron and B. Pasquet, J. Nucl. Mater. 188 (1992), 49
  2. K. Maeda, Comprehensive Guide Nucl. Mater. 3.16, Elsevier Ltd (2012), 443
  3. M. Lainet, V. Bouineau, T. Helfer and M. Pelletier, FR13, Paris, France, 4–7 March 2013
  4. C. Guéneau et al., J. Nuc. Mat. 419 (2011), 145
  5. C. Guéneau et al., J. Nuc. Mat. 344 (2005), 191
  6. B. Sundman et al., Integrating Materials and Manufacturing Innovation, 4:1 (2015), open access