# Model materials for irradiated fuels: Study of local mechanical behavior using nanoindentation and microstructural analysis

Nathalie Payraveau-Le Roux, Sylvain Meille, Cyril Langlois, Isabelle Zacharie-Aubrun, Thierry Douillard, Jean-Marie Gatt

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Nathalie Payraveau-Le Roux, Sylvain Meille, Cyril Langlois, Isabelle Zacharie-Aubrun, Thierry Douillard, et al.. Model materials for irradiated fuels: Study of local mechanical behavior using nanoindentation and microstructural analysis. Microscopy and Microanalysis, 2014, 20 (Suppl. 3), pp.1832-1833. 10.1017/S1431927614010897. hal-01761575

HAL Id: hal-01761575

https://hal.science/hal-01761575v1

Submitted on 3 Nov 2025

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# Model materials for irradiated fuels: study of local mechanical behavior using nanoindentation and microstructural analysis

Nathalie Payraudeau - Le Roux¹, Sylvain Meille¹, Cyril Langlois¹, Isabelle Aubrun², Jean-Marie Gatt²

1. INSA-Lyon, MATEIS CNRS UMR5510, F-69621 Villeurbanne, France

2. CEA, DEN, DEC Cadarache, 13108 St Paul Lez Durance, France

Understanding and predicting nuclear fuels behavior during their use in reactors require knowledge of their mechanical properties such as elastic modulus, hardness and resistance to fracture. In this study, a method combining nanoindentation and EBSD analysis was developed on cubic zirconia (from Microcertec) and ceria (from Kurt Lesker), two model materials of irradiated UO₂. This approach aimed at overcoming the difficulty of making test-pieces for compression or flexion tests in fragmented and cracked irradiated fuels.

The model materials were chosen for their crystalline structure, microstructure and mechanical properties close to UO₂ ones. The yttria fully-stabilized cubic zirconia used in the study has a microstructure (grain size, porosity) similar to the UO₂ one while fracture toughness of ceria is closer to the UO₂ one and lower from fracture toughness of cubic zirconia, thus allowing the observation of crack propagation at lower loads.

Nanoindentation tests are performed inside grains and on grain boundaries at loads high enough to induce cracks. SEM observations enable determination of the threshold load to initiae cracks, and of the types of cracks obtained. These results show a similarity with the fracture behavior of irradiated UO₂ (as illustrated in Figure1).

SEM images are coupled with EBSD analysis in order to determine the influence of grain orientation and preferential cleavage planes (as illustrated in Figure2).

Indentation tests are also performed at higher loads to study the influence of microstructural heterogeneities such as pores and grain boundaries and in order to link local mechanical properties of the material to more macroscopic results. The range of loads used in this study is ranging from 10 mN to 10 N.

Finally, instrumented indentation is used to assess fracture energy, as classical models using crack lengths for fracture toughness calculation [1-2] are not applicable in a large range of loads in the tested materials. Fracture energy is an important input for numerical simulations of irradiated fuel behavior. The method developed here will then be implemented on nuclear fuels.
References:

[1] Anstis, GR *et al*, (1981), Journal of the American Ceramic Society, 64:533-538.

[2] Dukino, R. D *et al*, (1992), Journal of the American Ceramic Society, 75: 3299-3304.

<figure><img src="image_2.png" /><figcaption>**Figure 1.** Vickers indentation at 661 mN in cubic zirconia (left) and at 735 mN in irradiated $UO_2$ (right).</figcaption></figure>

<figure><img src="image_3.png" /><figcaption>**Figure 2.** Berkovich indentation at 300 mN in cubic zirconia. Superposition of SEM image and EBSD cartography.</figcaption></figure>