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Abstract:In this paper we study the brittle behavior of UO 2 nuclear fuel. First, we present the interpretation of bending tests with three different approaches to evaluate rupture parameters (critical stress and toughness). Second, we present Vickers indentation tests on fresh UO 2 fuel. The comparison between bending and indentation tests on fresh fuel allows us to evaluate the constant parameters relative to indentation tests. Vickers indentation is then used to evaluate rupture parameters of irradiated fuels. At the end, we present some applications to fuel rod modeling taking into account the different rupture mechanisms. IntroductionThis paper deals with the rupture of UO 2 fuel in a nuclear reactor. At low temperature (up to 900 °C), this ceramic is brittle. We show that the fracture network has an important impact on the loading of clad due to pellet expansion (thermal expansion and swelling). The goal of this article is to present a methodology to study this problem. The measurement of rupture parameters (ultimate stress and toughness) is therefore important to have a good simulation of pellet cladding interaction in different operating conditions. For that, we must develop a methodology to evaluate rupture parameters on fresh fuel and irradiated fuel. This paper is divided into three parts. The first part shows how we measure the rupture parameter using bending tests. An interpretation of these tests using modeling is necessary to identify the real different parameters of the models (cohesive zone model and DDIF2 [1]). In the second part of this paper we present measurements by Vickers indentation. The interest in these tests is that it can be used on irradiated fuel. The last part of the article, describes the modeling used to simulate a fuel rod taking into account the different rupture mechanisms.

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