Experimental and numerical study of fracture mechanisms in UO2 nuclear fuel

Abstract

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.

Damage characterization of (U,Pu)O2 under irradiation by molecular dynamics simulations

“…The advantage of this approach is that CZM eliminates singularity of stress at the crack tip. This approach has been used for nuclear fuels and implemented into fuel performance codes [33,34]. The pick stress on the cohesive law is the cohesive strength of the material, σ c , and corresponds to the point at which the crack initiates for a critical separation, λ f .…”

Multi-scale modeling of inter-granular fracture in UO2

“…However, the fracture properties are not correlated to the porosity, pore and grain size distributions in the underlying microstructure. Similar correlations are missing in the fracture studies made in [12][13][14]. In the present work, a multi-scale method is pursued to obtain the microstructural dependent fracture parameters of a smeared crack model [14,15] at the engineering scale.…”

“…The fracture modeling of UO 2 fuel pellets have primarily focused on thermo-mechanical fragmentation process during power cycles. Numerical techniques like discrete element method (DEM) [11], or finite element method (FEM) with cohesive zone [12] or smeared crack models [13,14] have been employed in these investigations. In DEM [11], the fracture properties are randomly distributed in the pellet to obtain the fragmentation behavior.…”

Modelling of fine fragmentation and fission gas release of UO2 fuel in accident conditions

“…As a first approximation, we have treated these local material properties as constants in the calculations and neglected any dependence on temperature or other parameters. The best-estimate value for G gb is 4 × 10 −3 Jm −2 , which is more than two orders of magnitude lower than the measured macroscopic (bulk) fracture energy of unirradiated polycrystalline UO 2 [147][148][149]. No measured data are available for the local fracture energy of grain boundaries, but molecular dynamics simulations indicate that it is comparable to that of bulk polycrystalline UO 2 in unirradiated conditions [150,151].…”