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Fuel melting margin assessment of fast reactor oxide fuel pins using a statistical approach

Abstract : In the framework of the basic design of the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project, the design margins have to be defined with accuracy. The design criterion considered here is the fuel melting margin during nominal operation condition, which is given by the melting probability. Oxide fuel temperature and melting temperature are calculated with CEA dedicated fuel codes, a computation scheme coupling GERMINALV2 for entire fuel pin with a preliminary 3D thermomechanical model using LICOS for a fuel pellet. Results could be dependent on parameters like manufacturing processes, irradiation conditions and fuel behavior laws. The aim of this paper is to take into account uncertainties associated to these parameters in the melting margin evaluation and to quantify its sensitivity in order to reduce uncertainties. First step is the description of uncertain parameters by appropriate distribution. Uncertainty propagation is then done by using meta-models, a multi-linear regression and an artificial neural network. As a result, the melting margin is depending of the linear heat rate first, stoichiometry and initial gap after. Defect as the fuel pellet off-centering within the clad, which need a 3D thermomechanical model, has as well a non-negligible effect. In a last section, melting probability obtained by Monte-Carlo simulations is compared to FORM-SORM approximations. This preliminary study shows that results are in a good agreement for SORM method, and that the melting probability depends considerably on radial offsets.
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Submitted on : Wednesday, December 18, 2019 - 1:29:19 PM
Last modification on : Tuesday, April 28, 2020 - 11:28:15 AM


  • HAL Id : hal-02417756, version 1




V. Blanc, V. Dupont, T. Beck, T. Lambert, E. Thebaud, et al.. Fuel melting margin assessment of fast reactor oxide fuel pins using a statistical approach. IAEA-International Conference on Fast Reactors and Related Fuel Cycles Next Generation Nuclear Systems for Sustainable Development (FR17), Jun 2017, Yekaterinburg, Russia. ⟨hal-02417756⟩



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