Fast-running tool dedicated to power excursion simulations in sodium-cooled fast reactor - CEA - Commissariat à l’énergie atomique et aux énergies alternatives Access content directly
Conference Papers Year : 2017

Fast-running tool dedicated to power excursion simulations in sodium-cooled fast reactor


Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) within the CEA (French Commissariat a l'Energie Atomique et aux Energies Alternatives) is involved, the innovative reactor designs under severe accidents conditions have to be assessed. Such accidents have been mainly simulated with mechanistic calculation tools (such as SAS and SIMMER). As a complement to these codes, which provide reference accidental transients, a new physico-statistical approach is currently developed at CEA; its final objective being to derive the variability of the main results of interest to quantify the safety margin. This approach involves fast-running tools to simulate extended accident sequences, by coupling models of the main physical phenomena with advanced statistical analysis techniques. They enable to perform a large number of simulations in a reasonable computational time and to describe all the possible scenario progressions of the hypothetical accidents. In this context, this paper presents a physical tool (numerical models and result's assessment) dedicated to the simulation of the beginning of the primary phase of the Unprotected Transient OverPower accidents (i.e. before large pin degradation).At the beginning of this primary phase, the increase of nuclear power induces a strong temperature gradient in the fuel pellets leading to specific mechanical behaviours, such as swelling and thermal expansion, before their meltdown. The fuel pin thermal evolution during slow power increases, such as control rod withdrawal accidents, and fast power increases have been performed. Validation on some CABRI experiments was carried out and focused on the axial distribution of melting points in the fuel column, the molten fraction at the end of the transient and the coolant temperature evolution during the transient. The results are consistent with experimental data and an application case on the CFV core of ASTRID was performed in order to give tracks for the safety of the reactor. In the future, mechanical and neutronics models will be developed.
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Dates and versions

cea-02434010 , version 1 (09-01-2020)


  • HAL Id : cea-02434010 , version 1


K. Herbreteau, N. Marie, F. Bertrand, J.-M. Seiler, P. R. Rubiolo. Fast-running tool dedicated to power excursion simulations in sodium-cooled fast reactor. 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH 17), Sep 2017, Xi'An, China. pp.95-108. ⟨cea-02434010⟩
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