The aim of this contribution is to deeply introduce the ALCYONE Fuel Performance Code.
ALCYONE is being implemented within the PLEIADES numerical framework and is …

Mixed oxides of uranium and plutonium U 1-y Pu y O 2-x are currently studied as reference
fuel for Sodium-cooled Fast Reactors (SFRs). To predict the margin to fuel melting, an …

The diffusion phenomena in uranium–plutonium mixed oxides U1− yPuyO2 dictate the
physicochemical properties of mixed oxides (MOX) nuclear fuel throughout manufacturing …

This paper presents simulations of the xM3 power ramp with the fuel performance code
ALCYONE performed during an international simulation exercise organized within the …

A combined experimental and modelling study into the Pb-Mo-O system has been
conducted in view of the safety analysis for lead-cooled nuclear systems. The thermal …

Currently, a knowledge gap exists in the available data and understanding of
thermophysical properties relating to fresh fuel salts, especially those containing plutonium …

This paper presents the results of the Power To Melt and Maneuverability (P2M) Simulation
Exercise on past fuel melting irradiation experiments, organized within the Organisation for …

In the context of the nuclear fuel recycle, U 1-y Pu y O 2±x Mixed Oxides (MOx) are currently
the most studied fuels. These materials have to meet accurate physico-chemical properties …

A new formulation for redistribution of constituents in metallic U-Pu-Zr nuclear fuel is
presented that can incorporate the contributions from fission products. The formulation is …

With dual goals of efficient and accurate modeling of solvation thermodynamics in molten
salt liquids, we employ ab initio molecular dynamics (AIMD) simulations, deep neural …