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International audience; This paper presents the coupling of the thermochemical solver ANGE (Advanced Gibbs Energy Minimizer) with an oxygen thermo-diffusion model. The coupling is implemented within the fuel performance code ALCYONE co-developed by CEA, EDF and FRAMATOME within the PLEIADES environment. An application to a 3D simulation of a power ramp on a Cr-doped UO 2 fuel is developed. Post-ramp EPMA measurements of chromia doped fuel show reduction of chromium and molybdenum oxides in the central part of the pellet, indicative of thermo-diffusion of oxygen. These phenomena are well reproduced by the coupled thermo-chemical-mechanical simulations. Impact of oxygen redistribution on speciation of fission products is then studied. Chemical state of caesium, iodine and tellurium is important as regard PCI, as they can form gaseous species (CsI$_{(g),}$ I $_{(g),}$ I$_{(2g),}$ TeI$_{(2g)}$) that can react with the cladding and induce SCC. Release of gaseous species and concentration of chemically reactive iodine compounds near the cladding are calculated in order to investigate I-SCC.

International audience; A speciation and release kinetics study of the gaseous fission products freed fromnuclear fuels in conditions of severe accident (temperature range 1300°C –2500°C) as a function of the applied thermodynamic conditions is carried out usingthe experimental feedback collected at CEA within the VERCORS and VERDONprograms. The nature of the major species involved are deduced from the analysis of thespecific features of the Cs-Mo-Ba-I-O chemical system in the composition range ofirradiated nuclear fuels. Stable gaseous Cs and Ba molybdate species, i.e. $Cs_2Mo_2O_7$ (g)and $BaMoO_4 (g)$, clearly predominates in oxidising conditions, whichcontributes to increase significantly the released fraction of molybdenum ascompared to reducing conditions.An analytical model accounting for thermo-chemistry in the release rate ofchemically reactive elements is applied to estimate the release of Cs, I, Mo and Bain severe accident conditions. Good agreement is achieved with on-linemeasurements

International audience

International audience; The ALCYONE multidimensional fuel performance code co-developed by the CEA, EDF and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial Pressurized Water Reactors (PWRs), power ramps in experimental reactors or accidental conditions such as Loss Of Coolant Accidents (LOCAs) or Reactivity-Initiated Accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms as close to physics as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, its development and validation shall include tests on PWR-UO 2 fuel rods with advanced claddings such as M5® under " low pressure-low temperature " or " high pressure-high temperature " water coolant conditions. This paper first presents the ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to non steady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This paper then compares some simulations of RIA transients performed on UO 2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE – starting from base irradiation conditions it itself computes – is currently able to handle both the first stage of the transient, namely the Pellet Cladding Mechanical Interaction (PCMI) phase, and the second stage of the transient, should the boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5® is a trademark or a registered trademark of AREVA NP in the USA or other countries.

International audience; The ALCYONE multidimensional fuel performance code co-developed by the CEA, EDF and AREVA NPwithin the PLEIADES software environment models the behavior of fuel rods during irradiation incommercial Pressurized Water Reactors (PWRs), power ramps in experimental reactors or accidentalconditions such as Loss Of Coolant Accidents (LOCAs) or Reactivity-Initiated Accidents (RIAs). Asregards the latter case of transient in particular, ALCYONE is intended to predictively simulate theresponse of a fuel rod by taking account of mechanisms as close to physics as possible, encompassingall possible stages of the transient as well as various fuel/cladding material types and irradiationconditions of interest. On the way to complying with these objectives, its development and validationshall include tests on PWR-UO2 fuel rods with advanced claddings such as M5 under low pressurelowtemperature or high pressure-high temperature water coolant conditions.This paper first presents the ALCYONE V1.4 RIA-related features and modeling. It especially focuses onrecent developments dedicated on the one hand to non steady water heat and mass transport and onthe other hand to the modeling of grain-boundary cracking-induced fission gas release and swelling.This paper then compares some simulations of RIA transients performed on UO2-M5 fuel rods inflowing sodium or stagnant water coolant conditions to the relevant experimental results gained fromtests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. Itshows in particular to what extent ALCYONE starting from base irradiation conditions it itselfcomputes is currently able to handle both the first stage of the transient, namely the Pellet CladdingMechanical Interaction (PCMI) phase, and the second stage of the transient, should the boiling crisisoccur.Areas of improvement are finally discussed with a view to simulating and analyzing further tests to beperformed under prototypical PWR conditions within the CABRI International Program.

International audience; The French Alternative Energies and Atomic Energy Commission (CEA) and its industrial partners electricite de France (EDF) and AREVA have been developing the collaborative platform PLEIADES for more than 10 years. This platform now supports the development of several state-of-the-art fuel performance codes Alcyone and Cyrano for Pressurized Water Reactor (PWR), Germinal V2 for Sodium Fast Reactors (SFR), Licos for innovative fuel elements and experimental irradiation devices, MAIA for Material Testing Reactors (MTR), etc.This paper first provides an up-to-date overview of the PLEIADES platform and then focuses on recent improvements of the thermomechanical modelling abilities recently introduced in PLEIADES's fuel performance codes. Various topics will be discussed and illustrated using Alcyone simulations of the PWR fuel rod in normal and off-normal situations, including LOCA and RIA- PWR fuel pellet cracking during the reactor start-up;- description of grain boundary decohesion in the fuel oxide during a RIA transient;- multi-fragments modelling of the PWR in 2D(r; );- various numerical improvements ;- finite strain modelling in 1D.

International audience

International audience; This paper presents 2D generalized plain strain simulations of the thermo-mechanical response of a pellet fragment and overlying cladding during a RIA transient. A fictitious hydride blister of increasing depth (25 to 90% of the clad thickness) is introduced at the beginning of the calculation. When a pre-determined hoop stress is exceeded at the clad outer surface, radial cracking of the blister is taken into account in the simulation by a modification of the mechanical boundary conditions. The hoop stress criterion is based on Finite Element simulations of laboratory hoop tensile tests performed on highly irradiated samples with a through-wall hydride blister. The response of the remaining clad ligament (beneath the cracked blister) to the pellet thermal expansion is then studied. The simulations show that plastic strains localize in a band orientated at ~ 45° to the radial direction, starting from the blister crack tip and ending at the clad inner wall. This result is in good agreement with the ductile shear failures of the clad ligaments observed post RIA transients. Based on a local plastic strain failure criterion in the shear band, ALCYONE simulations are then used to define the enthalpy at failure in function of the blister depth.

International audience; This paper presents finite elements analyses of the lift-off experiments performed in the HALDEN reactor: IFA-610.3/.5 for UO$_2$-Zy4 fuel rods, IFA-610.2/.4 and IFA610.7 for MOX-Zy4 fuel rods. The 1D and 2D($r$,$\theta$) schemes of the multi-dimensional fuel performance code ALCYONE are both used to study the overpressure conditions leading to the onset of temperature increase in the experiments. The 1D scheme is based on a rather standard axisymmetric description of the complete fuel rod discretized axially in slices. The 2D($r$,$\theta$) scheme allows one to study the plane strain thermo-mechanical behaviour of a pellet fragment (usually 1/8 th of the complete pellet) and its contact with the overlying cladding bore. It accounts explicitly for the additional free surface associated to pellet radial fractures and provides an estimation of the evolving pellet crack opening during loading sequences. In the proposed application to lift-off experiments, the impact of overpressure applied on the radial pellet crack borders has been studied. In the first part of this paper, the main features of ALCYONE 1D and 2D($r$,$\theta$) modelling schemes are presented. In the second part, simulations of the lift-off experiments performed with the ALCYONE 1.4 release are presented (in particular this release allows changing the nature of the filling gas in order to assess its impact on the fuel thermal behaviour). Generally, for the lift-off experiments simulated with ALCYONE code 1D scheme, a rather good agreement is obtained between predicted and measured temperature evolutions and rod axial elongations, especially when a clad-pellet bonding hypothesis is retained. Since the same material models are used in 1D and 2D, a good agreement with the measured temperature is also obtained from the 2D simulations. It is however shown that the application of the overpressure on the radial pellet crack borders has a strong impact on the onset of pellet-clad gap reopening. The resulting tangential stressing of the pellet fragment leads to radial fuel creep which tends to increase the external radius of the pellet and hence delay reopening with respect to 1D simulations results. The "mechanical lift-off" is thus better estimated when the pellet fragmentation is considered in the simulations.