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International audience; A numerical FEM (Finite Element Method) model was implemented to represent the mechanical state of the wooden panel of the Mona Lisa, as it is conserved in its exhibition case, and constrained in its auxiliary frame. The model is based on the integration of advanced numerical analysis and various experimental examinations carried out non-invasively on the artwork by the authors during over 15 years. This includes visual, microscopic and X-ray observations together with mechanical measurements and monitoring of panel deformations and constraining external forces. In addition to the development of non-invasive techniques to characterise the mechanical properties of the panel, the FEM model reliably evaluated the strains and stresses generated in the panel by the various actions it experiences. The paper consists of the following parts: (i) a short summary of the experimental measurements and other observations, (ii) a detailed description of the FEM numerical model, of the hypotheses it is based on, and of its advantages and limits, (iii) the main results obtained by running the model. This includes the identification of local strains and stresses, the location of most critical areas, an evaluation of the risk that the existing ancient crack may propagate, and an evaluation of safe ranges for the forces acting on the wooden panel, (iv) the validation criteria for such results, and (v) a discussion about the significance of the mechanical model.

International audience; This article describes the MFrontInterface.jl package which aims at allowing mechanicalbehaviours generated by the MFront code generator to be used in JuliaFEM. MFrontInterface.jl isbuild on top of the MFrontGenericInterfaceSupport library.

International audience

International audience; In tokamak environment, plasma facing components made with tungsten as armor material will reach extreme temperature value up to 2000°C under stationary heat loadings. At such temperature, tungsten recrystallization and creep phenomenon become non negligible, this is why the particular creep behaviour of recrystallized ITER-grade tungsten is of interest. Previous study performed experiments on rods tungsten grades up to 1300°C and revealed that creep behavior of material is strongly dependent on chemical composition and manufacturing process. In order to fulfill the literature database, 3-points bending tests are performed from 1400°C to 1600°C on rolled tungsten, a potential candidate for future divertor. Experiments are achieved using thin rectangular samples of recrystallized tungsten. Tests are carried out at CEA Cadarache in a device usually used to study ceramics such as uranium dioxide. The purpose of these tests is, in-fine, to identify a creep material law for tungsten that may be used in fatigue analysis to describe the in-service behaviour of the component. In order to identify such a law which is dependent upon stress, strain rate, temperature and the material microstructure, constant crosshead speed experiments have been carried out under different conditions. Using a finite element model to describe the 3-point bending device, an attempt is made at modelling the creep behaviour of ITER-grade tungsten using a kinematic hardening law. Frederik-Armstrong law is used and numerical results obtained highlight that law parameters are well identified. In addition to the macroscopic bending tests, complementary SEM/EBSD examinations are carried out with the aim of identifying specific microstructural mechanisms based on the former works on uranium dioxide

International audience; Future GENIV sodium-cooled fast reactor (SFR) may include Americium (Am) recycling in the radial fertile blankets positioned in the outer core regions. An issue related to this option is the lack of knowledge regarding the behaviour under irradiation of the Am-bearing blanket (AmBB). The DIAMINO separate-effect experiment was irradiated in the OSIRIS Material Test Reactor (MTR) with the goal of studying AmBB gaseous release and swelling as a function of temperature, blanket microstructure and gas production rate.A brief description of the experiment is first provided, in terms of objectives and planning. The innovative design developed for the DIAMINO experiment is described in details. The DIAMINO irradiation time history is then summarized and the first results of non-destructive post-irradiation examinations are given.The thermo-mechanical modelling of the DIAMINO experiment device using the LICOS code is then detailed, together with the validation of the computations on the basis of some experimental temperature measurements performed on an in-pile mock-up of the experiment. The use of the simulation tool for the monitoring of the AmBB sample temperature under irradiation and for the post-irradiation calculations is explained and illustrated through some examples.

International audience; Concrete exhibits visco-plastic behaviour when subjected to high temperature whilst supporting an external load; this component of plasticity is referred to as load induced thermal strain (LITS), which is also known as transient thermal creep (TTC). LITS phenomena can be important in pre-stressed concrete structures, because the evolution of LITS can potentially lead to a loss in pre-stress, and also residual tensile stress development under transient thermal conditions. Hence, structures which have been subjected to a high temperature thermal loading cycle may experience cracking and subsequent loss of rigidity; this could ultimately lead to a loss of functionality or even structural failure. Whilst models for LITS have only recently become available in the public domain, there are currently no available models that can capture cracking with plasticity, to an acceptable industrial reliability. Indeed, most damage models available in commercial packages do not adequately capture anisotropy and plasticity in addition to the unilateral effect exhibited by concrete, i.e. the Mazars damage model. However, recent research has led to the development of the Fichant-La Borderie (FLB) damage model, which has shown that these additional effects can be captured using an energy regularisation method; this makes the FLB model suitable to capture cracking due to LITS effects. In this project a method for coupling LITS behaviour with a FLB model is proposed. The constitutive behaviour is coupled to the damage law via the total stress tensor. By decomposing the visco-plastic and thermal strains, the damage evolution is driven from the elastic strains only. Finally, the effective stress is computed and used to obtain the total stress at each iteration with unilateral effect included. This type of model is most suitable for fully implicit integration schemes and will allow more accurate assessment of structural damage from transient thermal events, such as fire.

International audience; L’étude du comportement viscoplastique du dioxyde d'uranium par des essais de compression uniaxiale a fait l'objet de nombreux travaux au CEA. L'objet de ce travail est de présenter des progrès récents accomplis sur la maîtrise des conditions expérimentales (contrôle de la pression partielle d’oxygène) d'une part et sur la simulation et l’interprétation des essais d'autre part.La maîtrise accrue des paramètres expérimentaux a permis d’améliorer la reproductibilité des essais et confirme le role majeur de la pression partielle d’oxygène sur le comportement viscoplastique du dioxyde d'uranium . La simulation des essais, basée sur les outils développés au sein de la plate-forme numérique PLEIADES de simulation du comportement des combustibles, est l'objet principal de la présentation. La simulation du dispositif expérimental a été améliorée et se base sur un cadre physique (lois en grandes transformations) et numérique rigoureux. Les simulations permettent de proposer différentes modélisations du dispositif expérimental complet, allant d'une simulation 1D suffisante pour une identification rapide des lois de comportement a des simulations multi-dimensionnelles permettant de décrire des effets de structure (mise en tonneau des pastilles) et les différents éléments du dispositif expérimental (cales, colonne). La modélisation fine de référence donne accès a une compréhension plus fine de l'essai, des biais associés et in fine du comportement du matériau.

International audience; Controlling and predicting the high temperature mechanical behaviour of nuclear oxide fuels is essential to guaranteeing the integrity of fuel rods during normal or incidental operating conditions. Although in the past many studies have been devoted to the creep behaviour of uranium dioxide, few have focussed in detail upon the effect of non stoichiometry. In those which have dealt with this issue, strain rate was usually considered as a function of deviation from stoichiometry although the true intensive thermodynamic variables are the chemical potential, temperature and stress.In this work we describe very recent developments involving a high temperature compression creep furnace which has been equipped with a system enabling the control and measurement of the oxygen activity in the gas phase. This guarantees the oxygen activity in the solid in equilibrium with it. We report the first creep experiments carried out under these controlled conditions. We also discuss the development of a material model and associated kinematic hardening behaviour law capable of reproducing the near constant strain rate experiments carried out. Although we show that the data may be interpreted in terms of uniaxial loading, due consideration is given to multi-dimensional effects. The physical significance of the material behaviour law parameters is presented, particularly when the parameter is sensitive to oxygen activity. Our ultimate aim is to relate internal variables of the model to the local microstructure. Post test Electron BackScatter Diffraction is shown to be a prospective technique for providing this information and based on it, parallels can be drawn between the mechanisms via which the material accommodates mechanically induced strain and its response to radiation damage.

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; Fixed point iterations are still the most common approach to dealing with a variety of numerical problems such as coupled problems (multi-physics, domain decomposition,...) or nonlinear problems (electronic structure, heat transfer, nonlinear mechanics, ...). For coupled problems, this partitioning approach has the main advantage to enable the use of efficient numerical solvers dedicated to each physic. However fixed point iterations converge generally only linearly and veryslowly.Methods to accelerate fixed point iteration convergence or more generally sequence convergence have been extensively studied since the 1960's. For scalar sequences, the most popular and efficient acceleration method remains the $\Delta^2$ of Aitken. Various vector acceleration algorithms are available in the literature, which often aim at being multi-dimensional generalizations of the $\Delta^2$ method.\\ In this presentation, we propose and analyze a generic residual-based formulation for accelerating vector sequences~\cite{RamiereHelfer}. We show that through this formulation, various existing acceleration algorithms for vector sequences are recovered~\cite{brezinki-zaglia91} and many other ones can be generated. \\ The question of the dynamic use of this residual-based transformation during the fixed point iterations for obtaining a online accelerated fixed point method is then raised. We show that two main classes of dynamic iterative algorithms can be derived. \\ We show the efficiency of such acceleration algorithms on two kinds of tests. First, we compare some of the residual-based acceleration algorithms in the field of nonlinear mechanics. In particular, well-chosen accelerated algorithms combined with a quasi-Newton method based on the elastic operator are prove to be competitive with respect to the standard second-order Newton-Raphson algorithm. Then, we present some numerical results obtained using acceleration techniques on a multi-physics thermal-mechanical-physical chemistral problem derived from the simulation of nuclear fuel behavior~\cite{Michel2013} and solved thanks a Gauss-Seidel fixed point algorithm.

International audience; This paper is devoted to the application of the phase field method to the case of damage in con-crete, in the particular framework of multiscale mechanics using 3D representative elemen-tary volumes (REVs) at the mesoscale. More specifically, this work consists in the treatment of instabilities in the crack propagation modeling in isotropic elastic solids, which may arise due to the dependency of Miehe scheme with respect to the choice of time step. Our al-gorithm allows for modeling crack propagation within complex microstructures, in a robust manner that is independent of the choice of time step. Several cases of concrete mesostruc-tures exhibiting different aggregate shapes have been analyzed under classical tension and compression loadings.

International audience; This paper reports on preliminary studies performed regarding the insertion reliability of theRBC-type Control Rods designed for the ASTRID Sodium-cooled Fast Reactor. At this stage, the primary aim of the analysis is to evaluate the mechanical behavior of RBC Control Rods under Emergency Shutdown conditions, for which reactor core structures are subjected to significant misalignments (including earthquakeimposed displacements). Using a Finite Element Model based on the Cast3M solver and developed specifically for these studies, computations are performed that allow assessing contact reactions (and the associated friction forces and contact pressures), deformations and stresses (mostly due to bending-induced deformations) which are considered for design. Based on these preliminary results, some optimization of the Control Rod design is proposed that ensures some stable behavior all along the rod drop, with substantial design margins

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.

National audience

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; 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; Itemising ASTRID SPECIFICATIONS  ENHANCED SAFETY  lessons learned from previous fast reactors operations  structural materials and guidance surface coatings/hard-facings  COST SAVINGS  reduce absorber subassembly length  extend residence time to match the residence cycle and, if possible, life span of the absorber elements with those for the fuel cycles (4 runs of 360 or 400 EFPD each)  reduce the total number of absorber subassemblies  keep the enriched absorber material requirement as low as possible  2 distinct and independent fast-acting reactor shutdown systems  Additional core safety features/systems as potential options based on passive insertion of negative reactivity in the core  Designed absorber subassemblies:  RBC subassembly (power regulation/compensation for reactivity change during lifetime/normal or emergency shutdown)  RBS subassembly (emergency shutdown)  SEPIA* subassembly (*abbreviation for SEntinal for Passive Insertion of Antireactivity) { R&D to support design options  MAIN QUALIFICATION ISSUES:  for RBC, RBS and SEPIA absorber element R&D: shroud behaviour, structural corrosion (cladding and/or shroud), absorber thermal behaviour; common thread: use the most durable materials possible  for RBC and RBS subassembly R&D: behaviour of hard-faced parts implicated in guidance problems  for SEPIA subassembly: actuation technology  POST-IRRADIATION EXPERIMENTS on already irradiated objects  cf. " Phénix Treasures " project  R&D ACTIONS on shroud behaviour and cladding/shroud corrosion  R&D ACTIONS on absorber materials:  " carbothermal " B 4 C  improve phenomenological modelling and formalisation of its irradiation behaviour  HfB 2 and TiB 2  changes in properties and behaviour under irradiation:  promising initial properties: high and distinctly metallic thermal conductivity; high refractoriness  the most capable of meeting the operating requirements for SEPIA  Hf metal (robust/durable alternative to boron materials + mitigation potential)  IRRADIATION outlook:  hard-facing materials for guidance parts exposed to neutron flux  absorber elements with representative design FOCUS on SEPIA preliminary thermohydraulic/hydraulic analysis RBC & RBS absorber element-CONCEPTS & DESIGN  Two coupled domains…  Fuel SA pin bundle + outside and central wrappers  Spherical absorber elements + capsule  … treated using two project codes  CADET  LICOS/Pléiades  Thermohydraulic feasibility OK  B 4 C is thus far no suitable material solution  Reference option: vented pin + B 4 C pellets + tubular shroud  Alternative (materials & design) options: HfB2/TiB2/Hf; splitted double shroud; single or double-donut shaped absorber column  Thermomechanical calculations  using LICOS/Pléiades project code  to check matching of criteria* * (no absorber or shroud melting; no ACMI; cladding integrity)  to assess the accessibility of higher performances 

Smirt 23 - Division V: Modelling, Testing and Response Analysisof Structures, Systems and Components; International audience; CEA is currently working on a new prototype of Sodium cooled Fast Reactor, ASTRID, which must demonstrate strong safety level. Spacer pads, which are small bosses stamped through faces of the hexagonal duct, play a central role in this framework. Indeed, this component maintains proper spacing of fuel subassemblies. This spacing impact significantly the core reactivity, which could be decreasing due to a hypothetical core-flowering phenomenon then, be increasing during the compaction phase. Therefore, the stiffness of spacer pads needs to be well characterized and improved in order to prevent from unsafe reactivity insertion due to radial core compaction. For this purpose, a finite element model is used in order to reproduce the stamping process and obtain a representative geometry of the bosses. Moreover, the mechanical response of this model is successfully compared to experimental results of duct crushing at nominal temperature condition. Finally, this model is used in order to achieve an optimization of the design of the spacer pads for ASTRID subassemblies. This study shows that stiffness of this component can significantly be improved only by modifying its geometry.

We have studied the relationship between the Radio Continuum (RC) and CO emission for a set of galaxies selected from the BIMA SONG sample. By combining new 20 cm VLA B-array observations with BIMA maps it has been possible to investigate the behaviour of the RC/CO correlation at matched resolution down to an angular size of about 6''. Preliminary results indicate that the RC/CO correlation is as tight as (if not better than) the famous RC/FIR one. This may have important consequences on the models proposed to explain them.