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• The results of cyclic tension/compression tests conducted in imposed stress show that the material can exhibit elastic shakedown at low stress amplitude and plastic shakedown up to ratcheting at higher stress amplitude. • By a introducing a damage variable to represent the evolution of micro-cavities or cracks and by coupling this variable with the kinematic and isotropic hardenings, the model allows of representing plastic end elastic shakedown. • The model generates ratcheting response when the damage variable evolution progresses significantly. • The modelling accounts for strong thermo-mechanical couplings to deal with transient thermal situations or to account of the impact of strong thermal gradients that can be encountered in nuclear steel structures. The ITER machine has been classified as a Basic Nuclear Installation French nuclear regulator (INB n°174), which implies that it will be the first fusion reactor to go through complete French nuclear licencing. The combination of mechanic and electromagnetic phenomena with the heat loads caused by neutron streaming requires a multi-physics approach to the damage assessment, which has not yet been implemented in the common nuclear codes and standards. The general damage prevention methodology consists in guaranteeing the structural integrity of a component. The development of design rules has mainly two origins: prevention of damage from monotonic mechanical loads and prevention of damage from repeated application of loads. In most cases, structural integrity is justified within a linear elastic behaviour but when this route is not enough to respect the design criteria, several non-linear approaches to the material's mechanical behaviour can be considered, requiring more elaborated demonstration of the design compliance. Nevertheless, the models proposed in the nuclear model database are sometimes not sufficient to properly describe the experimentally observed cyclic plasticity behaviour and, in particular, the ratcheting and shakedown phenomena. According to ITER community experts in materials and analyses, a thermo-mechanical behaviour model fitting the ITER Tokamak materials data will guarantee the best prediction of the damage considering a nuclear and a multi-physic loading condition. This paper describes the assessment of the non-linear behaviour of Vacuum Vessel (VV) material with a strong thermo-mechanical coupling and a damage parameter to prevent crack initiation. More precisely, Chaboche's models available in the literature (elasto-(visco)-plasticity models, with various types of hardening) have been enriched in order to explicitly take into account the influence of the temperature on the mechanical behaviour and, reciprocally, the influence of the mechanical behaviour on the temperature. Mechanical cycling tests have been performed on the VV constitutive material, emphasizing on the progressive deformation state up to failure mode, i.e., ratcheting. The proposed models have been tested on a homogeneous problem and the results compared with uniaxial test results; this type of simulation is commonly called "0D"analysis. The last part of this document describes the finite element implementation of the constitutive material model and its application to the ITER VV welded support. [ABSTRACT FROM AUTHOR]

The present work exposes the 3D steady state thermal hydraulic (TH) analysis of the Irregular Sector number 3 (IrS#3) of the ITER Vacuum Vessel (VV), including the Irregular Field Joints (IrFJ) by means of a commercial CFD (Computational Fluid Dynamics) code. The geometry of the IrS#3 has been simplified and healed in order to be suitable for CFD analysis. The simplified geometry has been meshed and converted to a polyhedral cell based mesh so as to enhance accuracy and calculation stability. Nuclear heat deposition ( H ) from the latest C Lite model [1] has been implemented through several User Defined Functions (UDF). Water coolant and stainless steel shell are solved together as a steady state conjugate heat transfer problem in order to assess the impact of the nuclear heat deposition on the IrS#3 cooling scheme. Hence, the IrS#3 is simulated as a whole, without splitting the domain. Results show the total IrS#3 pressure drop as well as the flow and temperature distribution all over the IrS#3. Moreover, the heat transfer coefficient ( h ) has been calculated at the fluid-solid interface. Velocity magnitude in the water coolant has an average value of 0.10 m/s and mass flow rate distribution is 11% and 89% respectively for Inboard and Outboard. The pressure drop, mainly concentrated at inlet and outlet ducts, is of 77.4 kPa. The average h is of 5 690 W/(m 2 K), way above the design limit of 500 W/(m 2 K). The fluid temperature stays at an average value of 101.7 °C. The results obtained have a significant importance regarding design and safety and give a valuable insight on the current cooling scheme and system behavior for the IrS#3 of the ITER VV. This follows a previous work presenting the methodology and the results of a 3D TH analysis of the Irregular Sector number 2 (IrS#2) of the ITER Vacuum Vessel (VV) by means of CFD [2]. [ABSTRACT FROM AUTHOR]

Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE). [ABSTRACT FROM AUTHOR]