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Abstract The role of the ITER Tokamak thermal shield system is to minimize heat loads transferred from the hot tokamak components to the cooled superconducting components. The system consists of two single wall structures fabricated from 304L-(N) stainless steel. Both structures are actively cooled with pipes permanently welded to the structure and filled with pressurized helium gas at 80K. One structure is the Vacuum Vessel Thermal Shield (VVTS), which protects magnet components from Vacuum Vessel (VV) radiation. The second structure is the Cryostat Thermal Shield (CTS), which protects magnet components from in-Cryostat components' radiations such as the Cryostat (at 293K), the VV Ports (at 393K), and the water cooling system (at 350K). In 2012, thermal analyses were performed in order to verify the thermal integrity of the ITER CTS 2010 baseline design. The heat loads to the CTS were estimated according to load specification as a homogeneous distribution based on cryostat temperature at 293K. Following this study, in order to determine a more realistic heat load distribution on the CTS panels, we performed a detailed 3D analysis of the Cryostat-CTS interspace. This paper describes the methodology used, based on view factors combined with Kirchhoff's law, to define the heat flux deposition on the CTS from its surrounding components via radiation. The previous thermo-hydraulic analysis models have been adapted using this method to define the temperature distribution and total power transmitted to the magnets. Despite an increase of 30% of the CTS total power deposition on the superconducting components in comparison to the previous analysis, we found the total heat power deposition onto the superconducting components at 4K to be within the limit of the design requirement (P≤1900W). [ABSTRACT FROM AUTHOR]

• 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]

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]