In order to develop and validate the high performance tungsten monoblock technology, the full-tungsten divertor qualification program was defined. As the first step, small-scale mock-ups were ...manufactured and successfully tested under the required high heat flux loads. The test results demonstrated that the technology is available in Japan and Europe. Post-tests observation of the loaded W monoblocks showed generation of self-castellation – a crack along coolant tube axis. The cause of the self-castellation was discussed and a tungsten material characterization program is being developed with the objective to understand mechanical properties that influence the occurrence of the self-castellation.
•The optimized ITER divertor design is presented.•Shaping of vertical target design was validated by 3D field line tracing calculation and thermos-mechanical analysis.•At the monoblock level, 0.5 mm ...deep toroidal bevel was implemented and a reduction of the thickness down to 6 mm was demonstrated to be acceptable.
The shaping of the ITER divertor vertical targets has been refined as a consequence of manufacturing and engineering considerations during the prototype manufacturing activities. In this paper, the optimized ITER divertor design is presented together with design validation by 3D field line tracing calculation and thermo-mechanical analysis by finite element calculations. Furthermore, the reduction of W monoblock armour thickness to 6 mm is also discussed.
•Detailed design development plan for the ITER tungsten divertor.•Latest status of the ITER tungsten divertor design.•Brief overview of qualification program for the ITER tungsten divertor and status ...of R&D activity.
In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R&D activity is summarized in this paper.
Measurements of ion energies in the boundary of tokamak plasmas in L-mode discharges and during ELMs are reviewed. A profile of the ion-to-electron temperature ratio Ti/Te from the edge of the ...confined plasma into the scrape-off layer (SOL) is produced by compiling the available Ti measurements. The picture that emerges is that in the SOL, as well as in the edge, Ti is systematically higher than Te (ratios up to 10 just outside the last closed flux surface) for most plasma parameter regimes. Far SOL ELM ion energies measured in JET, and more recently in MAST and AUG, agree with the models of the ELM transients, providing strong evidence that ELM ions can reach the first wall with significant fraction of the pedestal energies.
•New diagnostics is implemented at the IDTF (ITER Divertor Test Facility) and TSEFEY-M high heat flux test facilities.•The flatness of the electron beam profile can be controlled by an x-ray camera ...during high heat flux tests.•The x-ray diagnostics is successfully applied at both facilities for high heat flux test of the ITER first wall and divertor.
Ensuring an accurate visualization of incident electron beam in high heat flux test facilities is a key to improve the monitoring of high-heat flux (HHF) tests carried out in view to verify the thermo-mechanical performance of components. Recent progress in X-ray imaging sensors R&D made them well-adapted for such monitoring, in particular in the case of tested components with complex shapes. This paper describes the innovative X-ray imaging diagnostic implemented in both the TSEFEY-M and IDTF facilities (Efremov Institute, Saint-Petersburg, Russia) to improve the accuracy of ITER in-vessel components HHF tests monitoring.
•Focus is on protection of leading edges between toroidally adjacent monoblocks.•Summarizes conclusions of coordinated, multi-device ITPA Divertor and SOL task.•Leading loading found to be well ...described by optical approximation.•Shaping required to prevent deep edge melting during steady state and ELMs.•Shaping implies reduced parallel heat flux to avoid material crystallization.
The key remaining physics design issue for the ITER tungsten (W) divertor is the question of monoblock (MB) front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM) or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA) and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets.
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