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The base plasma heat load on the ITER first wall results from the projection of the plasma scrape off layer power profile onto the shaped surface of the first wall panels. The actual heat load on the first wall is larger than the base heat load, due to enhancement factors such as design approximations, manufacturing and assembly errors, and deviations to the ideally toroidal magnetic field configurations. The heat load increase is accounted for through penalty coefficients, ranging individually from a few percent to 50%. The probabilistic combination of individual penalty coefficients to give a global penalty coefficient is presented in the paper. The global penalty coefficient ranges from 1.37 to 2.44, depending of the first wall row and plasma phase. [ABSTRACT FROM AUTHOR]

This paper summarizes the latest progress in the ITER blanket system design as it proceeds through its final design phase with the Final Design Review planned for Spring 2013. The blanket design is constrained by demanding and sometime conflicting design and interface requirements from the plasma and systems such as the vacuum vessel, in-vessel coils and blanket manifolds. This represents a major design challenge, which is highlighted in this paper with examples of design solutions to accommodate some of the key interface and integration requirements. [ABSTRACT FROM AUTHOR]

Abstract: A concept for a shaped first wall for ITER is presented. While keeping most features of the 2004 FDR wall (modules segmentation, plasma facing components technologies, plasma facing material), this concept provides protection of the lateral faces of the first wall panels against the intense parallel heat flux coming from the plasma. Excessive beryllium temperatures at the panel edges are avoided during regular operation. The intense heat flux at the top of the vessel is accounted for and protection is provided against the shine thru heat flux. Start-up and ramp down using the wall as a limiter is possible for up to 7.5MW, both inboard and outboard. This is rendered possible by the use of 5MW/m2 technology panels for 40% of the panels. [ABSTRACT FROM AUTHOR]

Abstract: A fundamental difference between ITER and present devices is the need to shield against 14MeV neutrons. This has major consequences for plasma start-up/rampdown (su/rd) and also for protecting the first wall from plasma contact. This has led to design decisions: (a) not to place in front of the n-absorbing blanket a separate wall-limiter structure, (b) to modularize the blanket into ∼400 remote handling compatible blanket modules (BM), and (c) to shape the front face of the BMs for plasma contact. Combined protection-su/rd options are considered here for the inner and outer wall with regard to optimal shaping. Unfortunately, the modularity of the BM system (inter-BM gaps and misalignments) requires shaping of the BM faces that increases peak power loads by ∼10× relative to the ideal (continuous, circular) wall-limiter. Fortunately, the level may still be acceptable, ∼2MW/m2, even for su/rd power of 7 MW. [Copyright &y& Elsevier]

Abstract: Long pulse operation imposes severe constraints on plasma facing components. Tore Supra pioneers this operation mode, with a large area high heat flux limiter technologically representative of next step experiments divertor targets. The failure mode of the individual elements are described, along with the strategies employed to reduce the occurrence of accidents. Two are developed: the knowledge of the heat flux in the scrape off layer and particularly the ability to predict the power density on the component’s surface, and the feed back control of edge diagnostics. Emphasis is set on infrared thermography which delivers 2D+time data, particularly useful for the prevention of accidents. This diagnostic is sensitive to the growth of carbonaceous deposits, which are highly non-uniform in Tore Supra as is presented in the paper. [Copyright &y& Elsevier]

Abstract: Three lagging tiles – over 12054 – are present since 2006 on Tore Supra main limiter, an actively cooled high heat flux plasma-facing component. The deterioration is attributed to progressing cracking of the bond between the tiles and the copper based heat sink. It is observed by an infrared camera: the thermal time constant of the tiles during cool down increased by a factor of three during the experimental campaigns of 2006 where a high level of additional power was used repetitively during long pulses. An element with a defective tile is removed for inspection during the summer shut down of 2007. The bond is cracked on three quarters of the length. Although the defects are important, the defective tiles do not limit the operation. [Copyright &y& Elsevier]

Abstract: The ITER First Wall is being redesigned to address a number of issues identified during the 2007 design review. One of the main improvements concerns the handling of parallel plasma heat loads. The design must be optimised for maximum leading edge protection with acceptable power flux distribution, which is achieved by shaping the First Wall panels. The conceptual design presented in the paper can accommodate both inboard and outboard limiter plasmas for a total power in the discharge of 7.5MW at 7.5MA and allows the abandonment of the original dedicated port limiters. [Copyright &y& Elsevier]

Actively cooled plasma facing components (PFCs) for Tore Supra (TS) have been designed basically for heat exhaust of ‘normal’ (convected and radiated) plasma power. However, in some cases, fast particles have been observed, which locally increased the power flux density, leading to damage of these PFCs and other inner vessel components. Three different examples for irreversible component damage, such as component melting and water leaks, are described involving runaway and supra-thermal particle strikes. In view of the capability for TS to handle larger input powers and to control the particles over long pulse durations, inner vessel components have been completely redesigned. The improved design concepts retained for the CIEL upgrade and preliminary results in the new configuration are presented. [Copyright &y& Elsevier]

Abstract: The implementation of actively cooled high heat flux plasma facing components (PFCs) is one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10MW/m2 of nominal convected heat flux. Technical information is drawn from the whole development up to the industrialisation and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non-destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final ‘burial’ of about half of the injected gas during long discharges. [Copyright &y& Elsevier]

Abstract: In Tore Supra long pulses, particle balance gives evidence that a constant fraction of the injected gas (typically 50%) is retained in the wall for the duration of the shot, showing no sign of wall saturation after more than 6min of discharge. During the discharge, the retention rate first decreases (phase 1), then remains constant throughout the pulse (phase 2). Phase 1 could be interpreted as implantation of particles combined with a constant codeposition rate, while phase 2 could correspond to codeposition alone, once the implanted surfaces are saturated with deuterium. This paper presents a possible contribution of charge exchange neutrals to the implantation process, based on modelling results with the Eirene neutral transport code. A complex pattern of particle implantation is evidenced, with saturation time constants ranging from less than one to several hundreds seconds, compatible with the experimental behaviour during phase 1. [Copyright &y& Elsevier]

Abstract: Budget restrictions have forced the ITER Organization to reconsider the baseline divertor strategy, in which operations would begin with carbon (C) in the high heat flux regions, changing out to a full-tungsten (W) variant before the first nuclear campaigns. Substantial cost reductions can be achieved if one of these two divertors is eliminated. The new strategy implies not only that ITER would start-up on a full-W divertor, but that this component should survive until well into the nuclear phase. This paper considers the risks engendered by such an approach with regard to known W plasma-material interaction issues and briefly presents the current status of a possible full-W divertor design. [Copyright &y& Elsevier]