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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]

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: In magnetic fusion experiments, a simple technique to evaluate the heat flux on first wall components is a key to controlled plasma surface interaction. The heat flux can be characterized by the peaking factor which is the ratio of the peak heat flux to the average heat flux. The peaking factor can be calculated exactly using simple derivations and standard software tools. This analysis is applied to an Iter class experiment for plasma-wall contact during start up phases at 15MW, in idealised, realistic and misaligned situations. Even though the peaking factors are usually above 10, the peak heat load on the wall remains moderate at a few MW/m2. [Copyright &y& Elsevier]

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: A precise geometric method is used to calculate the power deposition on the future JET ITER-Like Wall beryllium tiles with particular emphasis on the internal edge loads. If over-heated surfaces are identified, these can be modified before the machining or failing that actively monitored during operations. This paper presents the methodology applied to the assessment of the main chamber beryllium limiters. The detailed analysis of one limiter is described. The conclusion of this study is that operation will not be limited by edges exposed to plasma convective loads. [Copyright &y& Elsevier]

Abstract: Future fusion experiments will rely on tungsten armour tile for their plasma facing components. In order to sustain steady state operation, the components need to be cooled through an attachment to a heat sink. All current reference concepts rely on contact bonds, unfavourable for long-term application (high temperature service, cycle fatigue, thermal shocks). Three routes toward the development of thick tungsten bonds are presented here, namely functionally graded tungsten copper assembly, electron beam welding of tungsten, and coating processes. All present favourable prospects, and tend to indicate that a thick bond is possible with tungsten. Dedicated programs as well as industrial implication are however required if such concepts are to be used actually for the fabrication of large components series. [Copyright &y& Elsevier]

Abstract: The material eroded from the surface of plasma facing components is redeposited partly close to high heat flux areas. At these locations, the deposit is heated by the plasma and the deposition pattern evolves depending on the operation parameters. The mapping of the deposit is still a matter of intense scientific activity, especially during the course of experimental campaigns. A method based on the comparison of surface temperature maps, obtained in situ by infrared cameras and by theoretical modelling is proposed. The difference between the two is attributed to the thermal resistance added by deposited material, and expressed as a deposit thickness. The method benefits of elaborated imaging techniques such as possibility theory and fuzzy logics. The results are consistent with deposit maps obtained by visual inspection during shutdowns. [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]

• Active wall monitoring systems are an increasingly important asset for safe operation of metallic wall steady state magnetic fusion experiments. • Wall Infrared diagnostic connection to plasma control system enables wall monitoring. • WEST operates the C4 experimental campaign in 2019 with an ITER relevant wall monitoring system (soft control plus hard interlock). • The control activates on 63 occurrences, principally on an upper divertor pipe being heated by particles losses. • It facilitated WEST path to high power operation during C4 campaign, by effectively managing the technical risks to critical wall components. A real time Wall Monitoring System (WMS) is used on the WEST tokamak during the C4 experimental campaign. The WMS uses the wall surface temperatures from 6 fields of view of the Infrared viewing system. It extracts the raw digital data from selected areas, converts it to temperatures using the calibration and write it on the shared memory network being used by the Plasma Control System (PCS). The PCS feeds back to actuators, namely the injected power from 5 antennae's of the lower hybrid and ion cyclotron resonance radiofrequency (RF) heating systems. WMS activates feed back control 63 times during C4, which is 14 % of the plasma discharges. It activates mainly as the result of a direct RF loss to the upper divertor pipes. The feedback control maintains the wall temperature within the operation envelope during 97 % of the occurrences, while enabling plasma discharge continuation. The false positive rate establishes at 0.2 %. WMS significantly facilitated the operation path to high power operation during C4, by managing the technical risks to critical wall components. [ABSTRACT FROM AUTHOR]

Highlights: [•] The ITER blanket is in the final stage of design completion. [•] Issues raised about the blanket heat loads and remote handling strategy are addressed, while integrating the in-vessel coils. [•] Key design justifications are presented, followed by a summary of the current status of the manufacturing plan and R&D activities. [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]

As a consequence of its position and functions, the ITER blanket system will be subjected to significant heat loads under nominal reference conditions. Therefore, the design of its cooling system is particularly demanding. Coolant water is distributed individually to the 440 blanket modules (BMs) through manifold piping, which makes it a highly parallelized system. The mass flow rate distribution is finely tuned to meet all operation constraints: adequate margin to burn out in the plasma facing components, even distribution of water flow among the so-called plasma-facing “fingers” of the Blanket First Wall panels, high enough water flow rate to avoid excessive water temperature in the outlet pipes, maximum allowable water velocity lower than 7 m/s in manifold pipes. Furthermore the overall pressure drop and flow rate in each BM shall be within the fixed specified design limit to avoid an unduly unbalance of cooling among the 440 modules. Analyses have to be carried out following a computational fluid-dynamic (CFD) approach based on the finite volume method and adopting a CFD commercial code to assess the thermal-hydraulic behaviour of each single circuit of the ITER blanket cooling system. This paper describes the code benchmarking needed to determine the best method to get reliable and timely results. Since experimental tests are available in ITER Organization on full scale prototypes of Shield Blocks #08 and #14, CFD analyses have been performed to investigate their fluid-dynamic behaviour under steady state conditions and compare the numerical and experimental results. Results obtained are presented and critically discussed. [ABSTRACT FROM AUTHOR]

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]

Abstract: During the last Tore Supra campaigns, an anomalous deformation in the near infrared spectrum of radiation has been observed on neutraliser underneath the Toroidal Pumped Limiter (TPL) on which we have observed the growth of carbon layer. The consequence is the difficulty to assess the surface temperature of the components and the power loaded. Laboratory experiment has been performed, using an Electron Cyclotron Resonance (ECR) ions source, to reproduce, characterize and explain this phenomenon. The luminance emitted by Carbon Fibre Composite (CFC) and pyrolytic graphite, have been observed under 95keV of H+ bombardments. The amplitude of the deformation was found to depend on the type of material used and the power density of the incident power loaded. This paper presents the possible hot spots explanation. The experimental luminance deformation is reproduced and these results are validated using a thermal model of dust in radiative equilibrium. [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]

Nitrogen (N2) will be used in ITER to enhance the radiative fraction to ∼90%, thereby cooling the edge plasma and preventing damage to the plasma-facing components. However, the reactivity of N2 with hydrogen isotopes can lead to the formation of tritiated ammonia (NT3). This should be considered in terms of the in-vessel tritium inventory, the regeneration of the cryo pumps, and the processes in the ITER de-tritiation plant. In the 'W' Environment in Steady-state Tokamak (WEST), a series of long L-mode discharges (∼50 s), with a constant N2 seeding from the outer strike point region has been performed on the upper actively cooled divertor. In the absence of active pumping, the N2 balance shows steady-state retention during plasma discharge, and is partially (∼35%) released in between discharges. Although a significant amount of N2(18.65 Pa m3) has been injected, the wall still exhibited N2 pumping capabilities. Under these conditions, as long as this N2 reservoir is not saturated, there is not enough N available for the detectable threshold of ND3 formation to be reached. In these WEST experiments, no ammonia is detected during the pulse or after the pulse in the outgassing phase. These results are consistent with and complementary to the N2 seeded experiments performed in the Joint European Torus (JET) with its ITER-like wall and in the Axially Symmetric Divertor Experiment (ASDEX) upgrade. [ABSTRACT FROM AUTHOR]

This paper presents the method toward quantifying the global uncertainty on the blackbody temperature measurement from the infrared (IR) diagnostics of WEST. To address this goal, the main functional elements of the IR diagnostics are identified. Then the temperature calibration and calculation principles are presented and analysed to extract the main potential uncertainty contributors, such as the optical transmission coefficients and their stray lights, or the accuracy of temperature references used for calibration. These contributors are individually estimated, or experimentally measured when a supposed effect on the overall uncertainty is identified, like environmental conditions or parasitic radiations. In particular, effects of environmental temperature on the transmission lines and camera is thoroughly studied. All contributions are then aggregated in the uncertainty propagation calculation and results in an overall temperature uncertainty versus the temperature estimation. The uncertainty is in the range of 5–10% for blackbody temperatures above 200 °C, and progressively worsens when temperature decreases. [ABSTRACT FROM AUTHOR]

A tangential infrared thermography system is installed in the WEST tokamak to observe the thermal scene of the first wall and the divertor components through a wide-angle tangential view. The goal is to provide and monitor the heat load deposition on the plasma facing components. About one-sixth of the chamber is observed, including sectors of the baffle, the lower and upper divertors, one inner bumper, vertical displacement event (VDE) protections, one ICRH antenna, and other parts of the wall. It is used for real time machine protection by monitoring temperature thresholds in delimited region of interest, and for analysis of normal or specific heat load events during operation such as VDE, ELM, disruption or runaways. This wide-angle view uses one aspherical and one on-axis plane mirrors, a sapphire window and three lenses for the objective of the camera. The optical line is optimized for two wavelengths 1.7 and 4 µm. The field of view is 60° on a 512×640 pixels Focal Plane Array. The endoscope is fully actively cooled. The thermal scene is complex to interpret given the fully metallic and radiative environment and the uncertainties on the emissivity which is angular-dependent and changes with the surface properties. This involves significant inaccuracy on the recovery of the real temperature. In this context, an interpretation by modeling approach is better suited, based on ray-tracing simulations taking into account the optical properties of materials. For instance, this allowed discriminating reflections patterns from real thermal events in the wide-angle view. A description of the wide-angle infrared diagnostic and its performances is presented as well as experimental measurements obtained in the WEST Tokamak. [ABSTRACT FROM AUTHOR]

• Successful high heat flux test results of ITER First Wall mock-ups after neutron irradiation • High heat flux testing of actively cooled mock-ups manufactured with the relevant technologies in relevant geometry to the ITER First Wall, by the suppliers involved in the series production for ITER • High heat flux testing of neutron-irradiated beryllium flat tiles with hypervapotron heat sink • Details of high heat flux test by electron beam facility, JUDITH facility and details of neutron irradiation in the fission reactor, HFR Eight mock-ups imitating the ITER Enhanced Heat Flux First Wall panel were subjected to thermal fatigue test after neutron irradiation at two dose levels, ≈0.1 dpa and ≈0.5 dpa. All the mock-ups successfully withstood repeated thermal fatigue loads up to 4.7 MW/m2. The test results confirmed the performance of mock-ups in the tested condition, specifically the design and manufacturing technologies of the suppliers of two domestic agencies in charge of the series production. [ABSTRACT FROM AUTHOR]

Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed. [ABSTRACT FROM AUTHOR]

The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised. [ABSTRACT FROM AUTHOR]

This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm × 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale “fingers” and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels). The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Jülich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m 2 is described in detail. [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: The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. The blanket manifold system has been redesigned to improve leak detection and localisation. The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. This paper describes the status of the redesign of the in-vessel components and the associated integration issues. [Copyright &y& Elsevier]

Highlights: [•] 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. [Copyright &y& Elsevier]

Abstract: The ITER plasma-facing components directly face the thermonuclear plasma and include the divertor, the blanket and the test blanket modules with their corresponding frames. The divertor is located at the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimising the helium and impurity content in the plasma. The blanket system provides a physical boundary for the plasma transients and contributes to the thermal and nuclear shielding of the vacuum vessel and external machine components. It consists of modular shielding elements known as blanket modules which are attached to the vacuum vessel. Each blanket module consists of two major components: a plasma-facing first wall panel and a shield block. The test blanket modules are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to test blanket modules, each of them providing for the allocation of two breeding modules inserted in a steel frame and in front of a shield block. [ABSTRACT FROM AUTHOR]

Abstract: On the limiter of Tore Supra, the heat loads map computed from deconvolution of IR surface temperatures shows good agreement with calorimetry measurements. This experimental heat pattern allows deducing the heat fluxes in the scrape-off layer using a 3D magnetic calculation and assuming only parallel heat transport along field lines. This calculation leads to an underestimation of the power circulating in the edge plasma according to the power balance, similarly to RFA measurements. The comparison between experimental heat loads on the limiter and modelling also shows a spreading of heat fluxes near the LCFS that cannot be explained only by parallel transport in the SOL. [Copyright &y& Elsevier]

Abstract: The precursors and following sequential events leading to radiative disruptions in Tore Supra have been analysed using infrared imaging, together with visible and ultraviolet spectroscopy of impurity species. A common feature observed prior to the disruptions is the appearance of a small (∼cm2) hot spot on the main plasma facing component, the Toroidal Pumped Limiter (TPL), clearly localised in a zone of thick carbon re-deposition (>100μm). A MARFE (Multifaceted Asymmetric Radiation From the Edge) is often triggered, followed by disruption. Such hot spots have been observed in ∼24% of the analysed disruptions, which is consistent with the fact that only 4/18 (22%) of the total area of the TPL is monitored with infrared cameras. These results suggest that over-heating of thick carbon re-deposition layers may play a role in the operational limits (MARFE, disruption) encountered. [Copyright &y& Elsevier]

Abstract: A spatially resolved calculation of the charge exchange particle and energy fluxes on the Toroidal Pumped Limiter (TPL) of Tore Supra is presented, as a first step towards a better understanding and modelling of carbon erosion, migration, as well as deuterium codeposition and bulk diffusion of deuterium in Tore Supra. The results are obtained with the EIRENE code run in a 3D geometry. Physical and chemical erosion maps on the TPL are calculated, and the contribution of neutrals to erosion, especially in the self-shadowed area, is calculated. [Copyright &y& Elsevier]

Abstract: The ITER Ion Cyclotron Resonant Heating (ICRH) antenna provides plasma heating at a power of 20MW. Operation in the ITER environment imposes significant thermal power handling capability, structural integrity, shielding and operations requirements. The design will require a step change over any predecessor in terms of power, scale and complexity. This paper reports the main mechanical design features that address the challenges and often conflicting requirements during the conceptual design phase. [Copyright &y& Elsevier]

Abstract: The ITER reference materials have been tested separately in tokamaks, plasma simulators, ion beams and high heat flux test beds. In order to perform a fully integrated material test JET has launched the ITER-like Wall Project with the aim of installing a full metal wall during the next major shutdown. As a result of R&D projects in 2005–2006, bulk tungsten tiles are foreseen at the outer horizontal target and tungsten coating at the other divertor tiles. In some regions of the main chamber, beryllium coated Inconel tiles and bulk beryllium tiles are utilised which include marker tiles as erosion diagnostics. This paper gives an overview of the R&D carried out in the frame of the ITER-like Wall Project on the development of an inertially cooled bulk tungsten tile design and the characterization of tungsten and beryllium coating technologies. [Copyright &y& Elsevier]

Abstract: Using the Tore Supra infrared thermography diagnostics, a new real time feedback control has been successfully implemented to maximize additional RF power while preventing plasma facing components (PFCs) from overheating and damage. As a first step, a thermography feedback control has been used to detect and extinguish electric arcs on lower hybrid current drive (LHCD) launchers. Secondly, heating sources on PFCs have been identified highlighting the role of the power from each ion cyclotron resonance heating (ICRH) antenna and LHCD launcher and the interactions between them. A new feedback control algorithm was developed to control the additional power. The real time feedback control of PFC temperatures which makes part of an integrated feedback controller, is a reliable tool routinely used as a basic protection system. Furthermore, it has proven its capability to operate in parallel with other control schemes such as the current profile control. [Copyright &y& Elsevier]

Abstract: In this study, processing of W–Cu materials from attritor-milled W–CuO mixtures is described. Reduction steps of powder compacts are investigated by TGA, dilatometry and XRD analyses. Morphological characteristics are observed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Sintering of reduced powder with different compositions is analyzed by dilatometry and assembly of graded materials is briefly discussed. [Copyright &y& Elsevier]

Abstract: In the frame of JET’s ITER-like wall project for most of the divertor surface tungsten coatings are intended to be employed on bidirectionally carbon fibre reinforced carbon substrates. Since this is thermomechanically rather mis-matched, a variety of deposition conditions were considered. Mostly in cooperation with industry, five Euratom associations provided 14 different types of samples with respect to production method and coating thickness. In a step-wise selection procedure, these were subjected to a thermal screening test and a thermal cycling test in the ion beam facility GLADIS as well as to an ELM-like thermal shock test in the electron beam facility JUDITH. A general failure mode is crack formation upon cool-down. Coatings with several microns of thickness show a distinct delamination feature in addition. Further analysis included metallographic investigation, X-ray diffraction for film stress assessment, adhesion testing as well as measurements on the contents of light impurities. [Copyright &y& Elsevier]

Abstract: The uniform layer with thickness at least 50–100μm was found on the CFC tiles from the inboard midplane after more than four years of tokamak operation. The upper part of the uniform layer was amorphous, but at the depth of ∼5μm a structure consisting of micro-size regions with aromatic chains located parallel to the surface was found. Gradual transition from uniform layer to underlying CFC structure was observed. The reciprocating material probe was used for installation of CFC samples in the Tore Supra deuterium plasma. The thermal desorptional spectra of these samples are compared with the spectra of the samples irradiated in the laboratory stand and with the spectra of hydrogenated carbon film. The peculiarities of hydrogen isotope trapping under plasma irradiation and at the atmosphere are presented and discussed. [Copyright &y& Elsevier]

Abstract: This paper presents the results of a modelling study conducted to estimate the power crossing the separatrix (P SOL) in the ITER device during a standard start-up sequence. This is used to calculate the power intercepted by the start-up limiters and the resulting power load distribution. The models and methodologies applied to calculate P SOL and the power loads on the limiters are described in detail elsewhere ([e.g., M. Kobayashi et al., Nucl. Fusion. 47 (2) (2007) 61]) and only a brief mention of some of the main results is included here. These assessments show that for the range of conditions analysed, the maximum P SOL intercepted by the two ITER limiter start-up modules during the current ramp-phase is ∼6MW. The peak power load to each limiter is calculated to be ∼5MW/m2, but these values depends on assumptions on physical quantities (e.g., transport coefficients, i.e., D ⊥ and χ ⊥), which are uncertain and still await confirmation by experiments. Recommendations are made for modelling and experiments to extend the study presented here. [Copyright &y& Elsevier]

Abstract: We have re-examined the basic dependences of carbon generation due to physical (D+ and Cn+) sputtering and from thermally dependent sources (chemical erosion) by comparison with a spectroscopic database for carbon emission from localized regions of CIEL. To be able to compare with observations in this complex environment, a model for carbon generation and transport has been created to include contributions from the important, but non-ideal, processes of carbon generation from material in intra-tile gaps and from poorly adherent re-deposited layers. Consistency simulations have been carried out to assess the degree to which the spot observations represent local emission, due to possibly long mean free paths of high-energy emitted particles, or from impurities transported into the spectroscopic field of view from other areas. Model results are compared with the experimental trends in the ratio of CII and Dα emission with power and edge parameters. In the course of the analysis a potentially important vector has been found for transport of re-deposited material to more remote locations and its significance discussed. [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: The new CIEL (Composantes Internes Et Limiteur) configuration of the Tore Supra tokamak has as its main plasma-facing component (PFC) a Toroidal Pumped Limiter (TPL) [P. Garin et al., in: Proceedings of the 20th Symposium on Fusion Technology, Marseille, vol. 2, 1998, p. 1709], which must sustain the bulk of the energy leaving the plasma. Analysis of the heat deposition pattern on the TPL indicates that perpendicular heat transport may play at least as significant a role as parallel heat transport [F. Saint-Laurent et al., Nucl. Fusion 40 (2000) 1047, R. Mitteau et al., these Proceedings]. We present a new approach for modelling the heat deposited onto the TPL, which follows test ‘heat packet’ trajectories backwards from the TPL towards the hot plasma column. Results are compared with experimental data and trends due to plasma parameters dependencies are described. Because of ripple effects, the limiter is covered by wetted areas with long connection lengths (tens of meters), and shadowed areas with very short connection lengths (centimeters). Sharp transitions between the two are clearly seen in experiment and also reproduced in the model. [Copyright &y& Elsevier]

A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components in Tore Supra operation, which is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable of sustaining up to 10 MW/m2 of nominal convected heat flux. This success is the result of a long lead development and industrialization program (about 10 years) marked out with a number of technical and managerial challenges that were taken up and has allowed us to build up a unique experience feedback database, which is displayed in the paper. [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]

The ITER equatorial port visible/infrared wide angle viewing system concept is developed from the measurement requirements. The proposed solution situates 4 viewing systems in the equatorial ports 3, 9, 12, and 17 with 4 views each (looking at the upper target, the inner divertor, and tangentially left and right). This gives sufficient coverage. The spatial resolution of the divertor system is 2 times higher than the other views. For compensation of vacuum-vessel movements, an optical hinge concept is proposed. Compactness and low neutron streaming is achieved by orienting port plug doglegs horizontally. Calibration methods, risks, and R&D topics are outlined. [ABSTRACT FROM AUTHOR]

Abstract: In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today’s devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues. [Copyright &y& Elsevier]