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One of the challenges that International Thermonuclear Experimental Reactor (ITER) will face during plasma operation is to determine the absolute divertor surface temperature of the actively water-cooled (70 °C, 3 MPa) tungsten plasma-facing units (PFUs) to ensure their integrity. The expected steady-state heat flux up to 10 MW/m2 is close to the operational limit and so one of the goals of the thermographic system is to make reliable measurements (wavelength band: 3.5– $4.5~\mu \text{m}$ ) with relatively low error bars to avoid tungsten melting and material damages. The tungsten emissivity is low and dependent on wavelength, temperature, and surface state (roughness, cracks, oxidation, and erosion/deposition processes) which can evolve during the time along with plasma operation. An error on the absolute emissivity can lead to large absolute temperature errors, and consequently to either a reduction of the operational window to fulfill safety limits, or an increased risk regarding the integrity of the components. For the past years, a number of emissivity measurements have been performed with ITER-like PFU mock-ups made of different grades of tungsten, with different techniques, from different material makers, and from different manufacturers. The overall emissivity discrepancy is large, indicating that tungsten emissivity is a major issue for temperature measurement and wall protection. The accuracy of the temperature measurement is investigated for both monocolor and bicolor IR thermography techniques. We present laboratory tests performed with an IR camera equipped with a rotating filter wheel for bicolor temperature measurement. It shows that the bicolor technique can reach high precision $\Delta \text {T}/\text {T} at high temperature assuming the emissivity ratio is constant regarding the two selected wavelengths.

In preparation of the production of the actively cooled tungsten targets for the lower divertor of WEST (W -for tungsten- Environment in Steady-state Tokamak), the qualification of suppliers is done by exposing small-scale mock-ups to cyclic high heat flux tests in the electron beam facility JUDITH 1 at Forschungszentrum Julich. Thereby, the small-scale mock-ups are based on the ITER design, consisting of 7 monoblocks mounted on a CuCrZr cooling tube and connected via a soft Cu-interlayer. The results presented herein focus on the high heat flux test results of two mock-ups produced by AT&M in China. The testing was performed by thermal cycling at 10 MW/m2 and 20 MW/m2 with a maximum of 1200 cycles at 10 MW/m2 followed by 500 cycles at 20 MW/m2 without obvious damage formation. The qualification is supported by subsequent hardness tests and microstructural analyses. The latter focus on local variation of tungsten recrystallization and macro-crack formation and deformation induced changes in the heat sink materials.

International audience; ITER (https://www.iter.org) will take advantage of tungsten (W) actively cooled Plasma Facing Units (PFU) in the lower divertor. These PFUs will receive a steady state plasma heat flux of 10 MW/m² and up to 20 MW/m² in slow transients, pushing these components to their limit. For machine protection reasons and for the study of the plasma wall interactions, temperature measurements are foreseen with Mid Wave Infra-Red (MWIR: 3-5 µm) thermography systems to cover most of ITER chamber including the divertor (tungsten material) and first wall (beryllium material) [1]. Absolute temperature measurements from 70 °C, the cooling temperature of the PFUs, up to 3500 °C (W melting temperature at 3422 °C), are considered. Those measurements require the knowledge of the emissivity of the tungsten PFU surfaces as function of the temperature and wavelength used for the IR monochrome thermography system. This paper summarizes the emissivity measurements performed on tungsten surfaces in High Heat Flux (HHF) test beds and the WEST tokamak [2] and uses the experimental emissivity values to compute the expected uncertainties in ITER (T/T) using monochrome and bicolor thermography techniques. Results show that monochrome technique is not able to fulfill the T/T < 10% requirement, while bicolor technique is able to reduce the temperature uncertainty below 10%. Laboratory bicolor thermography measurements using a filter wheel (including 6 MWIR interference filters) have been successfully performed up to 830°C with temperature uncertainty T/T < 3%. In the WEST tokamak, the two dimensions emissivity map is varying by a factor of five within 27 cm along a PFU, from 0.13 to 0.65, showing that the monochrome IR thermography will have difficulties to stay within the ITER requirements T/T < 10% from 70 °C up to 3500 °C.

International audience; 4 probes with Fiber Bragg Gratings (FBGs) will be integrated in Plasma Facing Component (PFC) of the WEST lower divertor in order to measure the bulk temperature distribution in the poloidal direction (11 locations equally spaced). A 2D nonlinear unsteady calculation combined with the Conjugate Gradient Method (CGM) and the adjoint state is used in order to estimate the space and time evolution of the surface heat flux based on temperature measurements. Synthetic measurements are generated assuming optical projection of the heat flux taking into account physical effects such as heat flux decay length ( λq ) and power spreading factor ( S ) in the private flux region and geometrical effects such as surface chamfering and round leading edges of the component. The inverse method is applied on different synthetic measurements in order to evaluate the ability of using FBG measurements to characterize the heat flux poloidal profile.

International audience; Since tungsten (W) has been selected as material for the ITER divertor target the knowledge of the emissivity in the infrared (IR) wavelength range becomes mandatory. A dedicated setup has been developed to measure the emissivity in the wavelength range 1.7–4.75 μm and the temperature domain 200–850 °C. The paper presents the emissivity measurements on W samples coming from actively cooled bulk W components (ITER-like) and W-coated graphite component of the WEST divertor. W samples with damaged surfaces generated by transient heat load as those observed in fusion machines, including micro-cracks and crack network, are also investigated. The dependence on wavelength, temperature and surface state are shown and discussed.

Based on a monoblock concept (e.g. a tube-in-tile concept), each elementary tungsten plasma facing component (called Plasma-Facing Unit PFU) of the WEST lower divertor follows as closely as possible the same monoblock geometry, materials and bonding technology that is envisaged for ITER. A fatigue simulation of W PFU was used to validate its specific integration into WEST. The complex design, the material heterogeneities and the usage outside operational load design envelope are all possible causes of fatigue failure. This paper shows how the ITER like monoblocks and its U-shaped attachments technology are integrated into the WEST divertor by performing finite element analysis. The WEST lower divertor is designed to withstand 15 MW steady-state of injected power, with peaked heat fluxes up to 20 MW/m 2 . The integration and the design choices of a WEST ITER Like Plasma Facing Unit inside the WEST vacuum chamber is valid for an “expected life time” of repeated inter ELMs thermal steady state (>10 s) cycles and for 300 off-normal vertical displacement events.

International audience; Plasma Facing Components (PFC) temperature measurement is mandatory to ensure safe high power and long pulse tokamak operation. IR thermography systems which are widely used in magnetic fusions devices become challenged with the choice of tungsten as a PFC material in the I l'ER tokamak, mainly due to emissivity uncertainties and reflection issues in a hot environment. Embedded temperature measurements are foreseen to cross-check the IR thermography measurements. Fiber Bragg grating sensors are potentially of great interest for this application because they are immune to electromagnetic interference and allow the measurement of a large number of temperature spots on a single fiber. Four optical fiber temperature sensing probes, each of them including 11 regenerated fiber Bragg gratings equally spaced by 12.5 mm (equivalent to one ITER-like tungsten monoblock), have been specifically designed and manufactured for the WEST project (W-tungsten Environment and Steady State Tokamak). The four probes are embedded in W-coated graphite components at two different distances from the surface, 3.5 mm and 7 mm, to cover a wide range of temperatures up to 900 degrees C. This paper addresses the design and integration issues and the qualification and performance assessment performed in the laboratory. It also shows the first measurements of this new diagnostic achieved in a tokamak environment during baking of the machine and during early diverted plasma exposure.

The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is based on an upgrade of the Tore Supra tokamak, into a double X-point divertor device, while taking advantage of its long discharge capability. Therefore components with tungsten (W) as plasma-facing material will be used. This paper presents the upper divertor of the WEST project (design constraints, thermal performance). This component, with a total surface of 8 m2 is designed to exhaust 4 MW of conducted power in steady state with a maximum local heat load of 8 MW m−2. The actively cooled heat sink of this upper divertor target is made of CuCrZr. It is covered with a W coating up to 30 μm thickness. The suited thermal behaviour of the component was checked using finite element modelling. Moreover, under high heat flux tests this component had the expected thermal exhaust capability and survived without visible damage up to 10.5 MW m−2 in steady-state.

In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20 MW/m2 heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design. One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30° sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30° sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected.

The fusion machines under development or construction (ITER, W7X) use several hundreds of actively cooled plasma facing components (ACPFC). They are submitted to leak tightness requirements in order to get an appropriate vacuum level in the vessel to create the plasma. During the ACPFC manufacturing and before their installation in the machine, their leak tightness performance must be measured to check that they fulfill the vacuum requirements. A relevant procedure is needed which allows to segregate potential defects. It must also be optimized in terms of test duration and costs. Tore Supra, as an actively cooled Tokamak, experienced several leaks on ACPFCs during the commissioning and during the operation of the machine. A test procedure was then defined and several test facilities were set-up. Since 1990 the tightness of all the new ACPFCs is systematically tested before their installation in Tore Supra. During the qualification test, the component is set up in a vacuum test tank, and its cooling circuits are pressurized with helium. It is submitted to 3 temperature cycles from room temperature up to the baking temperature level in Tore Supra (200 °C) and two pressurization tests are performed (6 MPa at room temperature and 4 MPa at 200 °C) at each stage. At the end of the last cycle when the ACPFC is at room temperature and pressurized with helium at 6 MPa, the measured leak rate must be lower than 5 × 10 −11 Pa m 3 s −1 , the pressure in the test tank being −5 Pa. A large experience has been gained on ACPFCs with carbon parts on stainless steel and CuCrZr structures (toroidal pumped limiter (TPL) composed of 12 sectors of 48 fingers) but also on all metallic ones (ICRH antennas and LHH launchers, vacuum vessel protection walls ∼60 m 2 ). During the tests, some leaks were found and the components had to be repaired. The analysis of the results database allows defining guidelines to optimize the presently used procedure for the new ACPFCs generation which will be set-up in WEST or ITER (tungsten (W) elements, high mass components). This paper presents the experience of the Tore Supra teams in the implementation of ACPFC with regard to the tightness issues including the tests procedures which were developed and used since 1990. The prospects for the new generation actively cooled machines are also discussed.

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m2 range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program. WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.

The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.

O aborto abarca um campo de tensões circundado por controvérsias morais. Sua legalização em diversos países não foi capaz de diminuir os conflitos que emergem da discussão sobre o tema, tornando-se necessário ampliar a análise em distintas esferas, como a política, a econômica, a saúde e a ética. Constitui-se como um dos poucos assuntos capazes de provocar profundas divisões sociais. A compreensão sobre as questões envolvidas e os desdobramentos morais decorrentes dessa prática requerem a retomada de um breve histórico, o que possibilita a análise da forma como o tema vem sendo tratado ao longo da história da humanidade. Este trabalho teve como objetivo realizar uma análise histórica e jurídica subjacentes ao aborto e as possíveis consequências para as diferentes moralidades que emergem da vida em sociedade. Foi realizada uma revisão da literatura a partir de livros e artigos científicos, o que permitiu analisar do ponto de vista bioético os distintos posicionamentos existentes no mundo e consequentemente as posições morais frente aos novos paradigmas propostos na atualidade.

Lower hybrid current drive (LHCD) and ion cyclotron resonance heating (ICRH) are recognized as important auxiliary heating and current drive methods for present and next step fusion devices. However, these radio frequency (RF) systems generate a heat flux up to several MW/m 2 on the RF antennas during plasma operation. This paper focuses on the determination of the heat flux deposited on the lateral protections of the RF antennas in Tore Supra. The heat flux was calculated by finite element method (FEM) using a model of the lateral protection. The FEM calculation was based on surface temperature measurements using infrared cameras monitoring the RF antennas. The heat flux related to the acceleration of electrons in front of the LHCD grills (LHCD active) and to the acceleration of ions in RF-rectified sheath potentials (ICRH active) were calculated. Complementary results on the heat flux related to fast ions (ICRH active with a relatively low magnetic field) are also reported in this paper.

Using high-resolution diagnostics, this paper investigates experimentally two ways of influencing the radiofrequency (RF) sheath patterns (magnitude and spatial distribution) developing around active Tore Supra ion cyclotron antennae. Firstly two types of Faraday screen electrical designs were compared. The new design aimed at reducing the parallel RF electric field integrated along “long field lines” passing in front of the antenna, by interrupting all parallel RF current paths on its front face. This proved inefficient for attenuating the RF-sheaths on the screen itself, and only weakly modified their distribution. Secondly the ratio of strap RF voltage amplitudes was varied, either during spontaneous bifurcations of the antenna electrical operational point or through the power balance between RF generators. This affected the left–right symmetry of screen heat loads: the local fluxes tended to evolve as the RF voltages on the nearest strap. Both experiments question the method used to optimize the new screen.

International audience; The present paper deals with the surface heat flux estimation with embedded thermocouples (TC) in a Plasma Facing Component (PFC) of the WEST Tokamak. A 2D nonlinear unsteady calculation combined with the Conjugate Gradient Method (CGM) and the adjoint state is achieved in order to estimate the time evolution of the heat flux amplitude and decay length lambda_q. The method is applied on different synthetic measurements in order to evaluate the accuracy of the method. The synthetic measurements are generated with realistic values of lambda_q and magnitudes as those expected for ITER.

International audience; The peak power flux densities from plasma ion bombardment on tokamak armor surfaces can deviate from geometric-optical projection when the ion thermal Larmor radius becomes comparable to the characteristic size of the surface relief (gaps, misalignments, shaping). We present a fast and reliable tool to estimate the effect of finite ion Larmor radius on heat load deposition for complex 3D elements. The two possible geometrical design options foreseen for WEST tungsten monoblocks are asymmetric chamfer shaping and no shaping on top of monoblocks. This paper presents an evaluation of peak power flux densities for a WEST ITER-like tungsten monoblock plasma facing component which is vertically misaligned with respect to the other components of a maximum expected value of +0.3 mm, for steady state convected plasma power with ion temperature ranging from 10 to 500 eV. With the +0.5 mm asymmetric chamfer, leading edges between plasma facing components are shadowed and there is no significant difference between the optical calculations and the full ion orbit model. For an unshaped and +0.3 mm misaligned monoblock, when considering the full ion orbit model, the heat load on the exposed edge is reduced by a factor of 3-5 compared to the geometric-optical projection. Energy is found to be deposited over several millimetres on top of the monoblock just after the leading edge which prevents the tungsten armor from melting. (C) 2016 Elsevier B.V. All rights reserved.

The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m−2 for the required pulse length.

Tore Supra is an actively cooled superconducting tokamak allowing long pulse operation. In the last decade, two major upgrades were achieved. In 2002, all the Plasma Facing Components (PFCs) were replaced to withstand an injected power of 25 MW in steady state in the framework of the CIEL (Composants Internes et Limiteur) project. In 2012, a significant increase of the capabilities of the additional heating systems will be achieved, mainly on the LHH (Lower Hybride Heating) system, which will be able to inject up to 7 MW in the plasma during 1000 s (CIMES Chauffage et Injection de Matiere En Stationnaire project). This project also included the development and implementation of a new continuous pellet injection system. One of the objectives of the experimental program in 2012 and the following years is to perform long pulses with an injected power of up to 10 MW during 1000 s. The thermal power to be exhausted to the environment by the water cooling system, comes from the primary loops cooling the PFCs (up to 10 MW) and mainly from those that cool the RF heating system generators (10–20 MW). The secondary loop, the Heat Rejection System (HRS), has two cooling towers with an exhaust capacity of 11 MW in continuous operation. With the completion of CIMES, the water cooling system capability to withstand the overall injected energy in the new configuration of the machine and RF generators is questioned. This requires a detailed assessment of the operational limits of the cooling system. These limits depend on the wet-bulb temperature which governs the lowest achievable temperature level of the HRS. They will be different depending on whether the pulse is performed in summer or in winter. For this purpose, a transient thermo-hydraulic computer model of the cooling system was developed with matlab (CATSYS model). The simulations of already achieved long pulses with high injected energy were compared to the experimental measurements. The tool was then used to perform simulation and to compute the limits of the Tore Supra cooling system and to identify the equipments which have to be upgraded. The paper presents a description of the Tore Supra cooling system and main components as well as the CATSYS model. It will give the results of the experimental validation performed and the simulation results giving the limits of the analysed system. A new upgrade of the machine aiming at testing ITER W divertor components under the steady state heat fluxes expected in ITER is being investigated within the WEST project (tungsten (W) Environment in Steady-state Tokamak). It will lead to the replacement of the major part of the existing PFC's and of the dedicated PHTS. The CATSYS model will be used for the design of the new cooling system. Some preliminary results of the changes required for WEST will also be presented.

A Passive Active Multijunction (PAM) Lower Hybrid heating and Current Drive (LHCD) antenna has been installed on Tore Supra in 2009. The design and the fabrication of this antenna based on the passive-active multijunction concept is a major part of the CIMES project (components for the injection of mater and energy in steady-state) on Tore Supra. The antenna made of copper and stainless steel has to withstand important electromagnetic forces due to eddy currents during plasma disruption. 3D electromagnetic analyses have been performed with the ANSYS code to calculate the eddy currents and the electromagnetic forces.

A 20 MW Lower Hybrid Current Drive system using an antenna based on the Passive-Active Multijunction (PAM) concept is envisaged on ITER. This paper gives an overview of the mechanical analysis, modeling and design carried out on two major elements of the antenna: the grill front face, and the RF feed-through or windows. The front face will have to withstand high heat and fast neutrons fluxes directly from the plasma. It will be actively cooled and present a beryllium coating upon ITER requirement. The RF window being a critical safety importance class component (SIC) because of its tritium confinement function, two of them will be put in series on each line to achieve a double barrier. A design of a water cooled 5 GHz CW RF “pillbox” window capable of sustaining 500 kW of transmitted power is proposed. Both studies allow to move forward, and focus on critical issues, such as manufacturing processes and R&D associated programs including tests of mock-ups.

A new Passive Active Multijunction (PAM) Lower Hybrid heating and Current Drive (LHCD) launcher, has been successfully manufactured and tested on Tore-Supra (TS). The design and the fabrication of this new actively cooled launcher based on the PAM concept, as the present ITER LHCD design, is a major component of the TS CIMES project (Components for the Injection of Matter and Energy in Steady-state), and will play a key role in the TS near term program. To achieve 1000 s pulses with a power flux of 25 MW/m 2 the PAM launcher has been designed for steady state (CW) operation (active cooling) with the objective of coupling 2.7 MW of LHCD power to the plasma at 3.7 GHz with a parallel index N ∥ = 1.7 ± 0.2. The launcher has achieved its qualifications tests, i.e. low power Radio Frequency measurements, vacuum and hydraulic leak tests, and has been installed on Tore-Supra tokamak in September 2009. It is commissioning on plasma started a month later, quickly achieving its design performance of 2.7 MW on a 35 s pulse. After a technical description of the PAM, this paper presents an overview of the project phases (RF optimization, manufacturing and qualification) and concludes with the first experimental results of the PAM.

Since the mission of Tore Supra is to produce quasi-steady-state discharges, the lower hybrid current drive (LHCD) system constitutes the most important method of additional heating and noninductive current drive. A description of the LHCD system is given, including the different launcher designs developed for the Tore Supra long-pulse program. Following the completion of the Composants Internes et Limiteur project, together with the installation of a high-performance LHCD launcher, world record discharges, injected and extracted energy exceeding 1 GJ, were obtained in 2003. With the flexibility of lower hybrid (LH) waves to tailor the current profile, an enhanced performance regime, the so-called LHEP, has been maintained in quasi-steady-state discharges. Detailed measurements of the fast electron distribution have allowed us to constrain LHCD ray-tracing models and to quantify parametric dependencies describing the fast electron tail. Localized heat loads on the LHCD launchers due to interaction with fast particles have been measured and quantified, using infrared imaging and calorimetric measurements on water-cooled plasma facing components. Furthermore, experimental results in the area of LH wave coupling are presented.

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.

International audience; The nuclear fusion research goal is to demonstrate the feasibility of fusion power for peaceful purposes. In order to achieve the conditions similar to those expected in an electricity-generating fusion power plant, plasmas with a temperature of several hundreds of millions of degrees must be generated and sustained for long periods. For this purpose, RF antennas delivering multi-megawatts of power to magnetized confined plasma are commonly used in experimental tokamaks. In the gigahertz range of frequencies, high power phased arrays known as " Lower Hybrid " (LH) antennas are used to extend the plasma duration. This paper reviews some of the technological aspects of the LH antennas used in the Tore Supra tokamak and presents the current design of a proposed 20 MW LH system for the international experiment ITER.

International audience; With WEST (Tungsten Environment in Steady State Tokamak) (Bucalossi et al 2014 Fusion Eng. Des. 89 [http://dx.doi.org/10.1016/j.fusengdes.2014.01.062] 907?12 ), the Tore Supra facility and team expertise (Dumont et al 2014 Plasma Phys. Control. Fusion 56 [http://dx.doi.org/10.1088/0741-3335/56/7/075020] 075020 ) is used to pave the way towards ITER divertor procurement and operation. It consists in implementing a divertor configuration and installing ITER-like actively cooled tungsten monoblocks in the Tore Supra tokamak, taking full benefit of its unique long-pulse capability. WEST is a user facility platform, open to all ITER partners. This paper describes the physics basis of WEST: the estimated heat flux on the divertor target, the planned heating schemes, the expected behaviour of the L?H threshold and of the pedestal and the potential W sources. A series of operating scenarios has been modelled, showing that ITER-relevant heat fluxes on the divertor can be achieved in WEST long pulse H-mode plasmas.

Fusion devices with carbon as the main armour material are experiencing a growth in carbonaceous deposits at the surface of the plasma facing components. Tore Supra presents such deposits, and has specific features which influence their growth: long pulse operation and cooled walls. Deposits have a low thermal transfer to the cooled structure so that they appear as hot areas with the infrared imaging system looking at the elements surface temperature during plasma discharges. A 'degree of (carbon) deposit' on the toroidal pumped limiter is estimated by establishing the ratio between the apparent power on the limiter derived from the infrared measure and the actual one, deduced from a power balance analysis between the injected and the radiated power. This criterion is used to monitor the evolution of the deposit average thermal resistance. Successive shots have a similar 'degree of deposit', showing that the evaluation makes sense. Two years of data have been compiled (2003 and 2004), representing 3000 discharges (13 h of plasma, including 30 discharges longer than one minute). A three-fold increase in the 'degree of deposit' over six months is evidenced, following a limiter clean-up early in 2003. A comparison with calorimetric data produces a similar result, albeit less pronounced. Large steps in the degree of deposit are sometimes observed, usually correlated with identified events such as disruption, vessel opening, conditioning or plasma parameters change. It indicates that the deposit thermal resistance can change rapidly, although a systematic correlation with the above mentioned events could not be established.

Tore-Supra Tokamak ( I p = 1.5 MA, B t = 4 T) has been constructed with a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components (PFCs) for high-performance long pulse plasma discharges. When not actively cooled, plasma facing components can only accumulate a limited amount of energy since the temperature increases continuously during the discharge until radiation cooling equals the incoming heat flux. Such an environment is found in the JET Tokamak [JET Team, IAEA-CN-60/A1–3, Seville, 1994] and on TRIAM [M. Sakamoto, H. Nakashima, S. Kawasaki, A. Iyomasa, S.V. Kulkarni, M. Hasegawa, E. Jotaki, H. Zushi, K. Nakamura, K. Hanada, S. Itoh, Static and dynamic properties of wall recycling in TRIAM-1M, J. Nucl. Mater. 313–316 (2003) 519–523] [Y. Kamada, et al., Nucl. Fusion 3 (1999) 1845]. In Tore-Supra, the surface temperature of the actively cooled plasma facing components reach steady state within a second. We present here the Tore-Supra thermographic system, made of seven endoscope bodies equipped so far with eight infrared (IR) cameras. It has to be noted that this diagnostic is the first diagnostic to be actively cooled, as required for steady state. The main purpose of such a diagnostic is to prevent the plasma to damage the actively cooled plasma facing components (ACPFCs), which consist of the toroidal pumped limiter (TPL), 7 m 2 , and of five radio-frequency antennae, 1.5 m 2 each.

Tore-Supra is a Tokamak (Ip=1.5MA, Bt=4T) aiming at doing researches in the field of controlled nuclear fusion. It has been constructed with a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T˜sqrt(time)) during the discharge until radiation cooling is equal to the incoming heat flux (T>1800K). Such an environment is found in the most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra system is described in details, the new JET infrared thermography system is presented and some insight of the ITER set of visible/infrared endoscope is given.

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 10 MW/m 2 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.

An imaging spectrometer using a sapphire prism as dispersing element has been conceived at Tore Supra for the spectral range of 1–4 μm. It measures simultaneously at various wavelengths the temperature on distributed high heat-flux elements under plasma impact with 36 optical fibers, four of which are ZrF4 fibers. It employs an InSb focal plane array detector (256*320 pixels) behind a silicon filter and a ZnS window yielding a dynamic range of 200–1500 °C with 20 ms temporal resolution. The fiber transmission and the spatial variation of gain and background of the camera are calibrated using a light source with integrating sphere. With a blackbody source one determines the nonlinearity of the average gain and controls its stability during operation. The spectral dispersion of about 24 nm/pixel is determined with interference filters and controlled with a spectral lamp. The measurements at various wavelengths allow to determine the temperature distribution in the field of view.

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.

The data acquisition for the infrared measurement system on Tore-Supra is a key element in ensuring the supervision of the new actively-cooled plasma facing components of the CIEL project. It will allow us to follow the thermal evolution of components of Tore-Supra, in particular the toroidal pumped limiter (LPT) (360°–15 m long) and the five additional heating launchers. When fully installed, the infrared measurement system will be composed of 12 digital 16-bit infrared cameras. They cover a 100–1200 °C temperature range and each picture has a definition of 320×240 pixels with a 20 ms time resolution. The objectives of the data acquisition system is real-time recording and analysis of each view element for further post-pulse analysis in order to understand the physics phenomenon and ensure the supervision of the plasma facing components and also to be part of the global feedback control system of Tore Supra.

A short review of diagnostics and techniques for active power exhaust control is given from the point of view of experiences gained at Tore Supra. The physical and technological aspects are presented with the aim of illustrating strengths, difficulties and potential for future long pulse and steady state fusion devices such as W7X and ITER-FEAT. One of the main features of the Tore Supra approach to power exhaust control is the reliance on infrared (IR) thermography to survey nearly 100% of the high flux target zones. This is applied to the largest elements, the toroidal pump limiter and the heating antennae, by a number of IR endoscopes. For high flux target areas inaccessible to direct view, optical fibres in the IR and the near-IR range are used. The large wall panels, which evacuate the rest of the power, are controlled by calorimetry. Bolometry is used to achieve a complete power balance. These measurements can be inserted in control loops acting typically on the injected heating power or the radiation cooling.

Tore Supra has chosen to protect the most heavily solicited and actively cooled target plates by thermography using optical fibres. In consequence to previous experience [J. Nucl. Mater. 290–293 (2001) 701], the prototype periscope for the 2001 experimental campaign was equipped not only with silica fibres but also with a ZrF 4 fibre, transparent up to 4 μm. As then, the measurements in the mid infrared range give lower temperatures than the ones in the near infrared range. The silica fibres are more robust and easier to calibrate, which may justify to use both types side by side using a cross calibration formula. Measurements with an silica fibre embedded in the interior of the target plate exhibited only during discharges with LHCD heating sufficient signal levels to be spectrally analysed. Under these conditions the same kind of additional near infrared radiation as on the silica fibres looking at the exterior of the target plates was observed.

Tore Supra has re-started operation in 2001 with a partial toroidal pump limiter as the main plasma facing component. Assembled with actively cooled high heat flux elements, it has the shape of a toroidal ring. Thermal equilibrium is achieved within 4 s and the stabilised surface temperature is comparable to the design calculation. The observed heat flux pattern is similar to calculations realised beforehand with the heat deposition code TOKAFLU, and is dominated by the large toroidal field ripple. The power deposition is a combination of parallel convection and cross-field transport of similar magnitudes. The heat flux decay length is evaluated by three independent diagnostics to 10 mm. The excellent agreement between experiment and modelling indicates that the description of the power deposition mechanism is reasonable, although parameters such as the heat flux decay length and fraction of perpendicular power are still difficult to predict.

In order to assist the design of the future ITER Lower Hybrid launcher, coupling codes ALOHA, from CEA Cadarache, TOPLHA, from Politecnico di Torino, and GRILL3D (Dr.Mikhail Irzak, A.F.Ioffe Physico-Technical Institute, St. Petersburg, Russia), from ENEA Frascati, have been compared with the initial (3 modules with 8 active waveguides per module) and updated (6 modules with 4 active waveguides per module) Passive-Active Multijunction (PAM) Lower Hybrid antennas. Both ALOHA and GRILL3D formulate the problem in terms of rectangular waveguides modes, while TOPLHA is based on boundary-value problem with the adoption of a triangular cell-mesh to represent the relevant waveguides surfaces. Several plasma profiles, with varying edge density and density increase, have been adopted to provide a complete description of the simulated launcher in terms of reflection coefficient, computed at the beginning of each LH module, and of power spectra. Good agreement can be observed among codes for all the simulated profiles.

The present work has been developed within the frame of the EFDA task “HCD-08-03-01: LH4IT, EU contribution to the ITER LHCD Development Plan” The use of rectangular oversized waveguides in the Main Transmission Lines (MTLs) of the Lower Hybrid Current Drive (LHCD) system of ITER, requires to investigate the problem of bends. The high number of involved waveguides (from 24 to 48) must be also taken into account. Thus, it has to consider not only the best choice in terms of curved framework, but also the proper allocation of all the waveguides. In this context, the principal specifications that characterize the design of the bends are: a) to minimize the reflection of the fundamental TE10 mode; b) to maximize the transmission of the fundamental TE10 mode; c) to minimize the coupling between the TE10 mode and other spurious modes that propagate at 5 GHz. This paper presents an overview about the bend options, and it compares the performances of several frameworks analyzed by using the Finite Element Method (FEM) commercial software, HFSS®. First of all, simple circular trajectory curves with different angulations, are considered. Then, the so called Mitre Bends alternatives are deeply analyzed. These curves are studied by several authors in the mono-modal configuration, with different techniques but the propagation in an oversized environment is a topic not much attended in literature. The only design parameter of the simple circular trajectory bend is the bending radius, so that the design is not flexible; the Mitre Bend structure is at least more flexible than the previous one and it is of great interest to study this type of bend to check the possible advantages. Finally an innovative modified Mitre Bend solution based on a cascade of trapezoidal elements is proposed.

A 20 MW Lower Hybrid Current Drive system using an antenna based on the Passive Active Multijunction (PAM) concept is envisaged on ITER for steady state operation, including the current ramp-up phase. In the frame of an EFDA task, a conceptual design of such a system has been performed. This paper gives an insight of the mechanical analysis, modeling and design carried out on two elements of the antenna: the front face, and the RF windows which constitute the first tritium barrier. The front face of the antenna will have to withstand high heat and fast neutrons fluxes directly from the plasma. It will be actively cooled and present a beryllium coating upon ITER requirement. The main issues were to ensure a reasonable margin versus a maximum temperature fixed by safety concerns (650 °C for plasma facing beryllium), and to understand and assess resistance of the assembly while in operation. Analyses were conducted on some variants of this PAM. Manufacturing scenarios considering ITER requirements and involving various techniques (brazing, high isostatic pressure bonding, explosion bonding) were explored, and simulations over subsequent residual stresses were conducted. Analyses showed that surface temperatures do not exceed 500°C, and stresses should not affect the integrity of the antenna mouth. The RF window is a critical component since it also has a safety function. We planed to design a water cooled 5 GHz CW RF “pillbox” window capable of sustaining 500 kW of transmitted power. An RF optimization conducted on this concept allowed to combine good RF properties and low dielectric losses. It resulted in a maximum thermal stress in the ceramic of 31 MPa, well below the static fatigue limit of 50 MPa. The residual stresses resulting from the brazing methods involved in the manufacturing of this multi-material assembly (copper, copper alloys and ceramic) were also analyzed. Both studies allow us to move forward and focus on critical issues, such as manufacturing processes (beryllium, bonding techniques…) and R&D associated programs including test of mock-ups.

A very preliminary design of the Lower Hybrid Current Drive system for ITER was developed in the early years of the 2000's. Many parameters of the system were defined on the base of knowledge and technology of those times. Experimental results obtained in the last few years, like the successful tests of a PAM launcher on FTU and on Tore Supra have definitely indicated this concept as a possible candidate for the LHCD launcher for ITER. The successful development of a prototype klystron at 5GHz with a target RF power of 500kW CW has indicated this power as an upper limit for the present high vacuum electron tubes technology at this frequency. From the physics point of view the ITER operational scenarios have been well defined, while the last experiments have pointed out the effectiveness of the LHCD waves at high plasma density. All these results imposed a revision of the existing design of the LHCD system. The RF power limited to a maximum of 500kW per klystron has determined a new modularity of the whole system and of the PAM launcher. In addition a relatively shorter distance between klystrons and ITER machine suggested the possibility of using rectangular oversized waveguides, excited in the fundamental TE01 mode, for the realization of the Main Transmission Lines (MTLs). Therefore all the main microwave components of the system have been revised on these basis and different and more suitable solutions have been investigated. In particular this paper compares the performances of the two different MTLs, based respectively on circular and rectangular waveguides, describes the main components of the proposed rectangular MTL and analyzes their main microwave characteristics. Companion papers in this Conference cover all the other aspects of the updated design of the LHCD system for ITER. The present revision has been carried out in the frame of the EFDA task “HCD-08-03-01: EU Contribution to the ITER LHCD Development Plan (LH4IT)”.

A time-of-flight diagnostic made of a double sheet sensor has been designed for the search of ions emitted by an electron–photon conversion target under the impact of an intense relativistic electron beam. This device was tested on the PIVAIR prototype accelerator. Signals attributed to ions emitted from the target were observed for different target elements. The ions C+ and H+ were identified by time-of-flight measurements, and their mean velocities were found to be equal to about 1.6×108 and 5×108 cm/s for C and Ta targets, respectively. The appearance, rise time and saturation values of the ionic current are close to those obtained from preliminary calculations based on an ionic emission model (VAPOR) which has been coupled to a Maxwell–Vlasov PIC code (M2V). The ionic current signal is characterised by a plateau, the level of which corresponds to the Child–Langmuir current (25 A for H+ ions). Furthermore, the fall-off of this signal is synchronised with the electron beam one. This behaviour represents the de-confinement of the ion beam under the effect of its own space charge.

The data acquisition for the infrared measurement system on Tore-Supra is a key element to ensure the supervision of the new actively cooled plasma-facing components of CIEL (Inner Components and Limiter), which is a major upgrade of Tore Supra. It will allow us to follow the thermal evolution of components of Tore-Supra, in particular the toroidal pumped limiter (LPT) (360/spl deg/-15 m long) and the five additional heating launchers. When fully installed, the infrared measurement system will be composed of 12 digital 14-bit infrared cameras. They cover a 100-1200/spl deg/C temperature range, and each picture has a definition of 320 /spl times/ 240 pixels with a 20 ms time resolution. The objectives of the data acquisition system are real-time recording and analysis of each view element for further postpulse analysis in order to understand the physics phenomenon and ensure the supervision of the plasma facing components, and also to be part of the global feedback control system of Tore Supra.