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The National Spherical Tokamak Experiment (NSTX) upgrade (NSTX-U) requirements lead to enhanced heat loads on plasma-facing components (PFCs) especially in the divertor regions, where normal heat flux density can reach up to 8 MW/m2. For these high heat flux (HHF) regions, the PFCs were redesigned, to use castellations which relieved the thermal stresses associated with high incident heat flux. Another design feature of HHF PFCs is the absence of front face mounting holes which create localized areas of high temperature and thermal stress concentrations. Optimized shaping of the front face of the HHF PFCs eliminates regions where the front face is perpendicular to the heat flux direction during normal helicity operation and spreads the heat load. A unique mechanism was designed to connect PFC tiles to the NSTX-U center stack casing using locking features accessible from the low heat flux regions. Isotropic graphite was selected as the HHF PFC tile material. Initial tile design parameters were accessed using analytical expressions for pulsed heat flux loading. A working prototype of the locking mechanism was created during the initial stages of the design to prove the concept performance. This article presents an overview of the divertor HHF PFC design and the results of the thermal and structural analyses performed using ANSYS software. The results of the analyses cover normal operating conditions and disruptions which impose electromagnetic (EM) loads from eddy and halo currents. The 3-D, transient, nonlinear analyses took into account the temperature-dependent properties of the materials, friction interfaces between the parts, and variable electric properties of the parts and interfaces. The results confirm that the HHF PFC tiles remain within the allowable limits for the loads defined by the NSTX-U Recovery Project. In addition, tolerance stack up analyses were performed to ensure tile performance in the worst possible assembly configuration. The design process was completed successfully, and the NSTX HHF PFC tiles are currently in the production phase.

The optimized superconducting stellarator device Wendelstein 7-X (with major radius , minor radius , and plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of with central electron temperatures were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of a temperature of with was transiently accomplished, which corresponds to with a peak diamagnetic energy of and volume-averaged normalized plasma pressure . The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above line integrated density and central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

The Wendelstein-7X stellarator will go through three operational phases, with successively more energy as more complex divertors are installed. The first operational phase operated without a divertor, instead using five limiters (one in each of the five field periods), the second phase uses ten uncooled test divertor units (TDU), and the third will include an actively cooled divertor. Simulations of plasma operation have shown that the evolution of a toroidal current can lead to significant heat fluxes sweeping across sensitive regions in the divertor, possibly exceeding their qualified limit. A “Scraper Element” (SE) has been proposed to protect these sensitive divertor regions. To examine the effectiveness of this high heat-flux SE, a TDU Scraper Element (TDU-SE) is scheduled to operate with the TDU. A thermal analysis has been performed to correlate to measurements to be obtained during operation to evaluate the location and intensity of the convective plasma flux. The analysis will give an understanding of what is occurring at the plasma edge. It will also allow for a quick evaluation of the safety margin during initial, low power pulses so that the experiment can nimbly move to higher power operation.

In order to get first insight into net tungsten erosion in W7-X, tungsten (W) marker layers were exposed during the operational phase OP 1.2b at one position of the Test Divertor Unit (TDU), at 21 different positions of the inner heat shield, and at two scraper elements. The maximum tungsten erosion rate at the TDU strike line was 0.13 nm s−1 averaged over the whole campaign. The erosion was inhomogeneous on a microscopic scale, with higher erosion on ridges of the rough surface inclined towards the plasma and deposition of hydrocarbon layers in the recessed areas of the rough surface. The W erosion at the inner heat shield was below the detection limit of 3–6 × 1012 W-atoms/cm2s, and all inner heat shield tiles were covered with a thin B/C/O layer with thickness in the range 2 × 1017–1018 B + C atoms/cm2 (about 20–100 nm B + C). W-erosion of the marker layers on the scraper elements was also below the detection limit.

The generic upper port plug (GUPP) structure in ITER is a 6 m long metal box which deploys diagnostic components into the vacuum vessel. This structure is commonly used for all the diagnostic upper ports. The final design of the GUPP structure, which has successfully passed the final design review in 2013, is described here. The diagnostic port plug is cantilevered to the vacuum vessel with a heavy payload at the front, so called the diagnostic first wall (DFW) and the diagnostic shield module (DSM). Most of electromagnetic (EM) load (∼80%) occurs in DFW/DSM. Therefore, the mounting design to transfer the EM load from DFW/DSM to the GUPP structure is challenging, which should also comply with thermal expansion and tolerance for assembly and manufacturing. Another key design parameter to be considered is the gap between the port plug and the vacuum vessel port. The gap should be large enough to accommodate the remote handling of the heavy port plug (max. 25 t), the structural deflection due to external loads and machine assembly tolerance. At the same time, the gap should be minimized to stop the neutron streaming according to the ALARA (as low as reasonably achievable) principle. With these design constraints, the GUPP structure should also provide space for diagnostic integration as much as possible. This requirement has led to the single wall structure having the gun-drilled water channels inside the structure. Furthermore, intensive efforts have been made on the manufacturing study including material selection, manufacturing codes and French regulation related to nuclear equipment and safety. All these main design and manufacturing aspects are discussed in this paper, including requirements, interfaces, loads and structural assessment and maintenance.

Due to this position close to the plasma, the port plug structure and the diagnostic first wall (DFW) contain water to allow cooling during operation and for heating during bake-out. To remove the heat coming from the plasma due to radiation and neutrons, the pressure inside these structures should be up to 44 bars. On the other hand, the dominant load expected to drive the design of these structures is of electromagnetic origin during the plasma disruption. Description of the loads acting on DFWs and generic port plug structures and the significance of the load due to the water pressure, with implications on the design and inspection, are discussed in this paper.

ITER will have a set of 45 diagnostics to ensure controlled operation. Many of them are integrated in the ITER ports. This paper addresses the integration process of the diagnostic systems and the approach taken to enable coordinated progress. An overview of the Port Integration hardware introduces the various structures needed for hosting tenant systems inside ITER diagnostics ports. The responsibilities of the different parties involved (ITER Organization and the Domestic Agencies) are outlined. The main challenges for diagnostic port integration engineering are summarized. The plan for a common approach to design and manufacture of the supporting structures, in particular the Port Plug is detailed. A coordinated design including common components and a common approach for neutronic analyses is proposed. One particular port, the equatorial port 11, is used to illustrate the approach.

The vacuum vessel (VV) design is being finalized including interface components, such as the support rails and feedthroughs of coils for mitigation of edge localized modes (ELM) and vertical stabilization (VS) of the plasma (ELM/VS coils). It was necessary to make adjustments in the locations of the blanket supports and manifolds to accommodate the design modifications in the ELM/VS coils. The lower port gussets were reinforced to keep a sufficient margin under the increased VV load conditions. The VV support design is being finalized as well, with an emphasis on structure simplification. The design of the in-wall shielding (IWS) has progressed, considering assembly and required tolerances. The layout of ferritic steel plates and borated steel plates will be optimized based on on-going toroidal field ripple analysis. The VV instrumentation was defined in detail. Strain gauges, thermocouples, displacement meters and accelerometers shall be installed to monitor the status of the VV in normal and off-normal conditions to confirm all safety functions are performed correctly. The ITER VV design was preliminarily approved, and the VV materials including 316L(N) IG were already qualified by the Agreed Notified Body (ANB) according to the procedure of Nuclear Pressure Equipment Order.

Procurement arrangements for ITER key components including the vacuum vessel (VV) have been signed and the ITER activities are now fully devoted towards construction. Final design reviews have been carried out for the main vessel and ports. One of the design review topics is the selection of materials, material procurement, and assessment of material performance during operation. The width of the inner shell splice plates was increased from 120 mm to 160 mm to minimize risk during the assembly of the Thermal shields and the VV. Instead of facet shaping, 3D shaping was introduced for the outboard inner shell. The material qualification procedures have been started for VV structural materials such as 316L(N) IG for licensing as a nuclear pressure equipment component. In accordance with the regulatory requirements and quality requirements for operation, common material specifications have been prepared in collaboration with the domestic agencies.

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.

After implementing a few design modifications (referred to as the “Modified Reference Design”) in 2009, the Vacuum Vessel (VV) design had been stabilized. The VV design is being finalized, including interface components such as support rails and feedthroughs for the in-vessel coils. It is necessary to make adjustments to the locations of the blanket supports and manifolds to accommodate design modifications to the in-vessel coils. The VV support design is also being finalized considering a structural simplification. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. The detailed layout of ferritic steel plates and borated steel plates was optimized based on the toroidal field ripple analysis. A dynamic test on the inter-modular key to support the blanket modules was performed to measure the dynamic amplification factor (DAF). An R&D program has started to select and qualify the welding and cutting processes for the port flange lip seal. The ITER VV material 316 L(N) IG was already qualified and the Modified Reference Design was approved by the Agreed Notified Body (ANB) in accordance with the Nuclear Pressure Equipment Order procedure.

Eddy current induced electromagnetic (EM) loads during plasma events are the design driver for ITER port plug (PP) structure, diagnostic first wall (DFW), shield module (DSM) and diagnostic system supported by the PP structure. Generic models using commercial software OPERA, MAXWELL and ANSYS are developed and benchmarks are performed for global EM analysis to obtain port-specific design driving EM loads. The 20 degree vessel sector models of the upper and equatorial PP structure take the same ITER TF, CS and PF coil and plasma current (15 MA baseline plasma scenario) as input, then solve for eddy current induced on all passive structural components for various DINA disruption cases. The worst load case can be exponential decay or linear decay of plasma current depending on the dimension and location of diagnostic components inside the PPs. Static and transient magnetic fields from generic models are mapped onto an excel datasheet to establish component design load specification. Three levels of modeling effort are suggested. The global model analysis is used to validate the impact of component design to the PP global system response as well as to study its component dynamic effect as a result of the disruption loads on the PP structure. The local sub-model analysis can be used to extract more accurate EM loads on diagnostics and the excel datasheet of static and transient field maps are used as the initial design load specification for in-port components.

Princeton Plasma Physics Laboratory and the Massachusetts Institute of Technology are preparing an experiment of current profile control using lower hybrid waves to produce and sustain advanced tokamak regimes in steady-state conditions in Alcator C-Mod. Unlike the Joint European Torus, ToreSupra, and JT60 couplers, the C-Mod lower hybrid coupler does not employ the now conventional multijunction design but will have similar characteristics, compactness, and internal power division while retaining full control of the antenna element phasing. This is achieved by using 3-dB vertical power splitters and a stack of laminated plates with the waveguides milled in them. Construction is simplified and allows easy control and maintenance of all parts. Many precautions are taken to avoid arcing. Special care is also taken to avoid the recycling of reflected power, which could affect the coupling and the launched n.

A frequency modulated Coherent Laser Radar ranging diagnostic is being used on the National Spherical Torus Experiment (NSTX) for precision metrology. The distance (range) between the 1.5 μm laser source and the target is measured by the shift in frequency of the linearly modulated beam reflected off the target. The range can be measured to a precision of

The design of the ITER blanket is presented together with the related technology which has been developed. The evolution of this component since the beginning of the EDA is explained in relation to the developing understanding of the thermal deformations and of the electromagnetic forces. These loads lead to a system composed of compact modules protecting a continuous support shell called a backplate. The backplate is a stiff double wall construction which conveys the coolant to the modules. The supports of the module are flexible and allow relative thermal expansions. They are connected and disconnected to the backplate by bolts operated through holes in the front face of the module. The coolant connections and the electrical straps located on the back of the modules are reached similarly. The first wall is integral with the module and cooled in series. A research and development program on materials and joining methods defined the construction path which has been tested in prototypes. The main body is built of stainless steel by forging and drilling or powder hot isostatic pressing (HIP), depending on the complexity of the shape. The first wall includes a dispersion strengthened copper heat sink which is hot isostatic pressed onto the steel body. Beryllium is the basic plasma facing material and is attached by HIP to the copper. Prototypes of the module attachment have been built and are under integrated tests.

The ITER blanket system removes the surface heat flux from the plasma and from bulk heating by the neutrons, reduces the activity in the vacuum vessel (W) structural material to the level allowable...

The blanket system comprises first wall/shield modules which are supported on a structural shell called a Back Plate containing water channels for cooling the modules. A modular design of the first wall/shield was chosen to allow independent maintenance through the horizontal mid-plane ports using in-vessel remote handling equipment. The modules are mechanically attached to the back plate for assembly and maintenance. The first wall/shield modules can be replaced with breeding modules for the EPP. In the first wall (FW) region, the water is contained in 316 LN stainless steel pipes surrounded by a copper heat sink except in high heat flux regions where copper pipes with a stainless steel liner are used in the limiter and baffles. The plasma facing surface of the FW will be beryllium except for the lower region of the baffles, where tungsten is used. Electromagnetic analysis and structural analysis has been performed on the back plate and FW/shield modules during a plasma disruption and for a VDE with toroidally asymmetric halo currents. A major R and D project is being conducted to provide needed input for design of the blanket system and to confirm the fabrication technology.

The International Thermonuclear Experimental Reactor (ITER) breeding blanket is designed to breed the necessary tritium for ITER operation during the enhanced performance phase by replacing the shielding blanket of the basic performance phase. Similar to the shielding blanket, it has to remove the majority of the fusion power generated by the plasma and to protect the vacuum vessel and the toroidal field coils from excessive nuclear heating and radiation damage. It has to produce a net tritium breeding ratio of more than 0.8 to satisfy the technical objectives of the enhanced performance phase. For compatibility with the ITER design and to satisfy the blanket functional requirements, a water-cooled modular solid breeder blanket with a beryllium neutron multiplier has been selected. Lithium zirconate is the reference breeder material based on the current database. Enriched lithium is used to enhance the tritium breeding capability, to reduce the radial blanket thickness, to decrease the breeder material volume, to lower the breeder thermal stresses, and to enhance the shielding capability. Similar to the shielding blanket, the breeding blanket uses Type 316LN-IG austenitic steel structural material. Both forms of beryllium material, porous and pebbles, are used at different blanket locations based on design requirements. This paper is concerned with the design analyses and design selections, including beryllium form, breeder material, tritium breeding, and heat transfer across the beryllium–steel interface. Also, the required research and development tasks for the ITER breeding blanket are summarized.

The ITER diagnostics generic upper port plug (GUPP) is developed as a standardized design for all diagnostic upper port plugs, in which a variety of payloads can be mounted. Here, the remote handling compatibility analysis (RHCA) of the GUPP design is presented that was performed for the GUPP final design review. The analysis focuses mainly on the insertion and extraction procedure of the diagnostic shield module (DSM), a removable cassette that contains the diagnostic in-vessel components. It is foreseen that the DSM is a replaceable component – the procedure of which is to be performed inside the ITER hot cell facility (HCF), where the GUPP can be oriented in a vertical position. The DSM removal procedure in the HCF consists of removing locking pins, an M30 sized shoulder bolt and two electrical straps through the use of a dexterous manipulator, after which the DSM is lifted out of the GUPP by an overhead crane. For optimum access to its internals, the DSM is mounted in a handling device. The insertion of a new or refurbished DSM follows the reverse procedure. The RHCA shows that the GUPP design requires a moderate amount of changes to become fully compatible with RH maintenance requirements.

A TFTR Lower Hybrid Current Drive Project has been undertaken to scope out the design and the details of construction of a Lower Hybrid (LH) system to provide up to 4 megawatts of 4.6 GHz rf source...

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma.The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral‐beam heating, in a supershot discharge and 6.7 MW in a high‐βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter‐H‐mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high‐βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first‐orbit loss mechanism. No loss due to alpha‐particle‐driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

During the past three Tokamak Fusion Test Reactor (TFTR) vacuum vessel machine openings, an extensive effort was undertaken to optimize the distribution of heating of the bumper limiter tiles. The ...

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fullymore » explored.« less

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium–deuterium (D–D) run to simulate expected deuterium–tritium (D–T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D–T experiments using D–D have been performed. The physics objectives of D–T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium–tritium plasmas, evaluation of α‐particle heating of electrons, and collective effects driven by alpha particles and testing of diag...

Carbon fiber composite (CFC) materials have been used as the plasma facing surface on limiters in TFTR since 1989. We changed from all POCO graphite tiles on the inner bumper limiter (BL) to about ...

ELM mitigation is of particular importance in ITER in order to prevent rapid erosion or melting of the divertor surface, with the consequent risk of water leaks, increased plasma impurity content and disruptivity. Exploitable “natural” small or no ELM regimes might yet be found which extrapolate to ITER but this cannot be depended upon. Resonant Magnetic Perturbation has been added to pellet pacing as a tool for ITER to mitigate ELMs. Both are required, since neither method is fully developed and much work remains to be done. In addition, in-vessel coils enable vertical stabilization and RWM control. For these reasons, in-vessel coils (IVCs) are being designed for ITER to provide control of Edge Localized Modes (ELMs) in addition to providing control of moderately unstable resistive wall modes (RWMs) and the vertical stability (VS) of the plasma.

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation. © 2009 IAEA, Vienna.

The TFTR Bumper Limiter (BL)1,2 is a continuous toroidal belt limiter which subtends 120° poloidally on the inner wall of the vacuum vessel, is symmetric about the midplane of the vessel and has a total area of 22 m2. The plasma facing surface of the BL consists of graphite tiles mounted on water-cooled Inconel backing plates. During plasma operation with high power neutral beam injection (NBI), “carbon blooms”3 developed when surface temperatures at localized areas on the limiter exceeded 1700 C. The blooms severely limited plasma performance. During the February–April 1990 shutdown, 721 of the original POCO graphite tiles were replaced with 4D Carbon Fiber Composite (CFC)4 tiles to improve the performance of the Bumper Limiter. Another source of blooms was cutouts in the BL made for diagnostic access to the plasma. During the shutdown, the edge tiles at the cutouts were modified to reduce the power flux to levels insufficient to cause a bloom. This paper describes the tile replacement and modifi...

Previous tests of a High Power Prototype (HPP) comprising one quadrant of the JET ITER‐Like ICRF Antenna have indicated the need for some design modifications in order to achieve 10 s pulses coupling the full design power (7.1 MW) into the reference plasma load (R′ = 4 Ω/m). These modifications have now been made to the HPP, as well as to the design of the ITER‐Like Antenna itself. In particular, maximum current densities have been reduced or otherwise accommodated in key areas. New current straps for the HPP have been fabricated from stereo‐lithography‐based investment castings. Design modifications to the antenna enclosure have also been implemented. This work has been materially assisted through the use of CST Microwave Studio (MWS), a commercially available 3‐D electromagnetic modeling package. Essentially the full engineering CAD model of the HPP current straps and antenna enclosure has been ex‐ported from ProE to MWS. Computed current density profiles have been introduced into an ANSYS thermal model. These activities will be discussed, as well as the current status of the HPP test program.

Three potential methods for evaluating the surface tritium content of the TFTR vacuum vessel are described, each based on a different technique for measuring the in situ beta emission from tritium. These methods should be able to provide both a local and a global assessment of the tritium content within the top {approximately}1{mu}m of the inner wall surface.

Recent Alcator C-Mod experimental campaigns have focused upon the study of the Advanced Tokamak regimes, which includes characterization of the RF heating, the formation and dynamics of internal barriers, H-mode edge pedestal, and divertor and scrape-off physics. The ICRF system has been recently upgraded with the improved performance of the 4-strap antenna. Total ICRF power in excess of 5 MW has been launched successfully into the plasma during this campaign. Due to the compact nature of C-Mod, the power feeds for the antenna are vacuum strip lines. Their orientation, to the tokamak B-field, is governed by maintaining E

The paper reports on the status of the JET ITER‐Like ICRF Antenna project and highlights main challenges that have come up during its design phase.

ITER ECE diagnostic [1] needs not only to meet measurement requirements, but also to withstand various loads, such as electromagnetic, mechanical, neutronic and thermal, and to be protected from stray ECH radiation at 170 GHz and other millimeter wave emission, like Collective Thomson scattering which is planned to operate at 60 GHz. Same or similar loads will be applied to other millimetre-wave diagnostics [2], located both in-vessel and in-port plugs. These loads must be taken into account throughout the design phases of the ECE and other microwave diagnostics to ensure their structural integrity and maintainability. The integration of microwave diagnostics with other ITER systems is another challenging activity which is currently ongoing through port integration and in-vessel integration work. Port Integration has to address the maintenance and the safety aspects of diagnostics, too. Engineering solutions which are being developed to support and to operate ITER ECE diagnostic, whilst complying with safety and maintenance requirements, are discussed in this paper.

Significant improvements of the ITER blanket were possible with the better definition of the thermal and electromagnetic effects. A supporting double wall backplate with reinforcement about the ports has been defined. The module has been simplified and the intrusion of the connections reduced. The attachment has been assessed by system analyses. The cooling system is configured to simplify the leak detection.

The Princeton Plasma Physics Laboratory (PPPL) and the Massachusetts Institute of Technology (MIT) are preparing an experiment of current profile control using lower-hybrid waves in order to produce and sustain advanced tokamak regimes in steady-state conditions in Alcator C-Mod. Unlike JET's, ToreSupra's and JT60's couplers, the C-Mod lower-hybrid coupler does not employ the now conventional multijunction design, but will have similar characteristics, compactness, and internal power division while retaining full control of the antenna element phasing. This is achieved by using 3 dB vertical power splitters and a stack of laminated plates with the waveguides milled in them. Construction is simplified and allows easy control and maintenance of all parts. Many precautions are taken to avoid arcing. Special care is also taken to avoid the recycling of reflected power which could affect the coupling and the launched n(subscript ||) spectrum. The results from C-Mod should allow further simplification in the designs of the coupler planned for KSTAR (Korea Superconducting Tokamak Advanced Research) and ITER (International Thermonuclear Experimental Reactor).

A Lower Hybrid Current Drive system is being constructed jointly by Plasma Science and Fusion Center (PSFC) and Princeton Plasma Physics Laboratory (PPPL) for installation on the Alcator C-Mod tokamak, with the primary goal of driving plasma current in the outer region of the plasma. The Lower Hybrid (LH) system consists of 3 MW power at 4.6 GHz with a maximum pulse length of 5 seconds. Twelve klystrons will feed an array of 4-vertical and 24-horizontal waveguides mounted in one equatorial port. The coupler will incorporate some compact characteristics of the multijunction power splitting while retaining full control of the toroidal phase. In addition a dynamic phase control system will allow feedback stabilization of MHD modes. The desire to avoid possible waveguide breakdown and the need for compactness have resulted in some innovative technical solution which will be presented.

A frequency modulated Coherent Laser Radar ranging diagnostic is being used on the National Spherical Torus Experiment (NSTX) for precision metrology. The distance (range) between the 1.5 {micro}m laser source and the target is measured by the shift in frequency of the linearly modulated beam reflected off the target. The range can be measured to a precision of < 100{micro}m at distances of up to 22 meters. A description is given of the geometry and procedure for measuring NSTX interior and exterior surfaces during open vessel conditions, and the results of measurements are elaborated.

A peak fusion power production of 9.3±0.7 MW has been achieved on the Tokamak Fusion Test Reactor (TFTR) in deuterium plasmas heated by co and counter injected deuterium and tritium neutral beams with a total power of 33.7 MW. The ratio of fusion power output to heating power input is 0.27. At the time of the highest neutron flux the plasma conditions are: Te(0)=11.5 keV, Ti(0)=44 keV, ne(0)=8.5×1019 m−3, and 〈Zeff〉=2.2 giving τE=0.24 s. These conditions are similar to those found in the highest confinement deuterium plasmas. The measured D‐T neutron yield is within 7% of computer code estimates based on profile measurements and within experimental uncertainties. These plasmas have an inferred central fusion alpha fraction of 0.2% and central fusion power density of 2 MW/m3 similar to that expected in a fusion reactor. Even though the alpha velocity exceeds the Alfven velocity throughout the time of high neutron output in most high power plasmas, MHD activity is not substantially different from that in co...

TFTR plans to begin D—T experiments in mid 1990. The D—T experimental program will produce approximately one hundred shots, with a neutron generation rate of 10 19 neutrons per shot. This will result in high levels of activation in TFTR, especially in the vacuum vessel. The primary purpose of the Maintenance Manipulator is to provide a means of remotely performing certain defined maintenance and inspection tasks inside the vacuum torus so as to minimize personnel exposure to radiation. The manipulator consists of a six-link folding boom connected to a fixed boom on a movable carriage. The entire manipulator is housed in a vacuum antechamber connected to the vacuum torus, through a port formerly used for a vacuum pumping duct. The configuration extends 180° in either direction to provide complete coverage of the torus. The four 3500 1/s turbopumps which were formerly used in the pumping duct will be mounted on the antechamber. The manipulator will utilize two end effectors. The first, called a General Inspection Arm (GIA) provides a moveable platform to an inspection camera and an in-vacuum leak detector. The second is a bilateral, force-reflecting pair of slave arms which utilize specially developed tools to perform several maintenance functions. All components except the slave arms are capable of operating in TFTR's vacuum environment and during 150°C bakeout of the torus.

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3 × 1019 m−3 , central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre. EURATOM 633053