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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.

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.

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.

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 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.

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.