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The CFETR is the next generation magnetic confinement tokamak, its increasing heating needs and stringent shielding requirements present new challenges for the ECRH launcher. In an effort to evaluate the viability of the equatorial launcher, a port plug prototype is designed and its performance under steady state or some interrupted events is assessed. The well-configured steel/water ratio is beneficial to improve its shielding capability and effectively mitigate the 493 kW heat. The maximum temperature of the port plug is less than 300 °C when the inlet temperature is 70 °C, and the average outlet water temperature reaches 142 °C. The worst EM case is a major disruption of the plasma, which would induce about 200 kN force on the port plug at the end of the plasma current linear decay. The combined heat and EM force loads on the port plug would result in a total deformation of 1.7 mm and a maximum equivalent stress of 320 MPa, with little impacts on the port plug or the surrounding vacuum vessel. As the installation of the port plug, the H and He generation rates of the surrounding vacuum vessel have increased by about 2 % and 1 % separately, but remain lower than the limit of 1 appm/FPY, and the nuclear heat density of the toroidal field coils has increased by about 5%.

The Diagnostic Shielding Modules (DSM) are secondary containers of the ITER diagnostic Port Plugs where shielding and diagnostics components have to be integrated together. For the development of Equatorial Port DSMs several key requirements have to be met by the Port Integrators. First of all, the total dry Plug weight must not exceed the allowable weight (currently 48 ton) in order to guarantee the consistency with the specification of other interfacing systems involved in the Port Plugs maintenance. Secondly, a shielding capability able to limit Shut-Down Dose Rate in the Interspace Zone (man access corridors) below 100 μSv/h needed to allow the human accessibility for the Port Plug sealing disassembly, the inspection and the maintenance of diagnostic components behind the Plug closure plate and in the interspace. Finally, Remote Handling (RH) compatibility for refurbishment of the Port Plug and maintenance of systems integrated is necessary as well. These operations involve activated components that are processed in the Hot Cell. As a result of these demands, the ITER Port Integration team has developed the so called Modular Equatorial DSM able to answer the weight/shielding/RH requirements described above following the ALARA principle and offering a standard infrastructure for the common DSM integration problems. [ABSTRACT FROM AUTHOR]

Forschungszentrum Jülich and partners are developing the ITER core charge exchange spectroscopy (cCXRS) hosted by upper port plug#3 (UPP). The system should withstand loading due to the plasma transients resulting in electromagnetic (EM) forces acting on the UPP and its onboard components. On the other hand, the eddy currents and shock dynamic EM loads occur in the vacuum vessel (VV) and make it vibrate. As a result, it excites the UPP and, consequently, its onboard components. The computational problem is a superposition of these solutions: 1) the deterministically calculated time-history UPP response to the applied EM forces and 2) the UPP response to the VV excitations that are specified as floor response spectra (FRS) at the port stub (attachment to the VV). On the top of this, the EM loads should be combined with the seismic ones which are also specified as an FRS. This paper considers a potential methodology for combining plasma transients with seismic events. A numerical modeling of the UPP hosting a specific cCXRS component is presented. This methodology gives a direct and transparent engineering way to estimate the mechanical strength of the UPP onboard components. The analysis uses port stub FRS input and does not depend on a spectra-to-spectra recalculation procedure (from VV to component attachment) that is well established for the seismic-type response spectra but is not validated for the FRS due to the plasma transients. [ABSTRACT FROM AUTHOR]

Highlights • An exploratory risk analysis of ITER Cask & Plug Remote Handling System has been performed under a system engineering approach considering the CPRHS in various operational states with the associated loads. • A Functional Breakdown Structure was developed from the 4 main functions fulfilled by the CPRHS: to dock, to handle, to transport and to confine. • A Failure Mode Effects & Criticality Analysis has been made by identifying potential failures of basic functions to be fulfilled by CPRHS. • A sensitivity analysis on maintenance duration by means of statistics approach, the failure modes of CPRHS basic functions were introduced in the schedule Primavera Risk Analysis software. • CPRHS availability and operations time were then addressed while considering the impact of the failure modes on maintenance operations time and the benefit of mitigation actions. Abstract Since the ITER Casks will be not shielded, human access will be forbidden in trajectory zones and, in the event of failure of Cask and Plug Remote Handling System (CPRHS) functions, rescue operations will have to be remotely conducted. As potential failure modes could be severe in terms of time to repair, an inventory of CPRHS function outages and a risk analysis have been made at each stage of a diagnostic port plug maintenance process. CPRHS availability has been calculated in the framework of a RAMI analysis and the resulting time to perform the CPRHS operations for the maintenance of 2 Equatorial Port Plugs (EPP) has been confirmed by using a probabilistic Monte Carlo approach. [ABSTRACT FROM AUTHOR]

Abstract Four EC H&CD (Electron Cyclotron Heating and Current Drive) launchers will be installed into the upper ports #12, #13, #15 and #16 in ITER. Beside plasma heating their main purpose is to counteract plasma instabilities by injecting up to 20 MW of microwave power at a frequency of 170 GHz at dedicated positions into the plasma. The microwave power is generated in 24 Gyrotrons outside the Tokamak and transmitted into the dedicated ports by 8 waveguide lines per launcher. After passing a CVD (Chemical Vapor Deposition) diamond window and another section of circular waveguides, the microwaves are quasi-optically guided into the plasma by an adjusted set of mirrors. These mirrors and also the foremost segments of the waveguides are mounted into the Upper Port Plug, which is a massive steel structure capable to dissipate up to 800 kW heat and to withstand heavy mechanical loads induced mainly from plasma disruptions. This paper presents the most recent status of the design of the EC H&CD Upper Port Plug (UPP), featuring the construction of the supporting structure and the BSM (Blanket Shield Module) with its plasma facing First Wall, the PHTS (Primary Heat Transfer System) water cooling layout, the shielding components engineering, the microwave system integration and comprehensive manufacturing strategies. Also correlated computational analyses to prove the final design of the launcheŕs structural system are given. [ABSTRACT FROM AUTHOR]

The first mirrors of ITER diagnostic systems are the most vulnerable ones since they are directed to the plasma and are subjected to erosion and intensive impurity deposition. To keep their optical performance and maintainability, single crystalline molybdenum and rhodium have been considered as mirror materials, subject to intensive investigations including R&Ds and mirror cleaning studies. The paper presents specific design and structural issues of the first mirror of 300 mm in the size considered for the ITER core charge exchange recombination spectroscopy (cCXRS). Such a large mirror can be assumed as a generic one for solid middle (size) mirrors (up to 100 mm in diameter) and for composed large mirrors (diameter >150 mm) that are potentially made of an assembly of smaller pieces. The main design aspects like the first mirror material, the cooling capability of passively and actively cooled mirrors and the mechanical behavior of rhodium and molybdenum mirrors composed of single crystals pieces (i.e. plates and tiles) have been addressed and studied. The mirror design is supported by extensive structural and multi-field analysis. [ABSTRACT FROM AUTHOR]

To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill this requirement, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment. The ITER port plug test facility (PPTF) is composed of several test stands that can be used to test the port plugs whereas at the end of manufacturing (in a non-nuclear environment), or after refurbishment in the ITER hot cell facility. The PPTF provides the possibility to perform environmental (leak tightness, vacuum and thermo-hydraulic performances) and functional tests (radio frequency acceptance tests, behavior of the plugs’ steering mechanism and calibration of diagnostics) on upper and equatorial port plugs. The final design of the port plug test facility is described. The configuration of the standalone test stands and the integration in the hot cell facility are presented. [ABSTRACT FROM AUTHOR]

The welded lip seals form part of the torus primary vacuum boundary in between the port plugs and the vacuum vessel, and are classified as Protection Important Component. In order to refurbish the port plugs or the in-vessel components, port plugs have to be removed from the machine. The lip seal design must enable up to ten opening of the vacuum vessel during the life time operation of the ITER machine. Therefore proven, remote reliable cutting and re-welding are essential, as these operations need to be performed in the port cells in a nuclear environment, where human presence will be restricted. Moreover, the combination of size of the components to be welded (∼10 m long vacuum compatible thin welds) and the congested environment close to the core of the machine constraint the type and size of tools to be used. This paper describes the lip seal cutting and welding development programme performed at the VTT Technical Research Centre, Finland. Potential cutting and welding techniques are analyzed and compared. The development of the cutting, TIG and laser welding techniques on samples are presented. Effects of lip seal misalignments and optimization of the 2 welding processes are discussed. Finally, the manufacturing and test of the two 1.2 m × 1 m representative mock-ups are presented. The set-up and use of a robotic arm for the mock-up cutting and welding operations are also described. [ABSTRACT FROM AUTHOR]

The Diagnostic Port Plugs provide an infrastructure for integrating the diagnostics necessary to operate the machine, while withstanding the nuclear and mechanical loads. They have to provide neutron shielding in order to ensure accessibility to the Port Interspace (PI) region for maintenance operations. This report is about the study to mitigate the ITER Shutdown Dose Rate in Port Interspace. This is a strong design driver for the Equatorial Port Plugs (EPPs) requiring a neutron attenuation of 7 orders of magnitude while maintaining the weight limit. To achieve this challenging task, the design has evolved to an improved double labyrinth configuration with welded shims at the back of the EPP to reduce the gap with the Vacuum Vessel Port Extension. This proved to be very successful for this task. Neutron stoppers have been placed after the gaps necessary for the insertion of the Diagnostic Shielding Modules (DSMs) housed inside the EPP. Effort was also placed in the further development of the DSMs. The latest design relies on a stainless steel structure and lightweight efficient neutron absorbing materials inside. Studies of cross talk with lower ports and the influence of the streaming through the blanket area are presented to give the status of the radiation environment in the Port Interspace region and their contribution to the Shutdown Dose Rate (SDR). [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The Diagnostic Generic Equatorial Port Plug (GEPP) is designed to be common to all equatorial port-based diagnostic systems. It is designed to survive throughout the lifetime of ITER for 20 years, 30,000 discharges, and 3000 disruptions. The EPP structure dimensions (without Diagnostic First Walls and Diagnostic Shield Modules) are L2.9 × W1.9 × H2.4 m 3 . The length of the fully integrated EPP is 3174 mm. The weight of the EPP structure is about 15 t, whereas the total weight of the integrated EPP may be up to 45 t. The EPP structure provides a flexible platform for a variety of diagnostics. The Diagnostic Shield Module assemblies, or drawers, allow a modular approach with respect to diagnostic integration and maintenance. In the nuclear phase of ITER operations, they will be remotely inserted into the EPP structure in the Hot Cell Facility. The port plug structure must also contribute to the nuclear shielding, or plugging, of the port and further contain circulated water to allow cooling during operation and heating during bake-out. The Final Design of the GEPP has been successfully passed in late 2013 and is now heading toward manufacturing. The final design of the GEPP includes interfaces, manufacturing, R&D, operation and maintenance, load cases and analysis of failure modes. [ABSTRACT FROM AUTHOR]

At first a detailed fast shutter design was finalized for the ITER core charge exchange recombination spectroscopy (CXRS) diagnostic. The shutter has approximately 70 kg of mass and a length of 2.1 m. It operates in fractions of a second (0.7 s) protecting critical optical components against degradation and providing means of calibration for the optical system. The shutter structure is driven by a bidirectional frictionless helium actuator, with forces and axial strokes of 3.4 kN and 2 mm respectively. The shutter structure consists of: (a) two blades made of CuCrZr and stainless steel, calibration surfaces (currently Al 2 O 3 ) on the top and on the bottom a protective TZM (Mo–0.5Ti–0.08Zr) screens, (b) two arms interconnected that form one cooling circuit including the blades, (c) a bumper system to limit the arms movement, and (d) a support. A description of these components and their functions are given in this paper, followed by some issues, and their corresponding solutions or ongoing investigations, encountered during the design work. Detailed manufacturing drawings have been developed as the deliverable final product of this design stage, and are used in the prototyping phase which includes testing, numerical benchmarking, and validation of the shutter concept. [ABSTRACT FROM AUTHOR]

The ITER core charge exchange recombination spectroscopy diagnostics (cCXRS) occupies the vacuum vessel upper port #3 and includes, in its generic version, the following in-vessel components: an optical mirror system, a shutter, the diagnostic first wall and the neutron shielding block. The most vulnerable diagnostic mirror is obviously the first one (M1) directly observing the plasma. The M1 reference option is made of a single crystalline molybdenum (ScMo). The paper indicates major aspects influencing the first mirror design and identifies the most reasonable and reliable concept for cCXRS M1 at present. The applicability of the option presented is determined by many reasons, and especially, by the ITER generic upper port plug and its customization flexibility. The largest dimension of the mirror polished face is ∼300 mm. Such large ScMo workpieces are currently not available on the market. The mirror should be designed as an assembly of several ScMo pieces joined together. The M1 design is supported by multifield thermal, electromagnetic and structural analyses. The performed study confirms the feasibility of the proposed solutions. At the same time, the paper indicates numerous technological issues of the M1 unit to be solved in future. [ABSTRACT FROM AUTHOR]

Abstract: To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill these requirements, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment. The ITER port plug test facility (PPTF) provides the possibility to test upper and equatorial port plugs before installation on the machine. The port plug test facility is composed of several test stands. These test stands are first used in the domestic agencies and on the ITER Organization site to test the port plugs at the end of manufacturing. Two of these stands are installed later in the ITER hot cell facility to test the port plugs after refurbishment. The port plugs to be tested are the Ion Cyclotron (IC) heating and current drive antennas, Electron Cyclotron (EC) heating and current drive launchers, diagnostics and test blanket modules port plugs. Test stands shall be capable to perform environmental and functional tests. The test stands are composed of one vacuum tank (3.3m in diameter, 5.6m long) and the associated heating, vacuum and control systems. The vacuum tank shall achieve an ultimate pressure of 1×10−5 Pa at 100°C containing a port plug. The heating system shall provide water at 240°C and 4.4MPa to heat up the port plugs. Openings are provided on the back of the vacuum tank to insert probes for the functional tests. This paper describes the tests to be performed on the port plugs and the conceptual design of the port plug test facility. The configuration of the standalone test stands and the integration in the hot cell facility are presented. [Copyright &y& Elsevier]

Abstract: To counteract plasma instabilities, Electron Cyclotron Launchers will be installed in four of the ITER Upper Ports. The structural system of an EC Upper Launcher accommodates the MM-wave-components and has to meet strong demands on alignment, removal of nuclear heat loads, mechanical strength and nuclear shielding. The EC Upper Launcher has successfully undergone the Preliminary Design Review in 2009 and is now in the final design phase. Nuclear heat loads from 0.1W/cm3 up to 0.8W/cm3 will affect the front area of the launcher main frame. To guarantee save and homogenous removal of those heat loads, the front part of the launcher main frame is designed as a double wall steel-casing with cooling channels inside the shell structure. To finalize the design of this double wall component, the main emphasis is now to define the cooling channels geometry and to identify the optimum manufacturing route to assure adequate flow of coolant and sufficient mechanical strength in compliance with required dimension tolerances and quality of the welds. Several manufacturing options have been investigated and were evaluated by computational analysis and fabrication of pre-prototypes. To come to a final design, the most promising route will be chosen to manufacture a full-size mock-up of the double wall main frame. It will be tested at the KIT Launcher Handling Test facility to check the compliance with the design goals related to geometrical accuracy and thermo-hydraulic characteristics. This paper describes the design and the manufacturing routes of the prototypic double wall main frame. [Copyright &y& Elsevier]

Abstract: The radiation dose rate in ITER maintenance areas is one of the important safety-related quantities. Until recently it was assumed that acceptable values (i.e. <100μSv/h at a time 106 s after shut-down) could be reached. However, with the detailed design progressing it becomes evident that the presence of gaps in the design will lead to unacceptable high dose rates in important locations. In this paper an evaluation of the dose rate in the equatorial port interspace is presented. The current dose rate seems to exceed the ITER requirement by a factor of 10. Suggestions are presented to improve the situation. [Copyright &y& Elsevier]

Abstract: The Equatorial Visible/InfraRed Diagnostic is one of the key diagnostics of ITER, due to its crucial role for the machine protection, through the determination of plasma facing components temperatures (infrared range) and the first wall survey (visible range). Due to its location (close to the plasma) and its complexity (two types of wavelength must be transferred from the plasma to the detectors), the development of this diagnostic needs to address a lot of non-trivial issues. Significant technical advancements have been performed since a few years. This paper focuses on the recent progress made on the design and integration of the diagnostic. After a short description of the system, this paper shows the changes proposed for its integration in the port plugs where it is installed and the improvements brought to its design. It also presents the first results of a neutronic analysis which aims at optimizing the geometry of the dog leg mirrors forming a labyrinth for the neutrons in the port plug. [Copyright &y& Elsevier]

Abstract: For the optical system of the core CXRS port plug several designs were developed in the last years. A survey is given of the three most relevant designs. The first design is based on an off-axis ellipsoid mirror as main imaging component. This design has been the basis of opto-mechanical design as described in previous publications . This design is presented and assessed with regard to the optical properties, the sensitivity to first mirror degradation, the possibility to include a tube to allow replacement of the first mirror (as a possible mitigation of the first mirror degradation) and the manufacturability. Two alternatives are presented that alleviate the issues connected to the optical design. These alternatives may improve both lifetime and manufacturability of the core CXRS port plug. [Copyright &y& Elsevier]

Abstract: New optical configurations for the ITER core CXRS system offer the possibility of longer ducts between the first mirror and the plasma. This has led to a renewed optimization of the availability, using a simple model of the degradation of the first mirror that starts with the conditions of (a) the required measurement performance and (b) the geometry of the port plug. It is found that for a fully passive system the design should strive for the longest duct length possible. Given known data, this will result in a diagnostic lifetime still substantially shorter than ITER lifetime. When an option of cleaning the first mirror is introduced (assuming this is a feasible option) the optimum is less straightforward, because the lifetime of the second mirror then also becomes important. The optimum then depends on the ratio between the cleaning interval and the ITER lifetime. Options are presented for various sets of assumptions. Finally practical limitations of supporting subsystems (cleaning system, shutter, calibration system) may influence the final design. Examples of such limitations with their impact are presented. [Copyright &y& Elsevier]

This paper describes the preliminary engineering investigation of the optical labyrinth and the first and second mirror mounts of the ITER core LIDAR diagnostics. The optical labyrinth is in a very sensitive location being close and open to the plasma. Therefore, it has been examined carefully to be able to provide different solutions satisfying the needs of the diagnostics. The present conceptual design considered the covering of the light path, the integration of the labyrinth into the port plug, as well as the shape, color, and material of the labyrinth inner surface. It was absolutely necessary, since the undesirable phenomena of scattered light and sputtered particles reaching the mirroring surface must be handled. Nevertheless, cooling/baking has also been considered, and appropriate solutions have been presented. Regarding the mirror mounts, concepts have been created. It has been identified that the mirrors must be conditioned at a certain temperature; therefore, cooling/baking options were developed. The cooling/baking and, particularly, the neutron absorption functions were detected to be essential for the extension tubes. The development basically focused on the different installation methods of the neutron-shielding plates, which also allow the tubes to be kept in the required temperature range. The result of the development was that a water-steel mixture varied through the length of the extension tubes could provide the maximum neutron shielding. This solution could satisfy the very strict six-orders-of-magnitude neutron absorption required to maintain the postoperation safety. The conceptual engineering design has been supported by the preliminary thermal analysis of the mirror extension tubes. [ABSTRACT FROM AUTHOR]

Abstract: The exploitation of ITER tokamak will require diagnostics for machine protection, inputs to plasma control systems, evaluation and analysis of plasma parameters and performances. The equatorial visible/infrared wide angle viewing system and the radial neutron camera are the two main diagnostics of Procurement Package 11 (PP11), one of the diagnostic procurements under the responsibility of Europe, which also contains Equatorial Port Plug 1 and seven other diagnostics supplied by Europe or other ITER partners and integrated in the same Port Plug. Significant progress has recently been made in the development of the equatorial visible/infrared wide angle viewing system and the radial neutron camera. This paper gives an overview of the major technical achievements on these two diagnostics and points out the urgent needed R&D activities. [Copyright &y& Elsevier]

Abstract: Most of ITER diagnostic systems are integrated in port plugs, which are water cooled stainless steel structures inserted into the vacuum vessel ports. The port plug must provide basic functions such as neutron and gamma shielding, supporting the first wall armour (BSM), closing the vacuum vessel ports, while supporting the diagnostic equipments. ITER diagnostic port plug must resist a severe environment like high temperature due to neutron interaction with the structures and high electromechanical loading during disruptions events. CEA has contributed to the design and integration tasks in the frame of the representative equatorial port plug EQ#01, in particular on the engineering, structural and thermal finite element analysis. These detailed analyses have highlighted some design issues which were worked out through different solutions. This paper contains a description of the engineering activities performed such as: [-] The static mechanical calculations of the top plate closure system under disruption load. [-] The static mechanical calculations of the BSM attachment to the port plug. These two first studies led to design changes proposals which significantly improved the behaviour of the structures but also showed that the safety margin with respect to design limits is quite low. [-] The design of a Diagnostic Shield Module (DSM) integrated inside the port plug and a proposition of attachment scheme, with respect to disruption loads. The manufacturing of the DSM has been taken into account, as well as diagnostic integration inside the structure and maintenance aspects. [-] The thermal assessment of the port plug under neutronic load during normal operation, with the optimization of the cooling system. The maximum temperature calculated in normal operation has been reduced from 900°C to less than 400°C in the front plate; and the cooling arrangement at the back of the port plug has been simplified without important temperature increase. [Copyright &y& Elsevier]

Abstract: The European test blanket module (EU-TBM), first prototype of the breeding blanket concepts under development for the future DEMO power plant to produce the tritium, will be developed to be tested in three equatorial ports of ITER dedicated to this. The CEA Cadarache under the contract of Association EURATOM/CEA and in close relation with Association EURATOM/HAS works on the integration of the EU-TBM inside ITER tokamak. The installation of the TBM into the vacuum vessel is made with the help of a port plug, constituted with two components: the Shield module and the Port-Plug frame. The Shield module provides the neutron shielding inside the Port-Plug frame, which maintains in cantilever position the TBM and its shield module and closes the vacuum vessel port. This paper will describe the EU-TBM design and integration activities on the cooled shield module and on its interface with the TBM component. A particular attention, in term of thermal and mechanical studies, is dedicated to the design of the shield and test blanket module attachment, and also to the shield design and its internal cooling system. [Copyright &y& Elsevier]

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

Abstract: The integration system for installation of the European Helium Cooled Pebble Bed Test Blanket Module (EU HCPB TBM) in ITER being developed in the frame of the FZK Fusion programme uses three main interfaces to connect the TBM to its sub-systems. This paper describes how the occupation of the ITER hot cell for TBM installation into the port plug can be limited to the essential taking into account the update of the port plug design. The new design allows the dismantling of the radiation shield from the port plug frame. If a new shield is used for each new TBM the operations for connection of interface I between the TBM and shield can be done outside of the ITER hot cell as hands on operation. [Copyright &y& Elsevier]

Abstract: The international thermonuclear experimental reactor (ITER) vacuum vessel (VV) upper port has a trapezoidal/rectangular cross-section and consists of a port stub, a stub extension and a port extension with a connecting duct. To investigate the structural integrity and to increase the structural reliability of the VV and ports, the structural analyses of the upper port structure have been performed. Two advanced designs from the ITER international team have been deeply analyzed. To verify the strength of the reinforcing ribs for the connecting duct and of the fastening/sealing units, the local analyses utilizing the sub-modeling technique have been performed. The upper port flange based on the original design could withstand design loads, but the flange surface could be detached under the design condition. The advanced design using large superbolts is highly acceptable in the structural design point of view. The deflection of the plug for the other advanced design with a removable flange is relatively large since the stiffness of a port and a plug is reduced by the nut groove and the removable flange. [Copyright &y& Elsevier]

Abstract: The integration of the HCPB Test Blanket Module (TBM) systems into ITER has been investigated with the aim to check the compatibility between the considered systems and ITER. In particular the requirement of a simple and quick TBM replacement is considered in this paper. It presents the systems and operations in hot cell and port cell for the integration and exchange of the four different TBMs in succession during operation into the ITER-circuits. This concept shows that TBM integration using three interfaces is a simple and quick way that is compatible with ITER environment and systems. [Copyright &y& Elsevier]

Abstract: In-vessel diagnostic equipment in ITER integrated in six equatorial and 12 upper ports, 16 divertor cassettes and five lower ports is designed to be removed in modules and then repaired, tested and commissioned in the same location at the ITER hot cell. The repair requirements and tests on these components are described along with design features that facilitate repair. The testing establishes the repair strategy, qualifies the refurbishment work and finally checks the mechanical and diagnostic function before the return of the modules. At the hot cell, a dummy port is provided for tests of mechanical and vacuum integrity as well as commissioning of the diagnostic equipment. The scope of the hot cell maintenance and commissioning activities is summarised and an overview of the integration of the diagnostic equipment is given. [Copyright &y& Elsevier]

Abstract: Two ITER limiters, located in two diametrically opposite equatorial ports, perform the specific functions of defining the plasma boundary and protecting other regions of the first wall and antennas from direct contact with the plasma during start-up and ramp-down. The considerable thermal load in these phases (H.D. Pacher, G.W. Pacher, ITER Start-up Limiter Power Loads, INRS Energie et Matériaux, Varennes, Qué., Canada, Report No. NRG-EM-00-676, February, 2001), and the expected high loads during abnormal events, call for a design using a special movable and separable part of the limiter assembly, comprising high-heat–flux components (the limiter module) that require very accurate alignment to the field contour in order to reduce the heat load peaking. Additionally, the design should allow relatively simple remote replacement and refurbishment of the module. Furthermore, uncertainties on the optimum plasma configuration require that the plasma-facing surface can be adjusted to match different plasma scenarios. An engineering design of the limiter port plug with a remotely controlled mechanical supporting and alignment system has been developed to meet all the requirements, providing safe and reliable operation with good margins. The conceptual design of 1998 with its engineering design of 2001 is considered as the current reference for ITER. The paper describes the design and layout of the system and its main components, the principles of operating and assembly. [Copyright &y& Elsevier]

Abstract: Engineering analyses have been performed for the representative diagnostic upper port plug of ITER. Maintenance and integration design have been also carried out for the diagnostic components to be installed in the upper port plug. From the electromagnetic and structural analyses, it has come up an important problem to suppress the displacement of the upper port plug rather than to reduce the produced stress. Reducing the electro-magnetic (EM) force will help to decrease the severity of potential displacement. Maximum displacement of the port plug decreases with increase in the number of slits in a manner that the displacement would seem to be less than the design tolerance. A proposed low body roller and inner frame may enhance maintenance and integration. These studies and designs have established the design basis for the diagnostic upper port plug. [Copyright &y& Elsevier]

• A first pre-conceptual design concept for an DEMO EC launcher is presented. • It includes all relevant components on a pre-conceptual basis. • General design requirements are taken into account. For the EU DEMO Tokamak, Electron Cyclotron (EC) launching systems for plasma heating and stabilization are under development. Various concepts for the optical system are currently studied of which the Mid Steering Antenna (MSA) with a steering mirror at a recessed position behind the breeding blanket (BB), the Open Ended Waveguide (OEWG) concept with quasi-optical beam propagation by fixed mirrors only and the Remote Steering Antenna (RSA), currently seen as back-up solution, are the basic ones. In addition, hybrid solutions, which means a port plug with different launcher concepts, are taken into consideration. In parallel, design drafts for a generic equatorial port plug are sketched with the aim to provide a versatile structural system, which allows customized installation of the potential optical systems. This paper presents a pre-conceptual hybrid design of an equatorial EU DEMO EC port plug based on MSA for plasma stabilization and OEWG for plasma heating, taking into account the exact port position with respect to the toroidal field coil, mechanical integrity, heat dissipation, neutronic shielding requirements, design integration and maintenance concepts. [ABSTRACT FROM AUTHOR]