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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]

• 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]