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[en] Highlights: ► Thermo-hydraulic transient performance evaluation and optimization of the GEPP structure cooling/heating system under neutronic heating and baking conditions. ► The optimization of the GEPP box structure's cooling system includes positioning and minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions. - Abstract: The port-based ITER diagnostic systems are housed primarily in two locations, the equatorial and upper port plugs. The port plug structure provides confinement function, maintains ultra-high vacuum quality and the first confinement barrier for radioactive materials at the ports. The port plug structure design, from the ITER International Organisation (IO), is cooled and heated by pressurized water which flows through a series of gun-drilled water channels and water pipes. The cooling function is required to remove nuclear heating due to radiation during operation of ITER, while the heating function is intended to heat up uniformly the machine during baking condition. The work presented provides coupled thermo-hydraulic analysis and optimization of a Generic Equatorial Port Plug (GEPP) structure cooling and heating system. The optimization performed includes positioning, minimization of number and size of gun drilled channels, complying with the flow and functional requirements during operating and baking conditions.
[en] Under construction for the stellarator project Wendestein 7-X is a neutral beam heating system based on RF driven positive ion sources. It is planned to start operation with 2 sources capable of injecting 5 MW of heating power in deuterium. This paper gives the current status and future plans of the construction of the injector boxes and subsequent installation in the experimental hall. The fruitful collaboration with the National Center for Nuclear Research in Swierk, Poland is also detailed. Lastly, results from an initial study on fast ions in Wendelstein 7-X will be given
[en] Highlights: In the frame of the conceptual design phase of the EU DEMO an effort is made here: • To define the interface requirements among systems to be integrated in the VV and the BB. • To propose the integration strategies for the auxiliary heating, diagnostic and fuelling systems into the VV and the BB and for the BB and divertor supporting structures. • To define a schedule for the in-vessel components integration design analyses. • To identify the 3D supporting tools. - Abstract: In the EU DEMO design (Romanelli, 2012; Federici et al., 2014), due to the large number of complex systems inside the tokamak vessel it is of vital importance to address the in-vessel integration at an early stage in the design process. In the EU DEMO design, after a first phase in which the different systems have been developed independently based on the defined baseline DEMO configuration, an effort has been made to define the interface requirements and to propose the strategies for the mechanical integration of the auxiliary heating and fuelling systems into the Vacuum Vessel and the Breeding Blanket. This work presents the options studied, the engineering solutions proposed, and the issues highlighted for the mechanical in-vessel integration of the DEMO fuelling lines, auxiliaries heating systems, and diagnostics.
[en] The ion cyclotron resonance frequency (ICRF) heating system on JET is currently being upgraded to validate new matching concepts in view of coupling ICRF power to ITER-like plasmas and to further increase the total additional heating power on JET. The present paper reports on first testbed results from the new JET ITER-like antenna as well as on the first use of the newly installed hybrid couplers between two of the existing A2 antennas. Several other on-going improvements, such as improved trip management system, external conjugate-T matching circuit and arc detection systems are also discussed
[en] Power switching in RF heating systems is a delicate function as it is often linked to high power tube protection. In most RF systems, the end stage power tube is fed by a high voltage power supply (HVPS), which connection to the tube has to be interrupted in case of arc suspicion. The amount of energy that is allowable to be dissipated in the arc is in the range of 10-50 J, to limit the degradations observed on the tube structures. The protection function is usually performed by a crowbar. Furthermore, the HVPS is often shared by several power tubes, and the loss of all the power from the group of tubes is to be avoided to minimize the perturbation on the plasma experiment. A description of a 40 kV thyristor based crowbar and a 100 kV, 25 A MOSFET switch is given, as well as the contours of the existing components for high power switching applications. By combining small components, such as thyristors or MOSFET, in matrix, highly compact and reliable units have been built and implemented in Tore Supra RF systems
[en] To counteract plasma instabilities, electron cyclotron launchers will be installed in four of the ITER upper ports. An EC launcher consists of a structural system which accommodates the MM-wave-components and has to meet demands on precise alignment, sufficient removal of nuclear heat loads, mechanical strength and proper nuclear shielding. The structural system consists of the Blanket Shield Module (BSM) and the Mainframe. Depending on the expected heat loads, the launcher components are designed as double-wall or single-wall elements, connected by massive flanges. Double-wall segments feature narrow cooling gaps with stiffening ribs between stainless steel shells. Single-wall segments consist of welded stainless steel plates with substantial material thickness. To investigate industrial manufacturing routes, characteristic sections of the BSM and the Mainframe were addressed in detail. To identify the optimum manufacturing strategy for double-wall components, three different concepts, namely HIP (Hot Isostatic Pressing), brazing and machining were studied and a total of four mock-ups of double-wall components were produced. Also for single-wall components appropriate manufacturing routes were investigated, optimum production parameters were determined and typical segments were manufactured. These prototypes are under study at the FZK Launcher Handling Test facility (LHT) where various ITER operating conditions can be simulated. Analyses and test series related to feasibility, prevention of residual stresses, contour accuracy, thermo-hydraulic behavior and also economical aspects were performed.
[en] The ICRF system at the ASDEX Upgrade tokamak is in operation since May 1992. Following some modifications of which the major one was the installation of 3 dB couplers it has become a reliable additional heating system. The maximum power coupled into the plasma has been raised up to 7.2 MW (90% of the installed RF power) for short pulses and up to 6.2 MW for pulses several second long with energy of up to 29 MJ. A power of 5 MW is delivered on a regular basis to replace two NBI sources
[en] The development of a Pd-based membrane reactor to be applied in processes for tritium removal from various gaseous streams of tokamak systems has been carried out. In particular, the membrane reactor has been designed for decontaminating soft housekeeping wastes of JET. This membrane reactor consists of Pd-Ag permeator tube fixed in a finger-like mode into a stainless steel shell. The feed stream (gases to be detritiated) is fed inside the membrane lumen where the isotopic exchange takes place on to a catalyst bed while pure hydrogen (protium) is sent in countercurrent mode in the shell side. The feed stream consists of 200 Ncm3 min-1 of helium with 10% of tritiated water (tritium content 1.11 x 108 Bq h-1). The membrane reactor design has been based on a simplified calculation model which takes into consideration the very low tritium content of the gas to be processed and the complete oxidation of the tritiated species in the feed stream. The model considers a tubular Pd-Ag membrane divided into finite elements where the mass balances are performed according to both the thermodynamic equilibrium reactions and permeation rates through the membrane of the hydrogen isotopes. The reactor model has permitted to verify that a Pd-Ag commercial tube of diameter 10 mm, length 500 mm and wall thickness 0.150 mm is capable to attain a decontamination factor larger than 10. A new mechanical design of the Pd membrane reactor has been also developed: especially, harmful mechanical stresses of the long permeator tube consequent to the hydrogenation and thermal cycling has been avoided. Furthermore, an innovative effective heating system of the membrane has been also applied.
[en] Highlights: • Prototype ITER neutral beam injector to study the beam source behavior requires beam characterization. • Among several types of diagnostic, a calorimeter made of fiber carbon composite (CFC) tiles will be used. • The CFC thermal behaviour is study by laser beam and will be characterize in particel beam. • Experiments and simulations results are discussed. • Designs for tests in particle beams are presented. -- Abstract: For ITER operations, additional heating systems are required. One of these systems is the neutral beam injector (NBI). The SPIDER experiment, a small-scale NBI, is going to be built with the aim to optimize the beam source. For this reason it is provided with several diagnostics, among which the Short-Time Retractable Instrumented Kalorimeter Experiment (STRIKE). In this contribution, a characterization of the Carbon Fiber Composite (CFC) tiles, which are the main component of the diagnostic, is presented. Such analyses include tests with a power laser, exposure to particle beams and thermal stress tests. The results are discussed, which will drive the definition of the acceptance tests of the final supply of CFC tiles
[en] Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different