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ENEA and Ansaldo Nucleare S.p.A. are currently working on the fabrication of full scale prototypes of inner vertical target component for ITER (International Thermonuclear Experimental Reactor). The requirements is to manufacture a component by means of a qualified technology which uses armour and cooling pipe materials defined by ITER organization specification. A dedicated new Hot Radial Pressing facility was designed and finally installed in ENEA for the fabrication of the full scale Plasma Facing Units of the inner vertical target component prototype. HRP technology was widely tested with small mock-ups and medium qualification prototypes but its applicability to fabricate full scale units has to be anyway demonstrated and, therefore, their joint quality and also dimensions and shape has to be properly checked. The compliance of the joint quality with the acceptance criteria is then performed by ultrasonic water gap technique, using the qualified ENEA testing equipment. Dimensional check is performed as well by analyzing acquired data points compared to the reference drawings and 3D model. [ABSTRACT FROM AUTHOR]

The plasma in the tokamak JT-60SA is initiated mainly by operating the four Switching Network Units (SNUs) inserted in the four modules of the Central Solenoid. In the framework of the “Broader Approach” agreement, these four SNUs were procured by the European Union. This paper presents the final assembly of the SNU system and the main tests performed in the last years before shipping it to Japan. The experimental results demonstrated the capabilities of the SNUs to comply with the JT-60SA requirements, even with better performances. [ABSTRACT FROM AUTHOR]

The F4E, Amec Foster Wheeler and INL comprehensive methodology for fusion breeding blanket accident analysis starts with the selection of reference accident scenarios, the development of detailed accident analysis specifications and the assessment of analysis codes. Models of each TBS are then constructed using the selected codes (MELCOR and RELAP5-3D) and modelling approaches. The models are qualified according to a test matrix including comparison with TBM finite element design analyses, code-to-code comparisons for both TBS normal operation and transient cases, and sensitivity studies for accident scenarios. The qualification test cases that are executed gradually move from models of separate systems to complete TBS models, and from the simulation of steady-state and normal plasma pulse operation to consideration of power excursions, operational transients and accident events. Finally, both of the qualified models are used to analyse a selected accident scenario (a 32 h loss-of-offsite power) together with sensitivity studies dedicated to the evaluation of uncertainties. This step completes the qualification process. The results obtained in the qualification of the EU HCLL TBS models are reported in the paper. [ABSTRACT FROM AUTHOR]

The realization of the 19.6 m 2 highly heat loaded part of the divertor of Wendelstein 7-X (W7-X) requires the installation of 890 target elements. A target element is made of a CuCrZr copper alloy heat sink armored with carbon fibre reinforced carbon CFC NB31 tiles. The industrial production of the target elements by the Austrian company PLANSEE SE needed 5 years. The successful delivery of a total of 973 target elements (armored with ∼16,200 tiles) was based on an efficient quality assurance throughout the production by the manufacturer: 44 different quality examinations for 82 manufacturing steps. It was followed by an intensive quality assessment by IPP: visual inspections, dynamic pressure tests, He-leak testing in vacuum oven, high heat flux testing. The quality of the delivered elements, and in particular the reliability of the bonding between CFC tiles and heat sink, was confirmed by high heat flux testing on the basis of a statistical approach. The essential recovery of CuCrZr properties after ageing processes was confirmed by the measurement of hardness and electrical conductivity. 104 target elements (10.7%) were accepted after repair. [ABSTRACT FROM AUTHOR]

ASDEX Upgrade (AUG) is the only tokamak in Europe to have low activation ferritic steel as the inner vessel wall facing component. Together with the massive tungsten tiles in the lower divertor, AUG is the tokamak with the closest DEMO wall. The project is a first step towards the extensive use of ferritic steel in future fusion reactors. For example, the test blanket module of ITER is planned to have a ferritic steel wall and thick tungsten tiles as a plasma facing component. The ‘ad hoc’ ferritic steel built with low activation capability is known as Eurofer. As the low activation property is not a requirement for AUG, the material selected for the project is the martensitic steel P92 which is the most similar material to Eurofer from a magnetic point of view. The purpose of the project is to improve understanding of the magnetic perturbation of the ferritic steel both on the plasma and magnetic probes, evaluating and controlling these effects. Additionally, the effect of the additional forces on the supporting structure has been addressed. Bearing this in mind, in 2013 a step wise program has been started and part of the W coated graphite tiles in the region of the inner column were replaced by steel tiles [1]. The first campaign did not suffer any particular issue related to the new material. According to the calculations, the plasma was almost unperturbed, thanks also to the toroidal symmetry of the tiles inside the vessel, and the magnetic probe measurements were properly corrected [2]. Inspection of the machine pointed out some hardware problems. The graphite tiles adjacent to the steel tiles were damaged. The graphite tiles had broken edges in 5 from 64 positions and notches in many others. The coating of the graphite and steel tiles, made of tungsten and TiO respectively, was damaged. At first glance it was clear that the steel tiles were moving but it was definitely unexpected. In understanding the process, the location of the damage was the crucial hint. In fact all failures were located at the boundary between 2 vacuum vessel octants. To justify this failure mode inside the vessel, a hypothesis (about current flowing in the heat shield supporting structure) was made and FEM analyses were carried out in this direction. With extreme caution, in 2015 just one additional row of steel tiles was added together with diagnostics that confirmed the hypothesis. Now that a clear understanding of the problem has been reached, the project to add further rows of steel tiles can be continued. For the next campaign it is planned to replace all the tiles in the middle region of the heat shield together with stiffening and modification of the supporting structure. In this paper the learning process from the damage of the tiles and its causes, from the FEM analysis results to the data diagnostics will be reported. The future plans for steel tiles in AUG will be discussed. [ABSTRACT FROM AUTHOR]

In the framework of the EUROfusion DEMO project, studies are conducted in several European institutions for designing the tokamak magnet systems. In order to generate the high magnetic fields required for the plasma confinement and control, the reactor should be equipped with superconducting magnets, the reference design being based on Cable-In-Conduit Conductors (CICC) cooled at cryogenic temperatures by forced circulation of supercritical helium. In order to propose a toroidal field (TF) winding pack (WP) design compatible with DEMO requirements, CEA has developed several tools addressing the different areas related to magnet dimensioning. An accurate calculation of magnetic field along the conductors is provided by the TRAPS code, and conductor design is performed by using an integrated macroscopic home design code based on simplified models accounting for superconducting properties, mechanics and thermal. This multi-physic tool gives a realistic but not assessed design. Indeed it is based on an assumed operating temperature that must be validated with an elaborate code, since it is linked with temperature margin design criterion (1.5 K). A dedicated modelling tool was developed by coupling the THEA code for 1D thermo-hydraulics in cables and the Cast3M code for 2D transverse thermal diffusion in a limited number of coil cross-sections, enhancing the accuracy of the outputs as being a quasi-3D approach. This tool allows a better assessment of the flux exchange between WP and casing, and the modelling of inter-turn and inter-pancake thermal coupling. The coupling methodology is described, as well as its validation on the simulation of a heat exchanger. A calculation was performed on the CEA proposal for DEMO TF coil in a burn (steady state) scenario, and finally providing a realistic assessment of the temperature margin. [ABSTRACT FROM AUTHOR]

Unsteady loading rates can affect the proper operation of a biofilter. In addition, biofiltration is known to be less effective when dealing with poorly soluble substances, like some volatile organic compounds (VOCs). A non-thermal plasma (NTP) produced with a dielectric barrier discharge (DBD) was studied as an option to pre-treat an air stream contaminated by a mixture of VOCs. Therefore, the plasma reactor was operated upstream of a laboratory-scale biofilter. Air admixed with toluene, n -heptane, p -xylene, ethylbenzene and benzene with average concentrations of 95.6, 49.4, 60.8, 47.3 and 36.6 ppm, respectively, was used as a model polluted gas, as these contaminants represent the air stripped by an oil-refinery wastewater treatment plant. Peaks of loading rates at the inlet of the biofilter were simulated by the increase of the flow rates of VOCs. The operation of NTP, with specific energy densities between 92 J L −1 and 256 J L −1 allowed reducing the VOC concentrations down to the level of optimal biofilter operation. In addition, non-water soluble VOCs were converted to more soluble compounds by the plasma treatment. In this first attempt to investigate the synergies between NTP and biofiltration, NTP reveals as a promising option to pre-treat effluents upstream of biofilters for optimized operation. [ABSTRACT FROM AUTHOR]

The article is devoted to the problem of creating a system of automated control with boron regulation for reactor WWER-1000 series. Using the boron regulation to control WWER-1000 allows to extend its maximum output operation period, ensuring the economic efficiency of the power unit, as well as to maintain the reactor facility within relevant safety limits that prevents from emergencies occurrence and development. The results of this problem solution, related to the process simulation, optimization and prediction, were used at further development of computer-integrated control system increasing the efficiency of decisions, taken by operational staff at reactor control. [ABSTRACT FROM AUTHOR]

Highlights: [•] Various plasma reactors available for water remediation are addressed. [•] Combination of plasma with catalysts for pollutant removal is reviewed. [•] Some factors in pulsed electrical discharge are systematically discussed. [•] Further researches and challenges for plasma technology are presented. [Copyright &y& Elsevier]