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One of the main challenges for the construction of DEMO, the first fusion demonstration reactor, is the power exhaust issue. To deal with it, EUROfusion has undertaken the construction of DTT (Divertor Tokamak Test) facility which will be held in Frascati, Italy. It aims to test various divertor models which can be integrated and used under different plasma confinement configurations. The current design of DTT embed 54 divertor cassettes which must be anchored to the vacuum vessel using suitable fixation systems. This paper deals with the initial monotonic load assessment of the fixation system of the Fixation System which must ensure compliance with multiple design requirements. The Remote Handling compatibility is one of the most demanding, since it includes a preloading phase of the divertor necessary to mitigate the shaking of the cassette due to dynamic forces. Moreover, the system must withstand the conspicuous loads acting on the Divertor due to the full or partial loss of plasma confinement (Disruptive Events). Structural verifications through Finite Element Method (FEM) have been performed on the Divertor-Fixation System, considering a set of load combinations, with particular focus on the most demanding case related to Fast-transient Vertical Disruption Event-downward. [ABSTRACT FROM AUTHOR]

• The MITICA (Megavolt ITER injector and concept Advancement) experiment, currently in the installation and commissioning phase at the neutral beam test facility in Padua-Italy, will be equipped with a beam source (BS) expected to generate a 40A negative ion beam to the energy of 1MeV (values for D). It is composed of an ion source to generate a h or d plasma and of a multiple-Aperture multiple-grid (MAMuG) accelerator [1] , where ions extracted are accelerated in five consecutive stages of 200 kV each. • The engineering and fabrication of the beam source for MITICA BS started in october 2018, when a contract between fusion for energy (F4E) and alsymex (France) was signed. Presently, all the components and sub-assemblies have been manufactured, except for a few non-critical items. Assembly of the accelerator is well advanced and ion source is on-going. • A review of the fabrication of the main components and aspects of their assembly is reported in this paper, with a focus on the main achievements. The MITICA (Megavolt ITER Injector and Concept Advancement) experiment, currently in the installation and commissioning phase at the Neutral Beam Test Facility in Padua-Italy, will be equipped with a Beam Source (BS) expected to generate a 40A negative ion beam to the energy of 1MeV (values for D). It is composed of an ion source to generate a H or D plasma and of a multiple-Aperture multiple-grid (MAMuG) Accelerator [1] , where ions extracted are accelerated in five consecutive stages of 200 kV each. The engineering and fabrication of the beam source for MITICA BS started in October 2018, when a contract between Fusion for Energy (F4E) and Alsymex (France) was signed. Presently, all the components and sub-assemblies have been manufactured, except for a few non-critical items. Assembly of the accelerator is well advanced and ion source is on-going. A review of the fabrication of the main components and aspects of their assembly is reported in this paper, with a focus on the main achievements. [ABSTRACT FROM AUTHOR]

• Remote control of a pellet facility has been proved repeatable and reliable. • E3S platform has been successfully tested in remote control of a pellet injector. • Performance tests indicate that, for massive I/O data flow, E3S should be scaled-up. The four-barrel, two-stage gun Ignitor Pellet Injector (IPI) was developed in collaboration between ENEA and ORNL. The prototype injector is presently located at Oak Ridge (TN, USA), and is normally operated locally through a control and data acquisition system developed in LabVIEW. More recently, a remote-control system has been set up, based on RealVNC®, which allows to operate the IPI from a control room in Italy. Tools for data transfer and storage into ENEA ICT area have also been provided. A Staging, Storage and Sharing system, named E3S, developed using OwnCloud as architectural component, is used for file synchronization and sharing of the data acquired by the diagnostic systems. It provides a homogeneous platform able to store and share heterogeneous data produced by many data acquisition systems in large nuclear fusion experiments. This paper reports about the implementation of the IPI remote control, and presents the application of E3S to this specific case, allowing easy storage and sharing of experimental data onto a wide-area distributed file-system, as well as remote data access via web-services based on MDS+ tool, integrated with MySQL metadata. A performance analysis of the architectural components is also introduced. [ABSTRACT FROM AUTHOR]

The Divertor Tokamak Test (DTT) facility is in advanced design and construction phase at the ENEA laboratories in Frascati, Italy, to contribute to the study, the design and assessment of systems for the treatment of heat exhaust. The DTT Toroidal Field (TF) magnet system consists of 18 superconducting coils, fed in series by a single power supply (TFPS), which is a 24-pulses thyristor-based converter with an output DC current up to of 44 kA. In case of a quench in the superconductors, the magnetic energy stored in the coils is extracted by 3 Fast Discharge Units (FDUs) in series with 3 sectors of 6 TF coils. After the respective contracts were awarded, the TFPS and the FDUs are currently in the final design phase which is addressed in this paper. In particular, the analysis is focused on the most innovative aspects of the design, as the use of silicon carbide (SiC) varistors instead of discharge resistors and of a fully electronic and completely redundant high-current switch without any mechanical device. [ABSTRACT FROM AUTHOR]

In the framework of the Broader Approach Agreement for the construction of the JT-60SA tokamak, ENEA provides 9 of the 18 toroidal field (TF) coils of the JT-60SA magnet system. The 9 coils are being manufactured by ASG superconductors in Genoa under the supervision of ENEA in collaboration with the JT-60SA European home team. The manufacturing is composed of two main steps: one concerning winding pack assembly and impregnation, and the other devoted to the integration into the casing structure and associated final coil preparation. This paper describes the results of the validation activities set-up for the integration phase. Specifically, welding of casing components has been retained particularly critical for at least three reasons: (i) during welding the WP may be damaged by the intense heating; (ii) distortion caused by heating may determine incorrect coil geometry and then field errors; and (iii) flaws may reduce structural strength and then the overall lifetime of the machine. Similarly, final embedding has been demonstrated on a 1 m long mock-up of the coil. Main results and lessons learned are here described. [ABSTRACT FROM AUTHOR]

Abstract: The ITER Heating Neutral Beam injectors will be implemented in three steps: development of the ion source prototype, development of the full injector prototype, and, finally, construction of up to three ITER injectors. The first two steps will be carried out in the ITER neutral beam test facility under construction in Italy. The ion source prototype, referred to as SPIDER, which is currently in the development phase, is a complex experiment involving more than 20 plant units and operating with beam-on pulses lasting up to 1h. As for control and data acquisition it requires fast and slow control (cycle time around 0.1ms and 10ms, respectively), synchronization (10ns resolution), and data acquisition for about 1000 channels (analogue and images) with sampling frequencies up to tens of MS/s, data throughput up to 200MB/s, and data storage volume of up to tens of TB/year. The paper describes the architecture of the SPIDER control and data acquisition system, discussing the SPIDER requirements and the ITER CODAC interfaces and specifications for plant system instrumentation and control. [Copyright &y& Elsevier]

Abstract: PRIMA (Padova Research Injector Megavolt Accelerated) is the name of the ITER neutral beam facility currently under construction at the RFX Consortium located in Padua – Italy. PRIMA consists of two injectors which will test at the same time the ion source (SPIDER) and the whole system (MITICA). Both injectors accelerate negative deuterium and hydrogen ions with a maximum energy of 1MeV for MITICA and 100keV for SPIDER, with a maximum beam current of 40A. Accelerated ions are stopped on a CuCrZr alloy calorimeter. Operation in deuterium leads to production of an intense neutron flux through D–D and D–T reactions on the calorimeter panels which is likely to activate materials and components around the injectors, among which the corrosion products formed in the cooling loops; the radioactivity of the activated corrosion products (ACPs) may persist even after injector shutdown, representing a relevant radiation protection issue during repair and inspection activities. In the present paper the occupational radiation exposure (ORE) is evaluated. The total amount of radioactivity produced inside the cooling loops is calculated under different operative conditions and the radiation dose received by workers carrying out maintenance activities on the injector is also assessed for both experiments. [Copyright &y& Elsevier]

Abstract: The NIO1 (Negative Ion Optimization phase 1) experiment, whose construction has started in 2010 at Consorzio RFX in Padova in collaboration with INFN-LNL, is a radio frequency Hydrogen negative ion source designed to produce a beam current of 130mA with 60keV particle energy. The aim of the experiment is to provide a highly flexible system to test material properties of source components and to benchmark simulation codes for beam optics and plasma behavior. The paper focuses on a conceptual design for NIO1 power supply system on the basis of given operational and performance requirements. The requested ratings fall well within those of products available on the market and the design includes standard commercial solutions for the individual power supplies. The ion source operates at −60kV with respect to ground, leading to a number of design issues at system level and related to mutual connection between power supplies and load, layout and grounding. These aspects, along with the crucial point of ensuring protection to the arc prone beam source, are reviewed and discussed. The study leads to a scheme integrating high voltage power supplies, high current power supplies and a possible solution for passive protection components at the power supply output. [Copyright &y& Elsevier]

Abstract: In this paper we exploit the dynamic allocation scheme for input redundant control systems proposed in to achieve elongation control on FTU (Frascati Tokamak Upgrade). The scheme first serves as a means for regulating the current in the F coils. Then, due to the quasi-static relationship between the plasma elongation and the F coils current, elongation control is achieved by suitably generalizing the allocation scheme. Both simulation and experimental results are reported. [ABSTRACT FROM AUTHOR]

Abstract: The Neutral Beam Test Facility, which will be built in Padova, Italy, is aimed at developing the ITER heating neutral beam injector (HNB) and at testing and optimizing its operation up to nominal performance before installation on ITER. It requires the development of two independent experiments referred to as SPIDER (source for production of ions of deuterium extracted from Rf plasma) and MITICA (megavolt ITer injector & concept advancement). SPIDER will explore the full-size negative ion source for ITER, whereas MITICA will explore the full-size ITER neutral beam injector. Both experiments will be designed for long-pulse operation, up to 3600s, as ITER itself. MITICA includes three functional components: the heating neutral beam injector plant system (HNB), which is the device under test; the auxiliary plant system (AUX), which includes all equipment to operate the HNB in the test facility (e.g. the local electric grid to feed the HNB power supplies), and MITICA supervisory system that is an electronics/informatics infrastructure to operate the facility. The paper introduces the requirements for the control and data acquisition systems of the experiments and proposes a preliminary design for both systems. SPIDER, which is preparatory to MITICA and will be developed on a shorter time scale, has no constraints coming from ITER CODAC, whereas MITICA includes the ITER neutral beam injector and therefore must be fully compatible with ITER CODAC. [Copyright &y& Elsevier]

Abstract: In the framework of the construction of the Test Facility for the ITER Neutral Beam Injector to be installed at Consorzio RFX in Padova (Italy), a cryosorption cryopump, designed by FZK, will be employed to guarantee the appropriate pressure range for the D2or H2 pulsed beam. The cryopanels that constitute the cryopump will be indirectly cooled by means of supercritical helium at temperatures ranging from 4.5K to 6.5K and at a pressure of 0.5MPa. The thermal shields will be indirectly cooled by means of high pressure (1.5MPa) gas helium at temperatures from 65K to 90K. In order to provide the helium gas and the supercritical helium to the cryopump, a cryogenic plant has been further designed by RFX. The cryogenic plant consists of a Main Refrigerator to supply the supercritical helium, a Shield Refrigerator based on a Brayton cycle for the cooling of thermal shields, a Proximity Cryogenic System to provide the helium distribution and a Helium Recovery and Purification System to deliver pure helium to the refrigerators. The plant has been conceived to operate in several functioning modes according to the needs of the test facility: a cool-down operation, stand-by and NBI operation modes, a 100K regeneration mode, a warm up operation to ambient temperature, and a 470K regeneration mode. The full flexibility of the system has been pursued to assure its reliability and to chase unbroken operation. This paper presents the cryoplant design and the technical solutions adopted to optimize the system performances and its capability to implement all the required operative scenarios in a fully automatic mode. [Copyright &y& Elsevier]

Abstract: The present ECRH launcher installed on a single equatorial port of FTU tokamak was designed to inject four beams independently steered continuously in poloidal direction and in a set of predetermined toroidal angles. The launching mirrors can be moved only shot by shot. New experimental programmes for control of MHD instabilities with ECH/ECCD and heating of over-dense plasmas with ECBW require new launcher respectively with fast-steerable mirrors and increased toroidal capabilities. The required scanning speed for tracking the rational surfaces in the FTU plasma is 1cm in 10ms in poloidal direction, while the maximum toroidal angle needed for O–X–B heating scheme is around ±40°. Two ECRH lines, feeding the old launcher, will be switched to the new launcher, located in a different equatorial position, capable of launching two independent beams from small movable mirrors in the plasma proximity. A control on the power deposition width will be achieved by changing the beam radius in the plasma using an optical system composed by two mirrors (zooming range 17–28mm). Place has been reserved for future arrangements of additional components, e.g. a remote steering waveguide. A dedicated feedback control for the poloidal motion of the launching mirrors is being developed, in order to adapt the tracking of the power deposition location to the dynamic changes of magnetic surfaces in real-time. The maximum toroidal angle impacts strongly on the movable mirror design; dimensions (height around 90mm, width around 54mm) are limited by the port width (=80mm) and the need to preserve the maximum steering angle. Since the mirrors will not be actively cooled, temperature control will be achieved by covering the backside with a high emissivity coating, to obtain an efficient radiative cooling. A detailed description of the launcher is presented in the paper. [Copyright &y& Elsevier]

Abstract: In this paper we describe LiveMonitor, an integrated system realized for the distribution of information in fusion environments. The software tool is based on a client–server approach, where the server side consists of a set of Web Services that collect data from a variety of data sources. LiveMonitor has been successfully used at FTU, replacing and enhancing part of the core of the current message broadcasting system. The tool integrates all the information needed by the control room personnel during the experiments, namely the shot sequence status coming from the FTU Control System, videos of the plasma discharge from the FTU ports cameras, and fresh data from the databases. From the hardware point of view, the new system is made of a Linux node running the Web Services, while clients running on other machines can display information on large () LCD monitors. The tool has been tested during FTU experiments and can be further expanded to match the needs of the control room personnel and experimental physicists. [Copyright &y& Elsevier]

Abstract: In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1MV). These two different devices will separately address the main scientific and technological issues of the 17MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1h. The required negative ion current density to be extracted from the plasma source ranges from 290A/m2 in D2 (D−) and 350A/m2 in H2 (H−) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1m2. The main design choices will be presented in the paper as well as an overview of the design of the main components and systems. [Copyright &y& Elsevier]

Abstract: A prototype of a laser in vessel viewing and ranging system was developed at ENEA laboratories in Frascati, it uses the amplitude modulated laser radar concept and it is conceived to withstand the severe ITER conditions. The in vessel viewing system (IVVS) probe has been designed and built to perform sub-millimetric three-dimensional images inside ITER; it is based on an intrinsically radiation resistant concept and architecture. A first characterization of the IVVS probe under room conditions was made at Frascati Labs on a full-scale mock-up of the ITER first wall panel (FWP) and of the divertor vertical target. The first characterization demonstrated that an upgrade of the IVVS performances was necessary especially to cope with divertor surfaces made of carbon material, which is highly absorbent from optical point of view. The paper describes the new developments of IVVS prototype to increase range measurement performances that was obtained increasing the modulation frequency, the optical efficiency of the probe optics and the laser power. A new test campaign has been carried out on the upgraded IVVS and a compact characteristic curve describing its performances has been found both in mathematical and graphical form. As far as viewing is concerned, the system has confirmed the sub-millimetric viewing resolution, reaching in the worst cases ∼1mm of resolution. The image quality was excellent in almost all the cases. The range measurement performance of IVVS system has been strongly upgraded reducing the standard deviation of range measure of a factor varying from 6 to 12. The increased performance allows measuring surface shapes and erosion on first wall tiles and divertor also for inclination angles completely outside the previous IVVS characteristics. [Copyright &y& Elsevier]

Abstract: In the frame of the activities related to ITER divertor R&D, ENEA CR Brasimone was in charge by EFDA (European Fusion Development Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. the outer vertical target, the inner vertical target and the dome-liner, both in steady state and during draining and drying transient. The investigation was performed by means of both experimental test campaigns performed at ENEA CR Brasimone and theoretical simulation developed in RELAP5 Mod.3.3 environment at the Department of Nuclear Engineering of the University of Palermo (DIN). This paper presents the achieved experimental results for both steady state and transient tests. Moreover, the level of the implementation of the predictive hydraulic model, based on RELAP5 code, as well as its results are described, discussed and compared with the experimental ones. [Copyright &y& Elsevier]

To reach fusion conditions and control the plasma configuration in ITER, the next step in tokamak fusion research, two neutral beam injectors (NBIs) will supply 16.5 MW each, by neutralizing accelerated negative hydrogen or deuterium ions. The requirements of ITER NBIs (40A/1 MeV D- ions for ≤1 h, 46A/870 keV H- ions for ≤1000 s) have never been simultaneously attained. So in the Neutral Beam Test Facility (NBTF, Consorzio RFX, Italy) the operation of the full-scale ITER NBI prototype (MITICA) will be tested and optimised up to full performances, focussing on accelerator (including voltage holding), beam optics, neutralisation, residual ion removal. The NBTF includes also the full-scale prototype of the ITER NBI source with 100 keV particle energy (SPIDER), for early investigation of: negative ion production and extraction, source uniformity, negative ion current density and beam optics. This paper will describe the main results of the first two years of SPIDER operation, devoted to characterizing plasma and beam parameters, including investigation of RF-plasma coupling efficiency and magnetic filter field effectiveness in reducing co-extracted electrons. SPIDER is progressing towards the first caesium injection, which aims at increasing the negative ion density. A major shutdown, planned for 2021, to solve the issues identified during the operation and to carry out programmed modifications, will be outlined. The installation of each MITICA power supply and auxiliary system is completed; in-vessel mechanical components are under procurement by Fusion for Energy (F4E). Integration, commissioning and test of the power supplies, procured by F4E and QST, as the Japanese Domestic Agency (JADA), will be presented. In particular, 1.0MV insulating tests were carried out step-by-step and successfully completed. In 2020 integrated tests of the power supplies on the accelerator dummy load started, including the assessment of their resilience to accelerator grid breakdowns using a short-circuit device located in vacuum. The aggressive programme, to validate the NBI design at NBTF and to meet ITER schedule (requiring NBIs in operation in 2032), will be outlined. Unfortunately, in 2020 the coronavirus disease infection affected the NBTF activities. A solution to proceed with integrated power tests despite the coronavirus is presented. [ABSTRACT FROM AUTHOR]

• A new multiscale model for TRIEX-II facility has been presented. • The tool permits to fast compute both inventories and permeation fluxes. • The modelling results show well agreement with the experimental results. • The tool can be applied to more complex systems such as a breeding blanket and its ancillary systems. At ENEA Brasimone Research Centre, Italy, a new experimental facility named TRIEX-II (Tri tium Ex traction) was designed and installed. Its aim is to characterize, in the range of operating conditions foreseen for the European Test Blanket System WCLL-TBS, several extraction technologies for hydrogen isotopes (Q 2) solubilized in the flowing metallic LiPb alloy (15.7 at. % Li). One of these technologies is the packed column, an example of Gas/Liquid Contactors (GLCs). This paper proposes a multiscale modelling tool, combining different scales through two computational tools. The extraction column mock-up is described by a component-detail level model, developed in COMSOL Multiphysics, and integrated into a system level code of the whole TRIEX-II circuit, developed using MATLAB/Simulink. The integration is carried out by implementing the COMSOL component into an S-function of MATLAB/Simulink, preserving the process flow diagram of the loop. In this way, it was possible to quantify the Q 2 concentrations and the permeation fluxes, and to evaluate the theoretical extraction efficiency. The LiPb flow field inside the extractor was also derived. Finally, a comparison with the experimental results was performed. The results suggest that this tool could be adapted for analyses of complex systems, at a multiscale level, in view of design improvements and safety studies for the tritium cycle of ITER. [ABSTRACT FROM AUTHOR]