Test of ID carbon-carbon composite prototype tiles for the SPIDER diagnostic calorimeter

Additional heating will be provided to the thermonuclear fusion experiment ITER by injection of neutral beams from accelerated negative ions. In the SPIDER test facility, under construction at Consorzio RFX in Padova (Italy), the production of negative ions will be studied and optimised. To this purpose the STRIKE (Short-Time Retractable Instrumented Kalorimeter Experiment) diagnostic will be used to characterise the SPIDER beam during short operation (several seconds) and to verify if the beam meets the ITER requirement regarding the maximum allowed beam non-uniformity (below \ub110%). The most important measurements performed by STRIKE are beam uniformity, beamlet divergence and stripping losses. The major components of STRIKE are 16 1D-CFC (Carbon matrix-Carbon Fibre reinforced Composite) tiles, observed at the rear side by a thermal camera. The requirements of the 1D CFC material include a large thermal conductivity along the tile thickness (at least 10 times larger than in the other directions); low specific heat and density; uniform parameters over the tile surface; capability to withstand localised heat loads resulting in steep temperature gradients. So 1D CFC is a very anisotropic and delicate material, not commercially available, and prototypes are being specifically realised. This contribution gives an overview of the tests performed on the CFC prototype tiles, aimed at verifying their thermal behaviour. The spatial uniformity of the parameters and the ratio between the thermal conductivities are assessed by means of a power laser at Consorzio RFX. Dedicated linear and non-linear simulations are carried out to interpret the experiments and to estimate the thermal conductivities; these simulations are described and a comparison of the experimental data with the simulation results is presented

Complete compensation of criss-cross deflection in a negative ion accelerator by magnetic technique

During 2016, a joint experimental campaign was carried out by QST and Consorzio RFX on the Negative Ion Test Stand (NITS) at the QST Naka Fusion Institute, Japan, with the purpose of validating some design solutions adopted in MITICA, which is the full-scale prototype of the ITER NBI, presently under construction at Consorzio RFX, Padova, Italy. The main purpose of the campaign was to test a novel technique, for suppressing the beamlet criss-cross magnetic deflection. This new technique, involving a set of permanent magnets embedded in the Extraction Grid, named Asymmetric Deflection Compensation Magnets (ADCM), is potentially more performing and robust than the traditional electrostatic compensation methods. The results of this first campaign confirmed the effectiveness of the new magnetic configuration in reducing the criss-cross magnetic deflection. Nonetheless, contrary to expectations, a complete deflection correction was not achieved. By analyzing in detail the results, we found indications that a physical process, taking place just upstream of the plasma grid, was giving an important contribution to the final deflection of the negative ion beam. This process appears to be related to the drift of negative ions inside the plasma source, in the presence of a magnetic field transverse to the extraction direction, and results in a non-uniform ion current density extracted at the meniscus. Therefore, the numerical models adopted in the design were improved by including this previously disregarded effect, so as to obtain a much better matching with the experimental results. Based on the results of the first campaign, new permanent magnets were designed and installed on the Extraction Grid of NITS. A second QST-Consorzio RFX joint experimental campaign was then carried out in 2017, demonstrating the complete correction of the criss-cross deflection and confirming the validity of the novel magnetic configuration and of the hypothesis behind the new models. This contribution presents the results of the second joint experimental campaign on NITS along with the overall data analysis of both campaigns, and the description of the improved models. A general picture is given of the relation among magnetic field, beam energy, meniscus non-uniformity and beamlet deflection, constituting a useful database for the design of future machines

Beamlet scraping and its influence on the beam divergence at the BATMAN Upgrade test facility

For the ITER fusion experiment, two neutral beam injectors are required for plasma heating and current drive. Each injector supplies a power of about 17 MW, obtained from neutralization of 40 A (46 A), 1 MeV (0.87 MeV) negative deuterium (hydrogen) ions. The full beam is composed of 1280 beamlets, formed in 16 beamlet groups, and strict requirements apply to the beamlet core divergence (<7 mrad). The test facility BATMAN Upgrade uses an ITER-like grid with one beamlet group, which consists of 70 apertures. In a joint campaign performed by IPP and Consorzio RFX to better assess the beam optics, the divergence of a single beamlet was compared to a group of beamlets at BATMAN Upgrade. The single beamlet is measured with a carbon fiber composite tile calorimeter and by beam emission spectroscopy, whereas the divergence of the group of beamlets is measured by beam emission spectroscopy only. When increasing the RF power at low extraction voltages, the divergence of the beamlet and of the group of beamlets is continuously decreasing and no inflection point toward an overperveant beam is found. At the same time, scraping of the extracted ion beam at the second grid (extraction grid) takes place at higher RF power, supported by the absence of the normally seen linear behavior between the measured negative ion density in the plasma close to the extraction system and the measured extracted ion current. Beside its influence on the divergence, beamlet scraping needs to be considered for the determination of the correct perveance and contributes to the measured coextracted electron current

Direct current measurements of the SPIDER beam: a comparison to existing beam diagnostics

For negative ion beam sources there are several methods of measuring the accelerated beam current, most commonly electrical measurements at the power supply and calorimetric measurements. On SPIDER, the ITER Heating Neutral Beam full-scale beam source prototype, electrical measurements at the acceleration grid power supply (AGPS) are complemented by polarizing the diagnostic calorimeter STRIKE to provide an additional electrical measurement of the accelerated current. This is in addition to the calorimetric measurements provided by STRIKE. These diagnostics give differing measurements of the beam current. Exploiting the reduced number of open apertures on SPIDER a new beam diagnostic has been installed to measure the individual beamlet currents directly. The so called Beamlet Current Monitor (BCM) has been used to measure the current of five beamlets during the most recent SPIDER campaign. This work compares the BCM current to the electrical measurements at the AGPS and STRIKE. The average BCM current agrees well with the STRIKE electrical measurements, indicating that the AGPS overestimates the beam current. The individual beamlets are compared to the STRIKE calorimetric measurements, showing similar current trends with the source parameters

Start of SPIDER operation towards ITER neutral beams

Heating Neutral Beam (HNB) Injectors will constitute the main plasma heating and current drive tool both in ITER and JT60-SA, which are the next major experimental steps for demonstrating nuclear fusion as viable energy source. In ITER, in order to achieve the required thermonuclear fusion power gain Q=10 for short pulse operation and Q=5 for long pulse operation (up to 3600s), two HNB injectors will be needed [1], each delivering a total power of about 16.5 MW into the magnetically-confined plasma, by means of neutral hydrogen or deuterium particles having a specific energy of about 1 MeV. Since only negatively charged particles can be efficiently neutralized at such energy, the ITER HNB injectors [2] will be based on negative ions, generated by caesium-catalysed surface conversion of atoms in a radio-frequency driven plasma source. A negative deuterium ion current of more than 40 A will be extracted, accelerated and focused in a multi-aperture, multi-stage electrostatic accelerator, having 1280 apertures (~ 14 mm diam.) and 5 acceleration stages (~200 kV each) [3]. After passing through a narrow gas-cell neutralizer, the residual ions will be deflected and discarded, whereas the neutralized particles will continue their trajectory through a duct into the tokamak vessels to deliver the required heating power to the ITER plasma for a pulse duration of about 3600 s. Although the operating principles and the implementation of the most critical parts of the injector have been tested in different experiments, the ITER NBI requirements have never been simultaneously attained. In order to reduce the risks and to optimize the design and operating procedures of the HNB for ITER, a dedicated Neutral Beam Test Facility (NBTF) [4] has been promoted by the ITER Organization with the contribution of the European Union\u2019s Joint Undertaking for ITER and of the Italian Government, with the participation of the Japanese and Indian Domestic Agencies (JADA and INDA) and of several European laboratories, such as IPP-Garching, KIT-Karlsruhe, CCFE-Culham, CEA-Cadarache. The NBTF, nicknamed PRIMA, has been set up at Consorzio RFX in Padova, Italy [5]. The planned experiments will verify continuous HNB operation for one hour, under stringent requirements for beam divergence (< 7 mrad) and aiming (within 2 mrad). To study and optimise HNB performances, the NBTF includes two experiments: MITICA, full-scale NBI prototype with 1 MeV particle energy and SPIDER, with 100 keV particle energy and 40 A current, aiming at testing and optimizing the full-scale ion source. SPIDER will focus on source uniformity, negative ion current density and beam optics. In June 2018 the experimental operation of SPIDER has started

Status, scientific results and technical improvements of the NBH on TCV tokamak

The TCV tokamak contributes to physics understanding in fusion reactor research by a wide set of experimental tools, like flexible shaping and high power ECRH. A 1MW, 25 keV deuterium heating neutral beam (NB) has been installed in 2015 and it was operated from 2016 in SPC-TCV domestic and EUROfusion MST1 experimental campaigns ((similar to)50/50%). The rate of failures of the beam is less than 5%. Ion temperatures up to 3.5 keV have been achieved in ELMy H-mode, with a good agreement with ASTRA predictive simulations. The NB enables TCV to access ITER-like beta(N) values (1.8) and T-e/T-i (similar to)1, allowing investigations of innovative plasma features in ITER relevant ELMy H-mode. The advanced Tokamak route was also pursued, with stationary, fully non-inductive discharges sustained by ECCD and NBCD reaching beta(similar to)(N)1.4-1.7. Real-time control of the NB power has been implemented in 2018 and presented together with the statistics of NB operation on the TCV. During commissioning, the NB showed unacceptable heating of the TCV beam duct, indicating a higher power deposition than expected on duct walls. A high beam divergence has been found by dedicated measurement of 3-D beam power density distribution with an expressly designed device (IR measurement on tungsten target)

Continuous pulse advances in the negative ion source NIO1

Consorzio RFX and INFN-LNL have designed, built and operated the compact radiofrequency negative ion source NIO1 (Negative Ion Optimization phase 1) with the aim of studying the production and acceleration of H- ions. In particular, NIO1 was designed to keep plasma generation and beam extraction continuously active for several hours. Since 2020 the production of negative ions at the plasma grid (the first grid of the acceleration system) has been enhanced by a Cs layer, deposited though active Cs evaporation in the source volume. For the negative ion sources applied to fusion neutral beam injectors, it is essential to keep the beam current and the fraction of co-extracted electrons stable for at least 1 h, against the consequences of Cs sputtering and redistribution operated by the plasma. The paper presents the latest results of the NIO1 source, in terms of caesiation process and beam performances during continuous (6{\div}7 h) plasma pulses. Due to the small dimensions of the NIO1 source (20 x (diam.)10 cm), the Cs density in the volume is high (10^15 \div 10^16 m^-3) and dominated by plasma-wall interaction. The maximum beam current density and minimum fraction of co-extracted electrons were respectively about 30 A/m^2 and 2. Similarly to what done in other negative ion sources, the plasma grid temperature in NIO1 was raised for the first time, up to 80 {\deg}C, although this led to a minimal improvement of the beam current and to an increase of the co-extracted electron current.Comment: 11 pages, 7 figures. Contributed paper for the 8th International symposium on Negative Ions, Beams and Sources - NIBS'22. Revision 1 of the preprint under evaluation at Journal of Instrumentation (JINST

Status and future development of Heating and Current Drive for the EU DEMO

The European DEMO is a pulsed device with pulse length of 2 hours. The functions devoted to the heating and current drive system are: plasma breakdown, plasma ramp-up to the flat-top where fusion reactions occur, the control of the plasma during the flat-top phase, and finally the plasma ramp-down. The EU-DEMO project was in a Pre-Concept Design Phase during 2014-2020, meaning that in some cases, the design values of the device and the precise requirements from the physics point of view were not yet frozen. A total of 130 MW was considered for the all phases of the plasma: in the flat top, 30 MW is required for neoclassical tearing modes (NTM) control, 30 MW for burn control, and 70 MW for the control of thermal instability (TI), without any specific functions requested from each system, Electron Cyclotron (EC), Ion Cyclotron (IC), or Neutral Beam (NB) Injection. At the beginning of 2020, a strategic decision was taken, to consider EC as the baseline for the next phase (in 2021 and beyond). R&D on IC and NB will be risk mitigation measures. In parallel with progresses in Physics modelling, a decision point on the heating strategy will be taken by 2024. This paper describes the status of the R&D development during the period 2014-2020. It assumes that the 3 systems EC, IC and NB will be needed. For integration studies, they are assumed to be implemented at a power level of at least 50 MW. This paper describes in detail the status reached by the EC, IC and NB at the end of 2020. It will be used in the future for further development of the baseline heating method EC, and serves as starting point to further develop IC and NB in areas needed for these systems to be considered for DEMO

Status and future development of Heating and Current Drive for the EU DEMO

The European DEMO is a pulsed device with pulse length of 2 hours. The functions devoted to the heating and current drive system are: plasma breakdown, plasma ramp-up to the flat-top where fusion reactions occur, the control of the plasma during the flat-top phase, and finally the plasma ramp-down. The EU-DEMO project was in a Pre-Concept Design Phase during 2014-2020, meaning that in some cases, the design values of the device and the precise requirements from the physics point of view were not yet frozen. A total of 130 MW was considered for the all phases of the plasma: in the flat top, 30 MW is required for neoclassical tearing modes (NTM) control, 30 MW for burn control, and 70 MW for the control of thermal instability (TI), without any specific functions requested from each system, Electron Cyclotron (EC), Ion Cyclotron (IC), or Neutral Beam (NB) Injection. At the beginning of 2020, a strategic decision was taken, to consider EC as the baseline for the next phase (in 2021 and beyond). R&D on IC and NB will be risk mitigation measures. In parallel with progresses in Physics modelling, a decision point on the heating strategy will be taken by 2024. This paper describes the status of the R&D development during the period 2014-2020. It assumes that the 3 systems EC, IC and NB will be needed. For integration studies, they are assumed to be implemented at a power level of at least 50 MW. This paper describes in detail the status reached by the EC, IC and NB at the end of 2020. It will be used in the future for further development of the baseline heating method EC, and serves as starting point to further develop IC and NB in areas needed for these systems to be considered for DEMO

SPIDER Beam Homogeneity Characterization Through Visible Cameras

High energy beams of negative hydrogen and deuterium ions are needed to heat and sustain the plasma of future nuclear fusion reactors, in particular, in the experimental reactor International Thermonuclear Experimental Reactor (ITER). Besides the beam energy, low divergence ( ${\mathbf{2}}$ in deuterium and 350 $\mathbf{A/}\mathbf{m}{\mathbf{2}}$ in hydrogen. Source for the production of ions of deuterium extracted from a radio-frequency plasma (SPIDER), the full-size prototype of the ITER negative ion source, is equipped with a tomographic system consisting of 15 visible cameras, which observe the light produced by the interactions of the beam particles with the background gas after the accelerator, all around the beam itself, allowing a complete characterization of the beam shape and intensity. In fact, when the beam particles propagate in the background gas, they emit light in the visible range, due to the production of excited neutrals and ionization of the background particles. This light allows studying the beam properties since it is proportional to the beam current density itself. In the SPIDER ion source, magnetic and electric fields are used to optimize the beam current density, by reducing the electron temperature and density close to the extraction region. Also, cesium is evaporated in the plasma as a catalyst of negative ion production. In this work, the impact of these fields and of the cesium presence on beam properties will be discussed by means of visible tomography, using both the 1-D beam profiles and the 2-D tomographic reconstructions