Neutral Beam Heating Research Articles

Ablation of solid pellets induced by supra-thermal ions in the far scrape-off layer of DIII-D plasmas

Injection of solid pellets is a key element in several aspects of the operation of a magnetic confinement plasma reactor for fusion applications, including plasma fueling, control and diagnosis. This letter reports observations demonstrating that pellet ablation can begin outside the plasma boundary, by effect of supra-thermal ions, whose orbits, under appropriate conditions, can extend well into the vacuum region. The phenomenon was recorded during plasma discharges in the DIII-D tokamak, combining pulsed modulation of heating neutral beams, with high-frequency injection of sub-millimeter lithium pellets, and it is ascribed to the large fraction of trapped beam ions associated with counter current neutral beam injection. The effect was quantitatively evaluated by means of Monte-Carlo simulations of supra-thermal ion orbits, finding that the heat-deposition was of the order of 50–100 W mm−2, in the region traversed by the lithium pellets before reaching the plasma boundary, which is consistent with the severe pellet deterioration observed for low velocity pellets.

Impact of ELM control techniques on tungsten sputtering in the DIII-D divertor and extrapolations to ITER

The free-streaming plus recycling model (FSRM) has recently been developed to understand and predict tungsten gross erosion rates from the divertor during edge localized modes (ELMs). In this work, the FSRM was tested against the experimental measurements of W sputtering during ELMs, conducted via fast neutral tungsten (WI) spectroscopy. Good agreement is observed using a variety of controlling techniques, including gas puffing, neutral beam heating, and plasma shaping to modify the pedestal stability boundary and, thus, the ELM behavior. ELM mitigation by pellet pacing was observed to strongly reduce W sputtering by flushing C impurities from the pedestal and reducing the divertor target electron temperature. No reduction of W sputtering was observed during the application of resonant magnetic perturbations (RMPs), in contrast to the prediction of the FSRM. Potential sources of this discrepancy are discussed. Finally, the framework of the FSRM is utilized to predict intra-ELM W sputtering rates in ITER. It is concluded that W erosion during ELMs in ITER will be caused mainly by free-streaming fuel ions, but free-streaming seeded impurities (N or Ne) may increase the erosion rate significantly if present in the pedestal at even the 1% level. Impurity recycling is not expected to cause significant W erosion in ITER due to the very low target electron temperature.

Numerical analyses and tests for optimized and enhanced heat transfer solutions in DEMO

A set of novel design solutions for high performance cooling systems have been developed and tested by Consorzio RFX, achieving, with experimental tests, the challenging heat transfer conditions foreseen for Heating and Current Drive Systems of present and future nuclear fusion devices. The project, called Multi-design Innovative Cooling Research & Optimization (MICRO), has the triple objective of:•Verifying the full qualification of the manufacturing process and assess the cooling performance of the present solutions applied on the acceleration grids designed for ITER Heating Neutral Beam Injectors and MITICA experiment, which have similar heat loads as for DEMO1.•Identifying alternative designs realizable with presently qualified manufacturing processes for improving heat transfer mechanisms with limited pressure drops.•Developing further optimized solutions by means of the investigation of the interrelated effects of geometric parameters and thermofluid dynamic properties of the coolant fluid.The main aim of present work is to extend the fatigue life-cycle of high thermal stress components. In the first part of the paper the design description and constraints are presented together with the results obtained from numerical analyses carried out on the full-scaled model of the electrostatic grids; the outcomes of the experimental campaign, together with the characterization of the properties and performance of a set of alternative dielectric fluids, are shown in the second one.

Design and preliminary operation of a laser absorption diagnostic for the SPIDER RF source

The ITER Heating Neutral Beam (HNB) injector is required to deliver 16.7 MW power into the plasma from a neutralised beam of H−/D− negative ions, produced by an RF source and accelerated up to 1 MeV. To enhance the H−/D− production, the surface of the acceleration system grid facing the source (the plasma grid) will be coated with Cs because of its low work function. Cs will be routinely evaporated in the source by means of specific ovens. Monitoring the evaporation rate and the distribution of Cs inside the source is fundamental to get the desired performances on the ITER HNB. In order to proper design the source of the ITER HNB and to identify the best operation practices for it, the prototype RF negative ion source SPIDER has been developed and built in the Neutral Beam Test Facility at Consorzio RFX. A Laser Absorption Spectroscopy diagnostic will be installed in SPIDER for a quantitative estimation of Cs density. By using a wavelength tunable laser, the diagnostic will measure the absorption spectrum of the 852 nm line along 4 lines of sight, parallel to the plasma grid surface and close to it. From the absorption spectra the line-integrated density of Cs at ground state will be measured. The design of this diagnostic for SPIDER is presented, with details of the layout and of the key components. A preliminary installation of the diagnostic on the test stand for Cs ovens is also described, together with its first experimental results; the effect of ground state depopulation on collected measurements is discussed and partially corrected.

Understanding helium transport: experimental and theoretical investigations of low-Z impurity transport at ASDEX Upgrade

The presence of helium is fundamentally connected to the performance of a fusion reactor, as fusion-produced helium is expected to heat the plasma bulk, while He ‘ash’ accumulation dilutes the fusion fuel. An understanding of helium transport via experimentally validated theoretical models of the low-Z impurity turbulent transport is indispensable to predict the helium density profile in future fusion devices. At ASDEX Upgrade, detailed, multi-species investigations of low-Z impurity transport have been undertaken in dedicated experiments, resulting in an extensive database of helium and boron density profiles over a wide range of parameters relevant for turbulent transport (normalised gradients of the electron density, the ion temperature, and the toroidal rotation profiles, the collisionality and the electron to ion temperature ratio). Helium is not found to accumulate in the parameter space investigated, as the shape of the helium density profile follows largely that of the electron density. Helium is observed to be as peaked as the electron density at high electron cyclotron resonance heating fraction, and less peaked than the electron density at high neutral beam heating fraction. The boron density profile is found to be consistently less peaked than the electron density profile. Detailed comparisons of the experimental density gradients of both impurities with quasilinear gyrokinetic simulations have shown that a qualitative agreement between experiment and theory cannot always be obtained, with strong discrepancies observed in some cases.

The High Voltage Deck 1 and Bushing for the ITER Neutral Beam Injector: Integrated design and installation in MITICA experiment

MITICA is the full scale prototype of the ITER Heating Neutral Beam (HNB) currently under construction at the Neutral Beam Test Facility in Padova (Italy). In the ITER HNB, negative ions (H−/D−) are produced in the Ion Source (IS) polarized to ground at −1012 kV, then extracted and accelerated to ground at 1 MeV energy and finally neutralized. The complex power supply system feeding the IS includes two non-standard equipment, beyond the current industrial standard for insulation voltage level (-1 MVdc) and dimensions: 1) the High Voltage Deck1 (HVD1), a large air insulated Faraday cage (12.5 m (L) x 8.4 m (W) x 9.6 m (H)) hosting the IS power supplies and diagnostics; 2) the High Voltage Bushing Assembly (HVBA), a -1 MVdc air to SF6 Bushing which interfaces the HVD1 with the SF6 insulated Transmission Line (TL) connecting the Acceleration Grid Power Supply system with the IS, carrying inside its High Voltage electrode all IS power and diagnostics conductors. The HVD1 and HVBA are installed inside a large HV hall. The main design choices leading to the final hall layout, integrating the HV experimental equipment and the conventional building plants assuring the necessary clearance required by such very high electric insulation level, are presented. Moreover, the paper reports on the manufacturing design developed by the supplier, the factory type tests to validate the design of the HVD1, the factory tests carried out on the HVBA and the on-site installation, commissioning and testing activities.

Progress in the design of MITICA control and interlock systems

The construction of MITICA, the full-size prototype of the ITER Heating Neutral Beam Injectors (HNBs) is in progress at the Neutral Beam Test Facility (NBTF) located in Padova, Italy. The design of the central control (CODAS) and interlock (CIS) systems is progressing taking into account the requirements coming from the MITICA plant units, in terms of number and type of interface signals, data sampling frequency values, operating states, protection actions and required reaction times and reliability. As the HNBs will be ITER components, additional requirements derive from the ITER plant control design handbook that specifies the design criteria and standard technologies to be applied to ITER plant systems.A distinction must be made in MITICA between plant units that will be part of the HNB, such as the power supply systems, and plant units that are auxiliaries, such as the gas and vacuum system. In the NBTF these components shall interface with MITICA CODAS and CIS, whereas in ITER they will be autonomous plant systems providing services to the HNB. The distinction reported above suggests defining the architecture of MITICA Instrumentation and Control (I&C) as split into the following functional blocks: i) HNB plant system including components that will be part of the HNB; ii) Service plant system including all operational plant units that need control in the NBTF but are implemented outside the ITER HNB; iii) Diagnostics plant system; iv) Central I&C (CODAS and CIS).The paper presents the requirements for MITICA CODAS and CIS and derives from these the system architecture. The design of CODAS and CIS is discussed with reference to the lesson learned in the implementation of SPIDER CODAS and CIS (the NBTF ion source test bed) and to the technologies suitable for the implementation.

Conceptual design of Integrated Bivalent Auxiliary System (SABI) supporting the MITICA cooling plant and cryogenic plant

The first ever built full-scale prototype of the ITER heating neutral beam injector is the MITICA experiment at PRIMA Neutral Beam Test Facility, under realization in Padova, Italy.This experiment consists of several in-vessel components: most of them are actively cooled by a large Cooling Plant, still under construction. Coolant is deionized water produced by a Chemical Control System (CCS).In the PRIMA facility, a cryogenic plant (currently in manufacturing phase) is also foreseen to serve the cryopumps to be installed in the MITICA Beam Line Vessel.During the installation of cooling plant and the design phase of cryogenic plant, two different issues came into evidence: waste water (produced by the CCS during operations) could not be delivered directly into the public network and the warm compressor of the cryogenic plant, originally air-cooled, changed to a water-cooled system.For the first issue the solution was found in a Water Treatment Unit (WTU) that requires the supply of hot water to distil the wastewater and to concentrate it.The second issue and the need to balance the heat loads and the cost and space constraints lead to an innovative system: it is essentially based on a high temperature water-water heat pump, by means of which the heat rejected by the warm compressor is used to heat up the WTU.Moreover this will reduce the operational costs by introducing a significant energy saving.

Cavity-Enhanced Photodetachment of H− as a Means to Produce Energetic Neutral Beams for Plasma Heating

Neutral beam injection, for plasma heating, will supposedly be achieved, in ITER, by collisional detachment of a pre-accelerated D − beam. Collisional detachment, however, makes use of a D 2 -filled neutralisation chamber, which has severe drawbacks, including the necessity to set the D − -ion source at −1 MV. Photodetachment, in contradistinction, would have several advantages as a neutralisation method, including the absence of gas injection, and the possibility to set the ion source close to the earth potential. Photodetachment, however, requires a very high laser flux. The presented work has consisted in implementing an optical cavity, with a finesse greater than 3000, to make such a high illumination possible with a state-of-the-art CW (continuous-wave) laser. A 1.2 keV 1 H − -beam (only 20 times slower than the 1 MeV 2 D − ion beams to be prepared for ITER) was photodetached with more-than-50% efficiency, with only 24 W of CW laser input. This experimental demonstration paves the way for developing real-size photoneutralizers, based on the implementation of refolded optical cavities around the ion beams of neutral beam injectors. Depending on whether the specifications of the laser power or the cavity finesse will be more difficult to achieve in real scale, different architectures can be considered, with greater or smaller numbers of optical refoldings or (inclusively) optical cavities in succession, on the beam to be neutralised.

Simulation and verification of air cooling system for -1MVdc MITICA High Voltage hall in Padova

The first ever built full-scale prototype of the ITER heating neutral beam injector is the MITICA experiment at PRIMA-NBTF (Neutral Beam Test Facility), under realization in Padua, Italy.The MITICA experiment includes many auxiliary plants; this paper is focused on the simulation and verification of the air cooling system of the High Voltage Power Supply (−1 MVdc), hosted in a Faraday cage (HVD1, High Voltage Deck1) inside a dedicated Building (High Voltage Hall, HVH) at PRIMA-NBTF.The HVD1 hosts the MITICA ISEPS (Ion Source Power Supply) in a two-floor cage, air insulated to ground for -1MVdc, with huge dimensions 12.5 m (L) × 8.4 m (W) × 9.6 m (H).Severe cleanliness constraints are required in the HVH hosting the HVD1, due to the high voltage operative conditions (the −1 MVdc Faraday cage is air insulated).The thermal behavior and air distribution in HVH are investigated as described in this paper.

The fabrication and assembly of the beam source for the SPIDER experiment

The ITER Heating Neutral Beam injector (HNB) will be equipped with a beam source that will provide a negative beam of 40A (H or D). The R&D activities undertaken in Europe to pursue this challenging goal comprise three experiments:-ELISE – the half size ion source experiment operating in IPP Garching-SPIDER – the full size ion source experiment at Neutral Beam Test Facility (NBTF) site in Padua-MITICA - the full size full energy ITER injector prototype being also established at the NBTFThe procurement of the beam source for SPIDER started in October 2012, when a contract between Fusion for Energy and Thales, Microwave & Imaging Sub-Systems (as group leader of a Consortium created with CECOM Srl, Galvano-T GmbH and Zanon SpA) was signed. The scope of the contract has been completed in October 2017 with the delivery of the SPIDER beam source at the NBTF site in Padova Italy.A review of the fabrication of the main parts and of the aspects of their assembly is reported in the paper, with a focus on some of the major issue encountered and the achievement of tight alignment tolerances of the accelerator.

Spider beam source ready for operation

The SPIDER Beam Source (BS), the first prototype of a full scale ion source for the ITER Heating Neutral Beam injector, was delivered to the Neutral Beam Test Facility (NBTF) site in Padova (Italy) after about five years procurement phase.A huge effort was devoted during the procurement for quality controls and testing at the supplier’s workshops. Several activities were also carried out on NBTF site for verification/adjustment of interfaces, solution of still open issues, as well as final tests before and after installation inside the vacuum vessel.The NBTF Team undertook the BS site acceptance tests including: pressure and leak tests of the hydraulic circuits; electrical tests; measurement of magnetic field profiles; functionality tests of diagnostics installed on the BS; checks of grids alignment by means of laser tracker.Accurate positioning of the BS inside the vacuum vessel was performed and various service lines were connected in a tight space. Several improvements were undertaken in order to guarantee reliability and reduce the risks during the commissioning and experimental campaign in 2018.After installation, the integrated commissioning phase was initiated, powering the RF and high voltage circuits, followed by the first operation in vacuum.

Plasma light detection in the SPIDER beam source

The source prototype SPIDER in the Neutral Beam Test Facility at Consorzio RFX has been developed to demonstrate the target performances of the RF plasma source of ITER Heating Neutral Beam injectors. SPIDER diagnostics characterize the plasma in the source, optimize performances and provide signals for machine protection. Photodiodes measure the time evolution of lines of sight integrated visible radiation emitted by the plasma in the source. Some are equipped with interference filters to select specific spectral lines, e.g. those of cesium to control its evaporation, which assists the H¯ production, and of metal impurities, which might indicate wall damages. Other lines of sight are fed to grating spectrometers to measure the spectrum at a slower rate, either at low resolution for spectral survey of hydrogen Balmer lines and impurities or at high resolution for molecular Fulcher bands. The overall layout of the photodiode diagnostic and preliminary results during the first SPIDER operation are presented. The custom photodiode electronics is described in detail, including a low-noise variable gain amplifier, FPGA-based remote control and suitably shaped digital output for feedback control. Main characteristics are: linear response up to 10 V, 14 bit dynamic range and 1.2–135 kHz bandwidth depending on gain in range 100 kΩ to 5 GΩ. Measured light through the drivers has allowed to recognise plasma features and dependencies with RF power and filter field current, as well as dominant frequencies in signal fluctuations.

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 (˜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 βN values (1.8) and Te/Ti ˜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 βN˜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).

SPIDER integrated commissioning

The construction of SPIDER (Source for the Production of Ions of Deuterium Extracted from a Radio frequency plasma), the prototype of the full-size, radio frequency, negative ion source of the ITER Heating Neutral Beam Injectors, has been completed at the Neutral Beam Test Facility in Padova (Italy). All SPIDER components have been delivered and have successfully undergone site acceptance tests. The mechanical components (vessel, ion source, accelerator, and calorimeter), power supply systems, auxiliary systems (gas and vacuum and cooling), basic diagnostics (thermocouples, imaging, spectroscopy), control and interlock systems were prepared for operation by means of a step-by-step procedure, referred to as integrated commissioning, aiming to progressively achieve the coordinated interoperation of all SPIDER plant systems. The execution of the integrated commissioning requires a well-defined organization covering personnel safety, shift-based operation, definition of roles and their interaction. The paper, after briefly summarizing the status of the SPIDER experiment, initially focuses on the organization of the integrated commissioning. The execution of the commissioning campaigns is then presented along with the technical and scientific issues encountered and the solutions applied to solve problems. Finally, the paper discusses the transition from the SPIDER integrated commissioning to SPIDER operation.

Characterization of the SPIDER Cs oven prototype in the CAesium Test Stand for the ITER HNB negative ion sources

The ITER Heating Neutral Beam (HNB) injector is required to deliver 16.7 MW of power to the plasma from a neutralized beam of H−/D− negative ions, produced by an RF source and accelerated up to 1 MeV. To enhance the H−/D− production, Cs will be routinely evaporated in the source by means of specific ovens embedded in the source, to reduce its work function. Controlling and monitoring the evaporation rate of Cs inside the source will be fundamental to achieve the desired performance for the ITER HNB.The prototype RF negative ion source SPIDER has been developed and built in the Neutral Beam Test Facility at Consorzio RFX. In SPIDER, liquid Cs based ovens will be used to inject Cs vapors inside the source. The CAesium Test Stand (CATS) has been specifically designed and set up for testing, commissioning, and characterizing Cs ovens in vacuum, but also to study the Cs evaporation and deposition onto surfaces. A SPIDER Cs oven prototype has been manufactured and tested in CATS in order to characterize its thermal behavior, by means of thermocouples and thermal camera, and its Cs flux, by means of Surface Ionization Detectors and Laser Absorption Spectroscopy.

Conceptual design of a neutral beam heating & current drive system for DTT

The main purpose of the Divertor Tokamak Test (DTT) is to study solutions to mitigate the issue of power exhaust in conditions relevant for ITER and DEMO. The key feature of such a study is to equip the machine with a significant amount of auxiliary heating power (45 MW) in order to test different divertor solutions. According to the Italian project, the experiment is foreseen to operate with the following main parameters: BT = 6 T, IP = 5.5 MA, R0 = 2.1 m, a = 0.65 m and a pulse duration of 95 s. It shall be able to study different divertor magnetic configurations and reach a reactor relevant power flow to the divertor. The proposed mix of heating power foreseen to achieve the target value of 45 MW delivered to the plasma will be provided by Electron Cyclotron Resonant Heating (ECRH), Ion Cyclotron Resonant Heating (ICRH) and Negative-ion-based Neutral Beam Heating (NNBH). In this framework, the conceptual design of a NNBH system for DTT is here presented, with a particular focus on the technical solutions adopted to fulfil the requirements and maximize the performances. The proposed system features two beamlines providing deuterium negative ions (D−) with an energy not smaller than 300 keV and an injected power of 5–8 MW each. The design of the main components of the injectors is described in detail, explaining the motivations behind the main design choices and the related evaluations by means of physics and engineering simulations.

Bursting toroidal Alfvén eigenmodes in KSTAR plasmas

We report on observations of bursty mode activity during early neutral beam heating in KSTAR plasmas, before current flat top while the q profile is still evolving. The magnitude of the activity increases with early beam heating, and reduces with the addition of resonant magnetic perturbation magnetic field coils. A mode analysis yields a toroidal mode number of n = 2. The mode is observed to be downward chirping in frequency, and exists for the duration of the slowing-down of the beam. Motional Stark effect constrained equilibrium reconstructions are available at adjacent time slices: we have rescaled the total current to the measured value to obtain the q profile during the mode activity. From this we have computed the mode spectrum and identified a number of candidate gap modes. Wave-particle simulations with plausible distribution functions are computed, which demonstrate that the lowest frequency mode satisfies the condition for wave drive , where is the fast ion diamagnetic drift frequency. An interesting finding is the change from exponential growth of the mode above , whereby the mode continues to nonlinearly grow at a reduced rate over a period of 100 wave periods up to final saturated amplitude. We believe that this may be because the two spatial resonances at s = 0.4 and s = 0.8 overlap for sufficiently high fast ion density, and so the phase-space volume and fast ion density available to drive the mode increases.

Safety systems in the ITER neutral beam test facility

SPIDER, the experimental device of the ITER Neutral Beam Test Facility (NBTF) devoted to the study of the ion source, has recently started operation. The construction of MITICA, the full-size prototype of the ITER heating neutral beam injector, is in progress. SPIDER and MITICA operation presents many possible hazards including radiation, electrical, explosive gas, laser and fire risks. An integrated approach, referred to as the NBTF Safety System, was chosen to deal globally with all risks for people and environment safety. As many different systems collaborate to guarantee the NBTF safety, a coordinating system, called Central Safety System (CSS), was implemented to supervise safety-relevant systems and provide an effective human interface of the NBTF safety. The CSS is also required to implement directly safety instrumented functions. The requirements of the safety systems derived from the safety risk analysis. The CSS uses the Siemens SIMATIC S7 400FH PLC along with safe input/output remote nodes connected through double ring-topology, redundant Profibus network. Complying with the IEC 61508/61511 standards, the system architecture uses the PROFISafe profile and the human machine interface is based on WinCC Open Architecture. Special provisions are used for communication of safety-related information with the PLC.

Thermo-hydraulic analyses and fatigue verification of the Electrostatic Residual Ion Dump for the ITER HNB

In the ITER Heating Neutral Beam Injector (HNB) the remaining charged particles after the neutralization process will be removed by an Electrostatic Residual Ion Dump (ERID) where electrostatic fields are used to deflect the ions that are so dumped on to five panels, each one composed of 18 separate CuCrZr Beam Stopping Elements (BSEs). A customized finite element (FE) code was developed to allow parametric and coupled solid-fluid simulations in the sub-cooled boiling conditions expected in the swirl flow of cooling channels. Thereafter, thermal-mechanical analyses and fatigue verifications of the ERID panels were undertaken with non-linear material properties by using elasto-plastic properties of CuCrZr and by considering the most demanding condition produced by breakdown and beam-on/off cycles during the operation.