Neutral Beam Injection Research Articles

Suppression of anomalous impurity transport in NBI-heated W7-X plasmas

Radial impurity density profiles in two Wendelstein 7-X (W7-X) experiments heated by neutral beam injection (NBI) are analyzed with respect to their impurity transport properties. Local impurity densities are derived from charge exchange reactions with the W7-X NBI system using simulated neutral densities cross-validated with beam emission spectroscopy. Impurity profiles of argon and carbon are found to show an evolving central accumulation inside half radius. The properties of the underlying impurity transport are assessed using the one-dimensional transport code pySTRAHL. Comparisons between simulation and experiment indicate transport dominated by anomalous diffusion outside half radius. The observed central impurity accumulation is found to match best simulations with purely (neo-)classical transport in the accumulation region. This data implies a suppression of the anomalous impurity transport channel to below 35% of the (neo-)classical one. Experimental data is found to be matched best when invoking a time evolving, inward propagating zone where anomalous impurity transport is suppressed. An additional central power deposition into a plasma with central impurity accumulation via electron cyclotron resonance heating is found to affect the (neo-)classical transport components in case of operation times below 200 ms only. For longer operation times, it is found to re-introduce an altered level of anomalous diffusion. The existence of an inward directed anomalous pinch as an alternative explanation for the central impurity accumulation cannot explain the observed profiles.

TCV Tokamak Neutron Shielding Upgrade for Dual NBI Operation

The Tokamak à Configuration variable (TCV) is equipped with two neutral beam injection (NBI) systems delivering up to 1.2 MW each for pulse durations of up to 2 s. The first system (NBI1), designed for an injection energy in the range of 25 to 30 keV has been operational since 2016. The existing concrete neutron shielding of the experimental hall proved insufficient for fully protecting human accessible areas, limiting the number of daily plasma pulses using NBI1. The recently commissioned second system (NBI2) is designed for injection synergies in the range 50 to 60 keV. Both systems are tangentially oriented in opposite directions in order to permit experiments with low or no net torque. Calculations with the TRANSP and ORBIS heating codes show that neutron rates from deuterium-deuterium fusion reactions may be as high as 1014 n/s, up to 10 times higher than with the lower energy beam only. This is due both to the ~five times larger beam-plasma neutron rates from the higher energy beam and to an exceptionally high contribution from beam-beam reactions between the opposing beams. The radiation protection policy at the Swiss Plasma Center is that all staff members be considered as members of the general public, limiting the daily personal dose to 4 µSv. This is also the maximum admissible daily dose in any publicly accessible zone, whether occupied or not. Currently, with only the lower energy beam, this limit can be attained in the control room adjacent to the device hall after only five NBI pulses out of a possible 30 daily pulses. To allow for exploitation of the two beams at full specifications, the source side of the existing barite concrete walls of the 15 × 20 × 8 m large TCV hall will be covered with 20-cm-thick polythene (PE) cladding and a ceiling made of 35-cm-thick PE will be added. The total mass of PE will be 200 tons. The usage of PE at this scale for neutron shielding is unprecedented at any fusion research facility.

Comparison among possible design solutions for the Stray Field Shielding System of the DTT Neutral Beam Injector

Within the effort on the conceptual design of the Divertor Test Tokamak (DTT) Neutral Beam Injector (NBI), the design of the Stray Field Shielding System (SFSS) for DTT NBI is under development, to suppress the potentially harmful effects of the stray poloidal field from the tokamak on the accelerated charged beam. Various possible design solutions to solve this problem are presented and compared, with a particular focus on the stray field minimization procedure and particle tracing simulations, used during the design validation phase with the objective of maximizing beamline performances.

Enhancements of the Fast Interlock System for Wendelstein 7-X Operational Phase OP 2.1

An extended modification phase of the superconducting fusion experiment Wendelstein 7- X was completed late 2021 after installing actively cooled divertor units and the integration or modification of new technical components and diagnostics. Currently, the commissioning of the W7-X experiment is underway. First technical plasma operation is scheduled for the end of September 2022, and scientific plasma operation will start at the end of November. The Fast Interlock System (FIS) is an important component of the Safety Instrumented System (SIS) of W7-X. The main task of the FIS is to protect the inner plasma vessel components from thermal overload from both plasma heating systems (e.g. neutral beam injection) and the plasma itself. There are strict limits for reaction times after detection of interlock relevant events, which have to be observed in order to prevent overload situations of the first wall. The requirements for the FIS were substantially modified and extended, in order to address the challenges due to the actively cooled divertor. This led to extensive revision of the technical implementation and the software of the FIS. In this paper, after a short introduction of FIS design and operation in OP1, the modification of FIS for OP2.1 operation is presented. Finally, the first results during the OP2.1 commissioning and operation of the FIS are discussed.

Assessment of IEEE 1588-Based Timing System of the ITER Neutral Beam Test Facility

MITICA is one of the two ongoing experiments at the ITER Neutral Beam Test Facility (NBTF) located in Padova (Italy). MITICA aims to develop the full-size neutral beam injector of ITER and, as such, its Control and Data Acquisition System will adhere to ITER CODAC directives. In particular, its timing system will be based on the IEEE1588 PTPv2 protocol and it will use the ITER Time Communication Network (TCN) on a dedicated Ethernet LAN. Following the ITER device catalog, the National Instruments PXI-6683H PTP timing modules will be used to generate triggers and clocks synchronized with a PTP grandmaster clock. The “Lazy Trigger" technique, then, will be adopted to implement event-driven data acquisition without the need of any hardware link in addition to the Ethernet connections used to transfer data and timing synchronization. To evaluate the accuracy over time that can be achieved with different network topologies and configurations, a test system has been set-up consisting of a grand master clock, two PXI-6683H devices and two PTP aware network switches. In addition, a detailed simulation of the network and the involved devices has been performed using the OMNET++ discrete event simulator. The simulation parameters include not only the network and switches configuration, but also the PID parameters used in the clock servo controllers. A comparison between simulated and measured statistics is reported, together with a discussion on the possible optimal configuration strategies.

Status of SPIDER beam source after the first 3.5 years of operation

SPIDER is the negative ion source testbed for ITER neutral beam injector. It is currently the largest negative ion source ever built, equipped with a 100 keV accelerator, aiming at producing a negative ion beam with an extracted current density in hydrogen of 355 A/m2, beam-on time of one hour. During the first 3.5 years of operation several improvements on the SPIDER source and its operating conditions were implemented. This paper focuses on the description of materials degradation phenomena in SPIDER and on the lessons learnt that were derived from components inspection during the already planned shutdown in 2022. The aim of these inspections was to improve the design of some components and to repair or refurbish damaged components. In particular, this paper describes the characterisation of damaged molybdenum coating, testing of surface cleaning techniques and experimental evidence of caesium vapour uneven distribution inside the plasma chamber. Sputtering due to beam particles was also investigated experimentally to observe its effect on materials used in SPIDER.

Control method of DC component for NNBI accelerator power supply

In the high voltage power supply of the negative ion source neutral beam injection system (NNBI) accelerator, the three-phase three-level neutral point clamping (NPC) inverter, as an inverter module, has a detectable DC component in its output. When it flows into the step-up transformer, it will lead to increased transformer loss and excitation saturation, thus reducing the service life of the transformer and increasing potential safety hazards. This paper analyzes the causes of the DC component of the NPC inverter, and a reliable DC component suppression method based on double-loop control is proposed. A suitable PI controller can effectively suppress the output DC component during the power supply operation by establishing the transfer function between the output DC component and the regulation rate of the switch’s duty cycle. Finally, the prototype experiments verify the effectiveness of the control method successfully.

4D and 5D phase-space tomography using slowing-down physics regularization

We compute reconstructions of 4D and 5D fast-ion phase-space distribution functions in fusion plasmas from synthetic projections of these functions. The fast-ion phase-space distribution functions originating from neutral beam injection (NBI) at TCV and Wendelstein 7-X (W7-X) at full, half, and one-third injection energies can be distinguished and particle densities of each component inferred based on 20 synthetic spectra of projected velocities at TCV and 680 at W7-X. Further, we demonstrate that an expansion into a basis of slowing-down distribution functions is equivalent to regularization using slowing-down physics as prior information. Using this technique in a Tikhonov formulation, we infer the particle density fractions for each NBI energy for each NBI beam from synthetic measurements, resulting in six unknowns at TCV and 24 unknowns at W7-X. Additionally, we show that installing 40 LOS in each of 17 ports at W7-X, providing full beam coverage and almost full angle coverage, produces the highest quality reconstructions.

Development of non evaporable getter pumps for large hydrogen throughput and capacity in high vacuum regimes

In vacuum technology, capture pumps based on Non Evaporable Getters are commonly applied to ultra-high vacuum systems. Recent improvements in the absorption of hydrogenic species, with the introduction of Zr–V–Ti–Al alloys (ZAO®), make them an appealing and viable solution for the application in fusion research, and in particular for the vacuum system of neutral beam injectors (hydrogen pumping speed of thousands of m3/s, pressure of tens of mPa). This paper describes the characterization of the new NEG material in pumps of increasing dimensions, including the development, construction and test of a large mockup pump of modular design, to demonstrate the scalability of the technology. Effective pumping speeds of the order of 14 m3/s or higher at a concentration of 130 Pa m3/kg were achieved by the mockup pump, for an installed getter mass of about 16 kg, and a stability within 10% up to 1300 Pa m3/kg The measured effective pumping speed per unit area of sintered disks is of the order of 3.5 m/s, corresponding to 4.9 m/s at the disk surfaces as derived from numerical simulations. General guidelines for the design of large NEG pumps for hydrogen are discussed, including thermal aspects and duty cycle of the pump.

3D computational fluid dynamics analysis of PINI ion source back plate under high heat flux condition

Neutral Beam Injection (NBI) plays an essential role in Tokamak plasma heating and current drive. A positive ion-based NBI (PNBI) system is adopted in Steady State Superconducting Tokamak- 1 (SST-1). This PNBI system is capable of generating neutral hydrogen beam power of 1.7 MW at 55 keV. This system has a JET PINI (Positive Ion Neutral Injector) type of ion source. The Back Plate (BP) is a component of the PINI ion source. It consists of an SS 304 L magnet positioning & cover plate, and an OFE copper cooling plate. The back plate plays a vital role in removing the high heat load intercepted during beam operation. The present study describes 3D Computational Fluid Dynamics (CFD) analysis of the actual size back plate using the ANSYS version of R121. The 3D CFD analysis has been done under a 2.5 MW/m2 steady heat load incident on the OFE copper plate. Cooling water is provided in the Inlet header of the back plate with a mass flow rate of 1 kg/s at 34 °C. The analysis gives the surface temperature distribution over the OFE copper cooling plate with a maximum surface temperature of 174 °C and the average temperature is 122 °C. This result is in good agreement with High Heat Flux experimental results.

Model-based linear–quadratic–integral controller for simultaneous regulation of the current profile and normalized beta in NSTX-U

Achieving advanced scenarios that are characterized by steady-state operation, stable plasma confinement, and high-performance plasmas is one of the primary objectives of the National Spherical Tokamak eXperiment-Upgrade (NSTX-U). Active control of the plasma may be necessary to achieve these conditions. In particular, control algorithms that can simultaneously optimize the shapes and values of plasma profiles and scalars, respectively, may play a critical role in robustly achieving and sustaining these advanced scenarios. In this work, a model-based optimal control algorithm is developed for feedback control of the current profile in NSTX-U. The linear, finite-dimensional, control model is derived by discretizing and linearizing the magnetic diffusion equation in combination with empirical correlations for electron density, electron temperature, and noninductive current drives. The linear, time-variant model is then used for designing a linear–quadratic–integral (LQI) controller that is capable of regulating both the safety factor and the normalized beta around desired targets. The controller determines the neutral beam injection powers and the overall plasma current that are needed to achieve the desired current profile and normalized beta. The proposed controller is tested in higher-fidelity nonlinear simulations that employ 1D models for the evolutions of both current and temperature profiles using the Control Oriented Transport SIMulator (COTSIM). The closed-loop simulations show the effectiveness of the controller at shaping the safety factor in NSTX-U while achieving the desired normalized beta.

Predictive simulations for plasma scenarios in the SMART tokamak

The SMall Aspect Ratio Tokamak (SMART) is a new spherical machine that is currently being constructed at the University of Seville (Mancini et al., 2021; Agredano-Torres et al., 2021). The operation of SMART will cover three different phases reaching an inductive plasma current (IP) of more than 500 kA, a toroidal magnetic field (BT) of 1 T and a pulse length of 500 ms (Mancini et al., 2021; Agredano-Torres et al., 2021). The main goal of the SMART tokamak is to study high plasma confinement regimes in a broad triangularity range (-0.5≤δ≤0.5) (Doyle et al., 2021; Doyle et al., 2021). While in phase 1 the ohmic heating alone is expected to provide enough power to access the H-mode, in phase 2 and phase 3 the access to the H-mode will be ensured by applying Neutral Beam Injection (NBI) as external heating system. The NBI will consist of one injector at 25 keV and 1 MW of power. The overall design of the NBI, including injection geometry, energy and power have been optimized using the ASCOT5 code (Hirvijoki et al., 2021). The SMART scenarios have been developed with the help of the free boundary equilibrium solver code FIESTA (Cunningham, 2013) coupled to the linear time independent, rigid plasma model RZIP (Lazarus et al., 1990) to calculate the target equilibria for all the different operational phases. To assess the feasibility of those scenarios, predictive modelling needs to be included to evaluate properly the evolution of the temperatures, density profiles for both electrons and ions. To this extent, the 1.5D transport code ASTRA (Pereverzev and Yushmanov, 2002) has been used including models for the ohmic current, bootstrap current and current driven by NBI. This contribution discusses the electron and ion density and temperature profiles obtained for various scenarios for phase 1 and 2 and presents the design study of the NBI.

Predictive simulations of NBI ion power load to the ICRH antenna in Wendelstein 7-X

In Wendelstein 7-X (W7-X), a new ion cyclotron resonance heating (ICRH) antenna will be commissioned during the operational campaign OP2.1. The antenna will have to sustain power loads not only from thermal plasma and radiation but also fast ions. Predictive simulations of fast-ion power loads to the antenna components are therefore important to establish safe operational limits. In this work, the fast-ion power loads from the W7-X neutral beam injection (NBI) system to the ICRH antenna was simulated using the ASCOT suite of codes. Five reference magnetic configurations and five antenna positions were considered to provide an overview of power load behavior under various operating conditions. The NBI power load was found to have an exponential dependence on the antenna insertion depth. Differences between magnetic configurations were significant, with the antenna limiter power load varying between 380 W and 100 kW depending on the configuration. Qualitative differences in power load patterns between configurations were also observed, with the low mirror and low iota configurations exhibiting higher loads to the sensitive antenna straps. The local fast-ion power flux to the antenna limiter was also considered and found to exceed the 2.0 MW m−2 steady-state safety limit only in specific cases. The NBI system might thus pose a safety concern to the ICRH antenna during concurrent NBI-ICRH operation, but additional heat propagation simulations of antenna components are needed to establish more realistic operational time limits.

PFPO plasma scenarios for exploration of long pulse operation in ITER

Long Pulse Scenarios (LPS) in ITER foreseen during the Pre-Fusion Power Operation (PFPO) phase of the ITER Research Plan (IRP) are assessed using 1.5D transport simulations within the ASTRA framework. Such assessment is required to predict the operational space for LPS operation in PFPO, as well as to evaluate which physics processes for LPS operation during Fusion Power Operation (FPO) could be studied during PFPO. An important aspect in the development of LPSs in PFPO is to minimize lifetime consumption of the Central Solenoid (CS) for these scenarios. The maximum pulse length achievable for LPSs in PFPO with no consumption of CS lifetime (currents in CS coils ⩽30 kA per turn) has been assessed for a range of heating schemes and heating mixes, confinement regimes (L-mode and H-mode) and for helium and hydrogen plasmas. The operational space of LPS and pulse length has been explored through density scans with the Heating and Current Drive mix required for the FPO Q ⩾ 5 steady-state plasma scenario (namely Neutral Beam Injection and Electron Cyclotron Heating) including acceptable shine through losses on the first wall for both helium and hydrogen plasmas. Fast particle physics aspects that are common between FPO plasmas and LPS PFPO H-mode plasmas at low densities are studied including MHD stability analysis with the KINX code and non-perturbative critical gradient model based on high-n Toroidal Alfven Eigenmodes (TAE) stability kinetic ballooning code HINST calculations.

Neutronic response of the neutral beam after EAST NBI upgraded

In order to improve the heating effect of neutral beam injector, the neutral beam injector (NBI) of the Experimental Advanced Superconducting Tokamak (EAST) is planned to be upgraded. After the EAST NBI is upgraded, the neutron emission intensity will increase. Neutron leaking from the neutral beam port penetrate the beamline, resulting in the material radiation damage. The neutronic response has been investigated by using the Monte Carlo particle transport code MCNP with the two dimensional RZ model of the EAST NBI. The results demonstrate that a 40% increase of the neutron emission rate has been attained with the upgraded EAST NBI. However, the simulations for the activation of different materials, displacement damage and nuclear heating of the neutral beam still satisfy the requirement of material safety performance. The calculations of nuclear response show that the materials can meet the requirements of the NBI steady state operation after the EAST NBI updated.

A modular multilevel DC active power filter scheme based on optimal parameter design for the CFETR N-NBI accelerator grid power supply

A DC active power filter (DAPF) based on a modular multilevel converter scheme is proposed for the accelerator grid power supply (AGPS) in a neutral beam injector system, to reduce the DC bus voltage ripple of AGPS. It works better than adding capacitors directly to the DC bus. This will help reduce the output voltage ripple of the AGPS at rated operating conditions, improve the focus state of the extraction beam, and reduce the occurrence of breakdown conditions. The scheme is special in that each module can compensate one frequency component of the ripple current, and the maximum switching frequency is not more than 4 times the ripple frequency. The electrical parameters of all DAPF modules (switching states and module capacitor voltages) are optimal values obtained through an optimal parameter design algorithm. In the process of applying the algorithm, two types of APF modules are distinguished, which have different switching frequencies and ripple suppression effects to adapt to different occasions. The simulation results show that each module can reduce the relative frequency component of the ripple by about 80% under the optimal parameters. Similar results are obtained by the 400V AGPS prototype and 20 V DAPF prototype. The ripple reduction scheme can be applied to other high-power applications with DC bus.

Integration of new sets of magnets for improved plasma confinement in the SPIDER experiment

SPIDER is the full-scale prototype of the plasma source of the negative-ion driven neutral beam injector for the heating and current drive of the ITER plasma. The uniqueness and complexity of the system requested this ad hoc test stand aiming at optimizing the performance of the RF inductively generated plasma, negative ion production and extraction, electron filtering, and robustness and controllability of all systems required to work together. After about three years of operation, presently SPIDER is in a long shutdown, in which the whole plasma source and accelerator were dismounted. In this phase, additional modifications with respect to the original design will be introduced to improve the system performance, driven by the experience acquired in the last years.These include the addition of further sets of permanent magnets in the plasma source expansion chamber and around the RF drivers, with the aim of improving the plasma confinement and consequently its density and possibly its uniformity. The present paper reports the study and the analyses behind this modification, which impacts on the original already complex magnetic configuration, made particularly difficult by the limited space available and the high voltages. The use of ferromagnetic shields, necessary to limit stray fields possibly increasing the breakdown probability, make the design particularly complex because of the greater impact on the previous configuration. An iterative process between analyses to determine the ideal configuration and CAD verifications was required. The analyses had to take into account the new magnetic configuration to be created in the particular area of interest, and the overall configuration in order to not compromise its efficacy.

Progress on the conceptual design of the antennas for the DTT ECRH system

One of the main goals of the Divertor Tokamak Test (DTT) facility is to reach a ratio of power crossing the separatrix over the major radius of about 15 MW m − 1, as the one expected in DEMO. For this purpose, up to 45 MW of external additional heating power shall be coupled to the plasma, provided by Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating and Neutral Beam Injection. The foreseen total ECRH power installed at full power shall be 32 MW, generated using 1 MW/170 GHz gyrotrons, for 100 s. The present ECRH system foresees two launching antennas per DTT sector, based on the front-steering concept. The equatorial antenna is dedicated to plasma heating and current drive and the upper antenna to the stabilization of MHD instabilities. This paper focuses on the latest design concept for these two antennas and on the definition of the ex-vessel matching optics unit of the last section of the evacuated Transmission Line (TL). The design has been updated to be compatible with the insertion of CVD diamond windows, to separate the vacuum environment of DTT from the one of the TL. This choice requires adding corrugated waveguide sections between the last mirror of the TL and the first mirror inside the port, requiring some adaptation of the in-vessel optics and of the supporting structure. The possibility to modify the steering range for the launching mirror has been also investigated to be compatible with the new design of the first wall and for the upper antenna, to reach the q = 2 surface in the new plasma scenario.

Design of The Power Supply and Distribution Monitoring System for NNBI

The power supply and distribution system provides powerful support for the efficient operation of the negative ion source neutral beam injector (NNBI) experiment. To centrally manage the power supply equipment of each subsystem of NNBI, a power supply, and distribution monitoring system based on WinCC software and SQL Server database is developed. The system has an excellent man-machine interaction interface and realizes the monitoring and management of energy consumption in the power supply and distribution system. It has circuit process monitoring, fault alarm, operation log, historical data query, and other functions. After testing, the system can run stably and reliably control each device. At the same time, the system has good compatibility and expansibility and meets the needs of intelligent management and monitoring of the power supply and distribution system.

Research and Design of RF Ion Source Impedance Matching Network

The radio frequency (RF) ion source is the key component of the neutral beam injector to achieve high power and long pulse operation. The key technology is feeding RF power from the power source to the ion source and producing stable plasma. It is impossible to calculate the exact equivalent impedance of plasma during discharge, and the impedance will change with the change in discharge pressure and feed power. In order to transfer the maximum output power of the RF power source to the coil antenna of the ion source, an impedance-matching network must be added between the power source and the ion source. Four possible matching network structures are designed by the Smith diagram circle method, and two kinds of capacitor structures are calculated and analyzed. The influence of capacitance change on impedance matching state is analyzed, and the electrical parameters of capacitance under the two structures are compared. After comparison and analysis, the structure of the parallel capacitor is first, and then the series capacitor is selected. Finally, the plasma discharge can be stable through experiments. The research on the matching network is of great significance for the development and experimental operation of the RF feeding system of the high-power RF ion source in the future