Neutral Beam Injection Research Articles

Impact of T i/T e ratio on ion transport based on EAST H-mode plasmas

At the EAST tokamak, the ion temperature (T i) is observed to be clamped around 1.25 keV in electron cyclotron resonance (ECR)-heated plasmas, even at core electron temperatures up to 10 keV (depending on the ECR heating power and the plasma density). This clamping results from the lack of direct ion heating and high levels of turbulence-driven transport. Turbulent transport analysis shows that trapped electron mode and electron temperature gradient-driven modes are the most unstable modes in the core of ECR-heated H-mode plasmas. Nevertheless, recently it was found that the T i/T e ratio can increase further with the fraction of the neutral beam injection (NBI) power, which leads to a higher core ion temperature (T i0). In NBI heating-dominant H-mode plasmas, the ion temperature gradient-driven modes become the most unstable modes. Furthermore, a strong and broad internal transport barrier (ITB) can form at the plasma core in high-power NBI-heated H-mode plasmas when the T i/T e ratio approaches ~1, which results in steep core T e and T i profiles, as well as a peaked n e profile. Power balance analysis shows a weaker T e profile stiffness after the formation of ITBs in the core plasma region, where T i clamping is broken, and the core T i can increase further above 2 keV, which is 80% higher than the value of T i clamping in ECR-heated plasmas. This finding proposes a possible solution to the problem of T i clamping on EAST and demonstrates an advanced operational regime with the formation of a strong and broad ITB for future fusion plasmas dominated by electron heating.

Re-evaluation of shielding ability and induced activity of KSTAR with increased neutron yields

To determine the viability of the 2-fold neutron yield increase of the Korea Superconducting Tokamak Advanced Research (KSTAR) owing to the installation of a second neutral beam injection system to increase the plasma density and a divertor material change from graphite to tungsten to reduce the heat load, we performed a radiological evaluation and shielding analysis of an existing building structure and the KSTAR with an increased neutron yield per shot. According to the shielding analysis results, the weekly and annual doses outside the building did not exceed the criteria, and the fluxes and doses in the plasma experimental room were unaffected by the divertor material change. The maximum dose rate in the working area in the tokamak was 2.16E+03 μSv/h at the surface-plasma-facing components. However, the dose rate was reduced to the background level one month after a 1-shot operation of 300 s. Ar-41 was the major nuclide for the air activation calculations; however, the effluent level exceeded the derived air concentration after the 1-shot operation ended. The residual activity concentration of Ar-41 was reduced below the derived air concentration after 243 min. The annual operating condition of the KSTAR is suggested based on the evaluation results.

Influence of surface work function and neutral gas temperature on mechanism of H− production in negative hydrogen ion sources

Negative hydrogen ion sources (NHISs) based on surface production with cesium (Cs) seeded can fulfill the demanded parameters for neutral beam injection systems for ITER. In this study, the Global Model for Negative Hydrogen Ion Source based on volume-produced H− ions is developed to include surface-produced H− ions and is validated against experimental data obtained in a planar inductively coupled plasma discharge used for study of Cs effect on H− production. The H− density predicted by the model decreases three times with surface work function from 2.1 to 4.5 eV, achieving good agreement with the experimental results, as surface conversion yield of particles to H− ions shows exponential decline with surface work function. The model predicts the rise in neutral gas temperature remarkably enhances surface production but reduces volume production of H− ions, because of increase in surface conversion yield of H atoms to H− ions and in electron temperature, respectively. The dependences of H− production on surface work function and neutral gas temperature are analyzed by evaluating creation rates of the H− ions from different reaction pathways. The developed model can be applied for prediction of H− production in NHISs and ultimate parameter optimization of negative ion beams for fusion reactors.

3D modeling of a double-driver ion source considering ion magnetization: an investigation of plasma symmetry modulation methods

A three-dimensional fluid model of a double-driver negative hydrogen ion source for China Fusion Engineering Test Reactor (CFETR) neutral beam injection is developed. In this model, the magnetic filter field is generated by 16 permanent magnets, which are surrounded by a soft iron. In order to accurately describe the transportation of charged species in the presence of strong magnetic field, both the electron magnetization and ion magnetization are taken into account, and the accuracy of the model has been proved by comparison with experimental data. By employing this model, the spatial distributions of the plasma parameters have been investigated, and three methods are proposed to optimize the symmetry at the bottom of the expansion region of a double-driver source. The results indicate that by adjusting the power of Driver I while keeping the power of Driver II constant, the symmetry of the electron density and negative hydrogen ion density could be improved. Furthermore, the inclusion of partition improves the symmetry of the electron temperature and density but has no impact on the regulation of the negative hydrogen ion density distribution. Finally, the application of magnetic shield can not only improve the symmetry of the electron density and negative hydrogen ion density, but also increase their densities at the bottom of the expansion region.

Real-time control of NBI fast ions, current-drive and heating properties

Conventionally, neutral beam injection (NBI) in tokamaks is controlled via engineering parameters such as injection voltage and power. Recently, the high-fidelity real-time NBI code RABBIT has been coupled to the discharge control system of ASDEX Upgrade. It allows to calculate the NBI fast-ion distribution and hence the properties of NBI in real-time, making it possible to control them directly. We successfully demonstrate control of driven current, ion heating and stored fast-ion energy by modifying the injected beam power. A combined ECRH and NBI controller is also successfully tested, which is able to adjust the heating mix between ECRH and NBI to match a certain desired ion heating fraction at given total power. Further experiments have been carried out towards control of the ion heat flux (i.e. ion heating plus collisional heat transfer between ions and electrons). They show good initial success, but also leave room for future improvements as the controller runs into instabilities at too high requests.

Design of a Plasma neutraliser for a Fusion reactor or as an upgrade to the ITER heating neutral beam injectors

The global efficiency of neutral beam injectors to be used on a fusion reactor needs to be of the order of, or greater than, 60 %, which is considerably higher than the efficiency of presently operating NBIs. The required efficiency can only be achieved by increasing the neutralisation efficiency from the ≈56 % obtained with the gas neutralisers currently used, which can be done using a plasma or photon neutraliser. This paper examines possible designs of a plasma neutraliser taking into account the many constraints imposed by operation in the environment around a fusion reactor, and a schematic design is proposed that meets those constraints. The design chosen uses a beam driven plasma neutraliser, which requires no additional power systems, diagnostics or interlocks.Using a plasma neutraliser is shown to improve the global efficiency to ≈55 %. Although that is below the desired value, reasons why that might be acceptable are discussed.This paper also discusses the possible use and benefits of such a neutraliser as an upgrade of the ITER heating neutral beam system.

NBI optimization on SMART and implications for scenario development

The SMall Aspect Ratio Tokamak (SMART) under commissioning at the University of Seville, Spain, aims to explore confinement properties and possible advantages in confinement for compact/spherical tokamaks operating at negative vs. positive triangularity. This work explores the benefits of auxiliary heating through Neutral Beam Injection (NBI) for SMART scenarios beyond the initial Ohmic phase of operations, in support of the device’s mission. Expected values of electron and ion temperature achievable with NBI heating are first predicted for the current flat-top phase, including modeling to optimize the NBI injection geometry to maximize NBI absorption and minimize losses for a given equilibrium. Simulations are then extended for a selected case to cover the current ramp-up phase. Differences with results obtained for the flat-top phase indicate the importance of determining the plasma evolution over time, as well as self-consistently determining the edge plasma parameters for reliable time-dependent simulations. Initial simulation results indicate the advantage of auxiliary NBI heating to achieve nearly double values of pressure and stored energy compared to Ohmic discharges, thus significantly increasing the device’s performance. The scenarios developed in this work will also contribute to diagnostic development and optimization for SMART, as well as providing test cases for initial predictions of macro- and micro-instabilities.

A neoclassical validation of balanced and unbalanced rotations on EAST H-mode discharges

Predicting residual stress (RS) contribution to intrinsic rotation is one of the major challenges in the study of momentum transport in tokamaks. One efficient experimental means of quantifying RS torque magnitude is to generate radially-flat and near-zero rotation profiles, termed ‘balanced rotations’ in this work, using counter-Ip Neutral Beam Injections (NBIs) to effectively cancel the torques from co-Ip NBIs. One remaining question, however, is on whether or not the attained velocity profile is well zeroed and flat enough so that the predicted RS torques based on perfect balance assumption can be used for further studies such as fitting of diffusive and convective coefficients to match with experiments. This article presents a neoclassical means of validating the attained balanced and unbalanced rotations at EAST to consequently validate the predicted RS torque profiles, using TRANSP/NUBEAM and a recently-developed neoclassical rotation/transport code TransROTA (Bae et al 2024 Comput. Phys. Commun. 296 108992). Both balanced and unbalanced EAST H-mode discharges are analyzed to find that the suggested neoclassical validation methodology successfully validates attained balanced rotations. It also finds that neoclassical gyroviscous (NGV) torque serves as the balance-breaking mechanism for axisymmetric plasmas and the balance-breakings start from the core and propagate towards the edge. This work also suggests the possibility of using the neoclassical methodology to find locally-balanced rotations at an elevated velocity range near ∼20 km s−1, which implies possible scaling of RS torques up to a certain elevated velocity range.

Investigating the role of edge neutrals in exciting tearing mode activity and achieving flat temperature profiles in LTX-β

We present observations, numerical simulations, and analysis from experiments in the Lithium Tokamak Experiment-Beta (LTX-β) in which the electron temperature profile (Te (r)) shifts from flat to peaked and a tearing mode is also destabilized when the average density (ne ave ) exceeds ∼1019 m−3. Flat Te (r) is obtained routinely in LTX-β, with a lithium coated, low-recycling first wall, once the external fueling is stopped and density decays [Boyle et al 2023 Nucl. Fusion 63 056020]. In the present experiment, flat Te profiles can be sustained while maintaining constant ne ave below a line averaged density threshold (ne ave th ) of ∼1019 m−3. Above ne ave th, Te (r) shifts from flat to peaked and a tearing mode is destabilized. Due to low recycling, the achieved ne ave can be controlled precisely by external fueling and hence, a certain threshold of the edge neutral inventory from the external fueling is experimentally manifested through ne ave th . The goal of the present work is to investigate the role of edge neutrals in determining Te (r) and MHD stability in the unique low-recycling regime of LTX-β. Our hypothesis is that the peaking of Te (r) beyond ne ave th is due ultimately to the edge cooling by the cold neutrals beyond a critical fueling flux. At lower fueling flux, flat Te (r) results in broader pressure profile and lower resistivity, which in turn stabilizes the tearing mode. This hypothesis is supported by edge neutral density estimation by DEGAS 2 code. Mode analysis by singular value decomposition confirms the tearing mode structure to be m/n = 2/1 (m and n being the poloidal and toroidal mode numbers). Linear tearing stability analysis with M3D-C1 predicts that plasmas with ne ave > 1019 are highly susceptible to a n = 1 tearing mode. ORBIT simulations, however, confirmed that the tearing modes do not contribute to the loss of fast ions from neutral beam injection. This study shows for the first time that the neutral inventory at the edge could be one of the deciding factors for the achievability of the unique operation regime of flat Te (r) and the excitation of tearing activity that could be disruptive for the plasmas.

Progress of Lyman-alpha-based beam emission spectroscopy (LyBES) diagnostic on the HL-2A tokamak

An edge Lyman-alpha-based beam emission spectroscopy (LyBES) diagnostic, using a heating NBI (neutral beam injection) system, is currently under development on the HL-2A tokamak. The 20-channel edge LyBES, which is intended to measure the density fluctuation in plasma edge (from R = 1960 mm to R = 2026 mm) with an improved spatial resolution of 3.3 mm, is a complement to the existing conventional beam emission spectroscopy (BES) diagnostic. In this article, we introduce the progress of LyBES diagnostic, including the collection optics, the monochromator, and the detector system. The reflectance of the collection mirrors is measured to be ~82% at 122 nm, and the aberration geometrical radius of the collection optics is tested to be ~150 μm in the aimed area. The linear dispersion of the LyBES monochromator is designed to be ~0.09 nm mm−1. The bandwidth of the detector system with the 5×107 V A−1 preamplifier gain is measured to be ~280 kHz, and the peak-to-peak noise of the detector system is tested to be ~16 mV. The finalized design, components development and testing of the LyBES diagnostic have been completed at present, and an overall performance of the LyBES diagnostic is to be confirmed in the next HL-2A campaign.

Criterion for long sustained highly peaked ion temperature in diverted configuration of KSTAR tokamak

In the context of KSTAR plasma research, the discovery of the fast ion regulated enhanced mode is noteworthy due to its remarkable ability to maintain ion temperatures exceeding up to 10 keV for a few tens of seconds, avoid impurity accumulation, and keep low loop voltages. This new plasma operating scenario is achieved in a diverted configuration plasma by avoiding the H-mode transitions with sufficient additional power for the transition. Keeping the density low is the primary method for the avoidance. Additionally, adjustments to other parameters (plasma shape, neutral beam injection, and toroidal magnetic field) associated with the H-mode threshold power are applied to inhibit the transition process. This paper includes an experimental analysis and discussion of these findings.

Particle injection methods in 3D-PIC MCC simulations applied to plasma grid biasing

In negative ion sources for the ITER Neutral Beam Injection system, the co-extraction of electrons is one of the main limiting factors. The current of co-extracted electrons can be decreased by applying a positive bias voltage to the Plasma Grid (PG) with respect to its source walls. Simulations using three-dimensional Particle-in-Cell Monte Carlo Collision (3D-PIC MCC) model are a powerful tool for studying the extraction region of such ion sources. However, the inclusion of both PG and source walls in the simulation domain is difficult due to numerical constraints. This study uses the 3D-PIC MCC code ONIX to explore the effects of particle injection models on plasma characteristics, using a flux injection model to regulate particle influx for a flat transition in potential from the bulk plasma to the simulation domain. Biasing of the PG above floating potential is possible using the flux injection scheme and results in a notable reduction in co-extracted electrons, corroborating with established experimental observations.

Isotope effects on transport characteristics of edge and core plasmas heated by neutral beam injection (NBI) in an inward shifted configuration at the Large Helical Device

Isotope effects have been investigated in Neutral Beam Injection (NBI) heated plasmas on the Large Helical Device with similar operational parameters between Hydrogen (H) and Deuterium (D) plasmas. Experimental results show that the global energy confinement has no significant dependence on the isotope mass under similar discharge conditions with nearly the same heating power, line-averaged density ( nˉe ) and magnetic field. For both electron and ion energy transport, the transport coefficients, which are obtained based on local power balance analysis, have analogous profiles between H and D dominant plasmas. For neoclassical χe and χi values, they are almost equal between H and D dominant plasmas in low nˉe discharges, whereas in high nˉe cases they are lower in H plasmas than those in D ones. At low nˉe , the electron and ion thermal transport in both H and D plasmas are dominated by neoclassical transport at a certain zone ( ρ≈0.6−0.85 ), while the anomalous transport process has primary effects in the remaining area, and the density fluctuations exhibit ion temperature gradient mode nature. With increase of nˉe , the anomalous transport becomes prevailing and the density fluctuations propagate along electron diamagnetic drift direction. Bispectral analysis reveals that the H plasma has stronger nonlinear coupling in edge density fluctuations in both low and high density discharges, which is probably due to that the D plasma has stronger damping rate for the nonlinear interaction of turbulence. For a comparative study, the present results have been compared with those observed in the ECRH discharges (Tanaka et al 2019 Nucl. Fusion 59 126040). The reasons for the similarities and dissimilarities between these two different heating manners are not clear yet. To unravel the underlying physics, essential inputs from theories and simulations are required.

Soft H-L back transitions by applying resonant magnetic perturbations in the low q 95 EAST plasmas

Controlled ‘soft’ H-L transitions with a simultaneous ramp-down of plasma density and stored energy, generated by applying resonant magnetic perturbations (n = 2 RMPs), relevant to the control of ITER terminations, are achieved in the EAST tokamak. These H-mode plasmas are run at median electron densities and a low safety factor q 95 ( q95=3.5 –4.0) under neutral beam injection (NBI) dominant heating. During a slow ramp in the RMP current the internal transport barrier for the electron density collapses at the beginning and the edge transport barrier for the electron density collapses later. Before the controlled H-L back transitions, the plasma stored energy W MHD is linearly correlated with the core line-averaged density ⟨ne⟩ . These controlled H-L back transitions ensue from the collapse of the edge transport barrier and at a similar energy confinement time τeH-L∼48 ms. These H-L transitions are not sensitive to the NBI power but are very sensitive to the fueling rate and ramp-up rate of the RMP current.

β-induced Alfvén eigenmodes and fishbone driven by energetic electrons in EAST electron heating dominant plasmas

β-induced Alfvén eigenmodes (e-BAE) and fishbone (e-fishbone) driven by the energetic electrons have been identified on the experimental advanced superconducting tokamak in electron cyclotron resonance heating (ECRH), low hybrid wave (LHW) together with neutral beam injection plasma and in pure radio frequency heating plasma, respectively. The e-BAE is predominantly located near the pedestal region, and its frequency typically displays a chirping characteristic. The frequency is notably associated with the count of energetic electrons with energy between 30 keV and 80 keV driven by LHW and undergoes three distinct phases. A direct correlation is established between the frequency of e-BAE and the global parameters, including electron temperature, Alfvén velocity and βp . The e-fishbone emerges in the core and becomes noticeable shortly after ECRH initiation, in an electron heated dominant and core confinement enhanced discharge. Experimental analyses suggest that the e-fishbone ties closely with both the population and energy distribution of energetic electrons. These observations can provide valuable insights into the behavior of energetic electrons, which play a crucial role in the formation and sustainability of transport barriers within hybrid plasma, both in the internal and pedestal regions.

Characterization of the TJ-II stellarator plasma by means of reciprocating Langmuir probes

Experiments were conducted in deuterium plasma in the TJ-II stellarator by means of swept Langmuir probes mounted on reciprocating probes manipulators. The results were processed using the four-parameter fit, as well as the triple-probe and the first-derivative probe techniques. The parameters determined were the floating potential, the ion saturation current density, the electron temperature and density, and the plasma potential. The results were obtained for two plasma heating techniques – electron cyclotron resonance heating (ECRH) and neutral beam injection (NBI) heating. In the case of ECRH, employing the first-derivative probe technique resulted in finding that the electron-energy distribution function (EEDF) was not Maxwellian, but rather a bi-Maxwellian one with thermal (14-25 eV) and cold (4-5 eV) electrons. In comparison, during NBI heating we found a Maxwellian EEDF with the electron temperature being around 5 eV and slightly increasing in the confined plasma, but always remaining below 15 eV. We present a detailed analysis and discussion of the data for the plasma parameters as acquired by different techniques of using the reciprocating probe manipulator.

The impact of neutral beam parameters on current drive and neutron yield in DEMO-FNS tokamak

DEMO-FNS (DEMO Fusion Neutron Source) will be a hybrid reactor designed to combine fusion and fission technologies. In a hybrid reactor the power is mainly produced by a fission blanket which is exposed to the neutron flux generated by confined fusion plasma used as a Fusion Neutron Source (FNS). The steady-state operation of the FNS is easier to achieve as compared to pure fusion reactors, allowing the FNS to be more compact. The DEMO-FNS will produce 40 MW of fusion power. The neutral beam injectors (NBI) are designed to provide a steady state plasma heating, rotation, fueling, current drive and neutron generation in DEMO-FNS. Four injector units will deliver 30 MW power in deuterium with energy Eb = 500 keV. The NBI concept and main components are inspired from ITER HNBI design implementing acceleration of negative ions with their neutralization on gas.The high density of injected power in DEMO-FNS leads to a very specific operation scenario with ahigher ratio of high-energy ions in the plasma with respect to conventional tokamaks, where the mainpart of neutron flux will result from fusion between hot-tail ions and relatively coldbackground. The current drive and neutron yield produced by the tangentially injected beam in asteady-state operation are strongly affected by magnetic configuration and plasma kineticprofiles. The fast ions deposition and performance in plasma are limited by the losses associatedwith the neutral beam shape and aiming. The main channels of direct beam losses includeshine-through losses, as well as fast ion orbital losses. The detailed evaluation of NB parametersand geometry effects is a necessary step of scenario optimization.BTR code (Beam Transmission or, in earlier versions, Beam Transmission with Reionization) is intended for injector beamlines study, and BTR detailed beam model can be also applied to perform a detailed analysis of beam losses and performance in plasma. The analysis of beam in plasma shows the ways towards higher values of beam driven current (NBCD) and neutral yield; the beam-in-plasma behaviour can be tuned for a given tokamak configuration. The NB tracing model workflow is implemented in the BTOR suite (Beam in TORoids). The results for DEMO-FNS are compared with the NBI performance in ITER tokamak, as both machines are beam-driven and implement a similar NBI concept.

Commissioning of the imaging neutral particle analyser for the ASDEX Upgrade tokamak

An imaging neutral particle analyser (INPA) diagnostic has been installed and commissioned at the ASDEX Upgrade (AUG) tokamak. The AUG INPA diagnostic provides energy and radially resolved measurements of the fast-ion (FI) distribution, complementing the existing set of diagnostic which measure the confined FI population. To this end, it analyses charge exchange (CX) neutrals produced in reactions between FI and neutrals injected by a neutral beam injector. These CX neutrals are ionised by a 20 nm carbon foil and deflected towards a scintillator by the machine magnetic field. The use of a scintillator as active component allows us to cover the whole plasma radial range with an energy resolution of 9 keV and a spatial of 3 cm for 93 keV deuterons. First measurements demonstrate the high sensitivity of the INPA diagnostic to different AUG fast-ion distribution functions, from NBI and ion-cyclotron resonance heating origin, and show good agreement with the synthetic diagnostic.

Development of a neutron yield measurement system utilizing BF_3 detectors on the HL-3 tokamak

A neutron yield measurement system, comprising five BF3 proportional counters with excellent n/γ discrimination capabilities, has been meticulously developed and implemented on the HL-3 tokamak to provide time-resolved assessments of neutron emission rates. The size of moderator, optimized for the flat energy responses, was skillfully simulated and designed using the MCNP code. An in-depth analysis of measurement errors stemming from pile-up phenomena determined that the upper limit of the count rate, under the current electronics design, is 131 kcps with a relative error of less than 20%. Drawing from the pulse amplitude spectrum obtained from a 252Cf neutron source and experimentation on the HL-3 tokamak, the amplitude threshold for the BF3 neutron detector systems was judiciously established. Furthermore, it was conclusively affirmed that the neutron yield measurement system possesses the capability to effectively discriminate between neutron and gamma signals. During HL-3 discharges, deuterium-deuterium (DD) neutrons were reliably detected with a temporal resolution of 10 ms, and the variations in neutron count rates was found to be in alignment with the neutral beam injection. Ultimately, the total neutron yield was successfully estimated through the utilization of count rates from four BF3 detectors, thereby validating the robust performance of the neutron yield measurement system.

Overview of neutral beam injectors for plasma heating and diagnostics developed at Budker INP

An overview of the neutral beam injectors developed at the Budker Institute of Nuclear Physics in Novosibirsk during the last 10 years is presented. These neutral injectors are used for plasma diagnostics, heating and current drive in modern fusion devices with magnetic confinement. An arc or a radio-frequency (RF) discharge generates a plasma in the ion sources of the injectors, and a positive hydrogen or deuterium ion beam is extracted and accelerated by a multiaperture ion-optical system (IOS). The accelerated ion beam is converted into a neutral one in a gas target. The precision multiaperture IOS with spherically concave electrodes provides ballistic focusing of the neutral beam. The high-energy, high-power beam injector based on negative ions, which is currently under development, is described as well. It comprises a RF negative ion source and a wide-aperture electrostatic accelerator separated from the source by a low-energy beam transport line, thereby improving the injector reliability.