Neutral Beam Injection Power Research Articles

A numerical method for calculating the driven current of neutral beam injection in tokamaks

The current driven by neutral beam injection in tokamak is calculated, and the slowing-down distribution function of the fast ion is obtained by the backward Euler iteration method, including the pitch angle scattering collision. This study reveals that when the pitch-angle cosine is small, the trapped fast-ion current significantly contributes to the total driven current, particularly when the neutral beam is injected perpendicularly. In such cases, the current densities of passing and trapped ions are of the same order of magnitude, with the trapped fast-ion current contributing over 10% to the total neutral beam-driven current. This results in a parabolic profile of the total current in the radial direction, promoting the formation of a negative shear equilibrium structure in the core of the tokamak plasma. The numerical approach was validated against the NUBEAM code while considering electron shielding effects and applied to calculate the neutral beam-driven current in multiple tokamaks. The influence of pitch-angle cosine and neutral beam injection power on the driven current was studied at different radial positions.

Enhanced plasma performance in C-2W advanced beam-driven field-reversed configuration experiments

TAE Technologies’ fifth-generation fusion device, C-2W (also called ‘Norman’), is the world’s largest compact-toroid device and has made significant progress in field-reversed configuration (FRC) plasma performance. C-2W produces record breaking, macroscopically stable, high-temperature advanced beam-driven FRC plasmas, dominated by injected fast particles and sustained in steady state, which is primarily limited by neutral-beam (NB) pulse duration. The NB power supply system has recently been upgraded to extend the pulse length from 30 ms to 40 ms, which allows for a longer plasma lifetime and thus better characterization and further enhancement of FRC performance. An active plasma control system is routinely used in C-2W to produce consistent FRC performance as well as for reliable machine operations using magnet coils, edge-biasing electrodes, gas injection and tunable-energy NBs. Google’s machine learning framework for experimental optimization has also been routinely used to enhance plasma performance. Dedicated plasma optimization experimental campaigns, particularly focused on the external magnetic field profile and NB injection (NBI) optimizations, have produced a superior FRC plasma performance; for instance, achieving a total plasma energy of ∼13 kJ, a trapped poloidal magnetic flux of ∼16 mWb (based on the rigid-rotor model) and plasma sustainment in steady state up to ∼40 ms. Furthermore, under some operating conditions, the electron temperature of FRC plasmas at a quiescent phase has successfully reached up to ∼1 keV at the peak inside the FRC separatrix for the first time. The overall FRC performance is well correlated with the NB and edge-biasing systems, where higher total plasma energy is obtained with higher NBI power and applied voltage on biasing electrodes. C-2W operations have now reached a mature level where the machine can produce hot, stable, long-lived, and repeatable plasmas in a well-controlled manner.

Conceptional design of photoneutralization test system for negative ion-based neutral beam injection

Neutral beam injection is one of the effective heating methods in the field of magnetic confinement fusion, and neutralization is the most crucial link in the case of negative ions. To further increase the neutral beam injection power, improve the long pulse operation capability, and optimize the efficiency of the NNBI system, further research and verification about the neutralization mode are needed. Theoretically, photoneutralization can achieve more than 90% neutralization efficiency. However, maintaining stable operation of the megawatt laser cavity over extended periods poses corresponding challenges. Additionally, the cost associated with laser target surpasses the benefit gained from increased neutralization efficiency, leading to its lack of practical application thus far. This paper proposes a solution to these issues by designing a single-channel, multi-fold photoneutralization verification system utilizing the CRAFT NNBI one-quarter and one-half size negative source test equipment. An outline of the system's test and diagnostics approach is provided. Key parameters such as laser target thickness, negative ion energy, beam shape and efficiency of the photoneutralization system are numerically calculated. Combined with the experimental data of the negative source test platform, theoretical calculations show that the neutralization efficiency can achieve 63% with the system efficiency exceeding 40%. Even by increasing the incident laser power or the number of reflections, neutralization efficiency can be increased to 95%, with a simultaneous increase in system efficiency to 60%. Maintaining efficiency while increasing incident laser power could reduce the number of reflections to approximately ten, reaching an acceptable threshold. However, this adjustment will increase the irradiation density of a single mirror from 660W/mm2 increases to 3000W/mm2. This paper methodically designs a practical laser neutralization verification platform, which is expected to substantially improve the neutralization efficiency, and facilitate practical application and validation.

Characterisation of the scrape-off layer in JET-ILW deuterium and helium low-confinement mode plasmas

Langmuir probe measurements in neutral beam injection (NBI) heated, low-confinement mode plasmas in JET ITER-like wall showed that the current to the divertor targets, Idiv, in helium (He) plasmas was up to 70% lower on the low-field side (LFS) than in otherwise identical deuterium (D) plasmas. The edge plasma density at which the rollover of Idiv occurred i.e. the onset of detachment, was 10% higher in He plasmas on both the LFS and high-field side (HFS). The density of Idiv rollover increases by 25% for He when the NBI power increases 1MW to 5MW. The total radiated power was similar in He and D plasmas for densities below the Idiv rollover. At densities above the Idiv rollover density, the total radiated power and power from within the separatrix are higher in He, reducing the power across the separatrix and subsequently Idiv,LFS. In He plasmas, the peak radiated power was observed within the confined region above the X-point in tomographic reconstructions from bolometry.

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.

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.

Studying fast-ion populations using oscillations in solid-state neutral-particle analyzer signal and neutral-beam injection power.

A method for determining the fast-ion population density in magnetically confined plasmas as a function of pitch-radius, (λ, R), using a solid-state neutral-particle analyzer (ssNPA) signal and neutral-beam injection (NBI) power-output data has been developed. Oscillations in the NBI power output are replicated only in the active part of the ssNPA signal, allowing this to be separated from the passive and background signals, which usually complicate data from this diagnostic. Results obtained using this method are compared with those from standard techniques using data from the Mega-Amp Spherical Tokamak Upgrade spherical tokamak.

Tritium neutral beam injection on JET: calibration and plasma measurements of stored energy

Neutral beam injection (NBI) is a flexible auxiliary heating method for tokamak plasmas, capable of being efficiently coupled to the various plasma configurations required in the Tritium and mixed deuterium-tritium experimental campaign on the Joint European Torus (JET) device. High NBI power was required for high fusion yield and alpha particle studies and to provide mixed deuterium-tritium (D-T) fuelling in the plasma core, it was necessary to operate the JET NBI systems in both deuterium and tritium. Further, the pure tritium experiments performed required T NBI for high isotopic purity and reduced 14 MeV neutron yield. Accurate power calibrations are also essential to machine safety. Previously on JET there have been a number of questions raised on the NBI power calibration, in particular following the Trace Tritium Experiments (TTEs). Operator activities on the tokamak NBI system, including calibrations, were performed in 2020. Following these activities, a series of plasma experiments were devised to further corroborate the T NBI power by comparing the plasma response to the D NBI power. A series of stationary, L-mode plasmas were performed on JET with different beam combinations used in different phases of the same pulse. By comparing the plasma response for D and T NBI it was possible to corroborate the T NBI power calibration using the D NBI power calibration. The stored energy as measured by magnetic diagnostics, corrected for fast particle stored energy, show that the uncertainty in NBI power calibration in T is comparable to that in D.

Toroidal Alfvén eigenmodes observed in low power JET deuterium–tritium plasmas

The Joint European Torus recently carried out an experimental campaign using a plasma consisting of both deuterium (D) and tritium (T). We observed a high-frequency mode using a reflectometer and an interferometer in a D-T plasma heated with low power neutral beam injection, . This mode was observed at a frequency and was located at major radii . The observed mode was identified as a toroidal Alfvén eigenmode (TAE) using the linear MHD code, MISHKA. Beam ions and fusion-born alpha particles were modelled using the full orbit particle tracking code LOCUST, which produces smooth distribution functions suitable for stability calculations without analytical fits or the use of moments. We calculated the stability of the 21 candidate modes using the HALO code. These calculations revealed that beam ions can drive TAEs with toroidal mode numbers with linear growth rates , while TAEs with n < 8 are damped by the beam ion population. Alpha particles drive modes with significantly smaller linear growth rates, due to the low alpha power generated almost exclusively by beam-thermal fusion reactions. Non-ideal effects were calculated using complex resistivity in the CASTOR code, leading to an assessment of radiative, collisional, and continuum damping for all 21 candidate modes. Ion Landau damping was modelled using Maxwellian distribution functions for bulk D and T ions in HALO. Radiative damping, the dominant bulk damping mechanism, suppresses modes with high toroidal mode numbers. Comparing the drive from energetic particles with damping from thermal particles, we find all but one of the candidate modes are damped. The single net-driven n = 9 TAE with a net growth rate matches experimental observations with a lab frequency and location . The TAE was driven by co-passing particles through the resonance. Both co- and counter-passing alpha particles drive the TAE through the resonance. Additional sideband resonances contribute significant drive for both beam and alpha particles.

Hot spots of the main limiter induced by fast ions in experimental advanced superconducting tokamak (EAST)

The main limiter in EAST was observed to endure a high heat load and was cracked near the midplane at the right side during the plasma operation. To explore the heat load carried by fast ion loss toward the main limiter, the neutral beam injection (NBI) and radio frequency power proportion experiment was conducted in EAST where the plasma stored energy and line integrated density were kept almost constant. The hot spot at the right side of the main limiter was observed to be enhanced by co-current perpendicular (co-perp) NBI. An NBI ion loss simulation was performed in the presence of the toroidal field ripple and the Coulomb collision by using the orbit code GYCAVA and the NBI code TGCO. The result indicates that the NBI ion loss by ripple and collision mainly causes a bright area below the midplane of the right side of the main limiter as observed in the EAST experiment. The peak heat load of lost fast ions generated by co-perp NBI is ∼0.5 MW/m2 as obtained by 1 MW of NBI deposited power and comparable with the heat load carried by fast electrons induced by lower hybrid current drive. In addition, increasing the gap between the separatrix and the first-wall limiters in the simulation is found to reduce this heat load.

Preliminary results of the 105GHz collective Thomson scattering system on HL-2A.

A 105GHz collective Thomson scattering (CTS) diagnostic has been successfully developed for fast-ion measurements on the HL-2A tokamak, and it has been deployed during an experimental campaign. Enhanced signals exhibiting synchronous modulation characteristics have been observed across all CTS channels upon the launch of a modulated probe wave. Results show that the intensity of the CTS signal increases with Neutral Beam Injection (NBI) power and is proportional to neutron count, indicating that the scattering signal contains a contribution from fast ions. Compared with the signal without NBI, the enhanced scattering spectrum due to NBI is slightly wider than the predicted fast ion range. Such broadening might be attributed to the heating effects of the gyrotron.

Impact of toroidal rotation on the resistive ballooning modes in ASDEX Upgrade tokamak

In this work, we investigate the behavior of instabilities appearing between type-I edge localized modes (ELMs), with increasing neutral beam injection (NBI) power concomitant increase in toroidal rotation, and compare it to the modeling result of the linear magneto-hydrodynamic (MHD) code CASTOR3D. An injection of one NBI beam, increasing toroidal rotation, results in the mode slowing down from 12 kHz to 7 kHz, and its associated radial displacement decreases from 5 mm to 3.5 mm. In addition, modes shift radially outwards towards higher q, decreasing their poloidal mode numbers. The mode velocity is measured to be close to the E × B velocity with significant uncertainties. Through a set of CASTOR3D simulations with varying profiles, resistivity has been identified as the primary contributor to the growth rates. Only a small stabilizing effect due to toroidal rotation has been observed. While experimental results show a decrease of mode frequency with rotation, the opposite trend is observed in modeling. Reasons for discrepancies between modeling and experiment are discussed. Nevertheless, a main contributor to the mode frequency has been identified to be rotation velocity. CASTOR3D classifies modes as resistive ballooning modes as they do not appear unstable in ideal MHD.

Divertor detachment and reattachment with mixed impurity seeding on ASDEX Upgrade

Using newly developed spectroscopic models to measure the divertor concentration of Ne and Ar, it is shown that the experimental detachment threshold on ASDEX Upgrade with Ar-only and mixtures of Ar+N or Ne+N scales as expected in comparison with an analytical equation derived by Kallenbach et al (2016 Plasma Phys. Control. Fusion 58 045013). However, it is found that Ar radiates more efficiently and Ne less efficiently in the scrape-off layer than the model predicts. By separately increasing the neutral beam injection power and cutting the impurity gas flow, it is shown that the partially detached and strongly detached X-point radiator scenarios reattach in ≈100 ms and ≈250 ms, respectively. The former timescale is set by the core energy confinement time, whereas the latter has an additional delay caused by the time required for the ionisation front to move from the X-point to the target. A simple equation with scalable geometric terms to predict the ionisation front movement time in future machines is proposed.

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.

Tritium Operation of the JET Neutral Beam Systems and Tritium NBI Power Calculations

Neutral beam injection (NBI) is a very flexible auxiliary heating method for tokamak plasmas, capable of being efficiently coupled to various plasma configurations required in the tritium and deuterium–tritium experimental campaign (DTE2) to be undertaken in the Joint European Torus (JET) device. In particular, experiments for high fusion yield and alpha particle studies require high-power NBI heating, and for maximum performance and optimum fuel mixture control in deuterium–tritium (D–T) plasmas, it is necessary to operate the JET NBI systems in both deuterium and tritium. The technical aspects of the JET NBI systems for compatibility with T operation are discussed, and the associated commissioning is described. The characterization of the JET NBI system in the tritium gas mode will be presented, with particular focus on the power and species mix measurements; this will be the first time that such data have been collected and analyzed for tritium neutral beams. Deuterium operation in the tritium gas mode was successfully carried out in 2019 with no loss in reliability. In this period of operation, the NBI power has been measured using beamline diagnostics and corroborated with plasma measurements. The species mix of the beam has been measured on the neutral beam test bed (NBTB) and also corroborated by plasma diagnostics on JET. These results will be presented alongside tritium NBI results allowing comparison of possible JET NBI performance between deuterium and tritium. Measurements of the NBI power in tritium show that there will be a higher neutralized fraction than in deuterium and a higher full energy fraction. When the effect of particle mass is also accounted for, this will lead overall to reduced beam penetration and lower particle flux per MW.

Electron cyclotron current drive under neutral beam injection on HL-2M

Based on OMFIT framework and HL-2M parameters, this paper comprehensively considers the changes in plasma density, temperature, and other transport quantities caused by the interaction of neutral beam injection (NBI) and electron cyclotron wave (ECW) with plasma. The changes in the Shafranov shift of the plasma magnetic surface center are also evaluated. Theoretically, the influence of NBI on the deposition location and current drive efficiency of the ECW is studied. According to the findings, NBI affected the position location and efficiency of the electron cyclotron current drive (ECCD) deposited on both high field side (HFS) and low field side (LFS). NBI can relocate the ECW power deposition location to the core and increase the current drive efficiency when the ECW power is deposited on the LFS. When the NBI power increases to 7 MW, the ECCD deposition location can shift to the core by roughly 0.15 normalized small radii, and the current drive efficiency can be improved by 1.3 times. Moreover, as NBI power increases, the radial region where the dimensionless current drive efficiency equals to zero gets closer to the plasma edge. When ECW power is deposited on the HFS paraxial, increasing NBI power causes the ECW deposition location to move toward the plasma edge, thus lowering current drive efficiency. This trend is caused by an increase in NBI power, which can increase the Shafranov shift of the plasma center, increase the electron density, and change the electron temperature. These studies hold great significance for achieving more effective current drive and controlling the plasma current profile and neoclassical tearing mode instability.

Design optimization of slit aperture Faraday shield for RF ion source in CRAFT NNBI system

The Faraday shield (FS) is a key part of the radio frequency (RF) ion source for high power neutral beam injection system. The RF transfer efficiency and power loss ratio of the driver with two different types of FS are analyzed using a finite element model. These two types of FS are divided into inclined shape slits and Z shape slits, and each slit shape can be subdivided into six types. It is found that the change of the slit shape on the FS has little effect on the RF transfer efficiency and the power loss ratio. For the six drivers with inclined shape FS slit, the difference in RF efficiency between them does not exceed 1%, and the difference in the power loss ratio on the FS does not exceed 1.2%; for the six drivers with Z shape FS slit, the change in the RF efficiency does not exceed 0.3%, and the change in power loss ratio on the FS does not exceed 0.4%. Meanwhile, taking electrons as the research object, the interception ability of 12 kinds of FS to particles was compared, and the optimal structure of the FS slit is determined.

Enabling adaptive pedestals in predictive transport simulations using neural networks

We present PEdestal Neural Network (PENN) as a machine learning model for tokamak pedestal predictions. Here, the model is trained using the EUROfusion JET pedestal database to predict the electron pedestal temperature and density from a set of global engineering and plasma parameters. Results show that PENN makes accurate predictions on the test set of the database, with R 2 = 0.93 for the temperature, and R 2 = 0.91 for the density. To demonstrate the applicability of the model, PENN is employed in the European transport simulator (ETS) to provide boundary conditions for the core of the plasma. In a case example in the ETS with varied neutral beam injection (NBI) power, results show that the model is consistent with previous studies regarding NBI power dependency on the pedestal. Additionally, we show how an uncertainty estimation method can be used to interpret the reliability of the predictions. Future work includes further analysis of how pedestal models, such as PENN, or other advanced deep learning models, can be more efficiently implemented in integrating modeling frameworks, and also how similar models may be generalized with respect to other tokamaks and future device scenarios.

NSTX-U theory, modeling and analysis results

The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 s, and has available up to 15 MW of neutral beam injection power at different tangency radii and 6 MW of high harmonic fast wave heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, EP stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

NSTX-U theory, modeling and analysis results

Abstract The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 seconds, and has available up to 15 MW of Neutral Beam Injection (NBI) power at different tangency radii and 6 MW of High Harmonic Fast Wave (HHFW) heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle parameter space that is relevant for both -heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component (PFC) solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, energetic particle stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.