Beam Heating Power Research Articles

Prediction of DIII-D Pedestal Structure From Externally Controllable Parameters

The sharp increase of pressure at the edge of a high confinement mode (H-mode) plasma, the pedestal, strongly impacts overall plasma performance. Predicting the pedestal is a necessity to control and optimize tokamak operations. An experimental data-driven machine learning (ML) approach is presented, which predicts the pedestal heights and widths of electron density ( $n_{\mathrm {e}}$ ) and electron temperature ( $T_{\mathrm {e}}$ ) profiles as well as the separatrix $n_{\mathrm {e}}$ from externally controllable parameters such as the plasma shape, heating method and power, and gas puff rate and integrated gas puff. The one modeling framework for integrated tasks (OMFIT) framework was used with DIII-D data to efficiently, robustly, and automatically build a database of pedestal parameters to train ML models. Database creation was enabled by the search engine tool for DIII-D data, TokSearch, which parallelizes data fetching, enabling fast searches through basic signals of thousands of DIII-D shots and selection of relevant time intervals. Principal component analysis (PCA) separated the database into three clusters that represent classes of plasma shapes that are regularly used in DIII-D. The most important parameters for setting the pedestal structure were plasma current ( $I_{\mathrm {p}}$ ), toroidal magnetic field ( $B_{\phi }$ ), neutral beam heating power ( $P_{{\mathrm {NBI}}}$ ), and shaping quantities. The deep jointly informed neural networks (DJINNs) algorithm was applied to identify suitable neural network (NN) architectures that appropriately capture the features of the pedestal database. Separate NNs were implemented for each pedestal parameter, and ensembling methods were used to improve the prediction accuracy and allowed estimation of the prediction uncertainty. The pedestal predictions of the test dataset lie within the measurement uncertainties of the pedestal parameters. The NN outperformed simple linear regression (LR) analysis, indicating nonlinear dependencies in the pedestal structure. The presented achievements illustrate a promising path for future research, using feature extraction to infer experimental trends and thereby improve pedestal models as well as deploying NN for a fast pedestal prediction in DIII-D scenario development.

First neutral beam experiments on Wendelstein 7-X

In the previous divertor campaign, the Wendelstein 7-X (W7-X) device injected 3.6 MW of neutral beam heating power allowing for the achievement of densities approaching 2 × 1020 m−3, and providing the first initial assessment of fast ion confinement in a drift optimized stellarator. The neutral beam injection (NBI) system on W7-X is comprised of two beam boxes with space for four radio frequency sources each. The 3.6 MW of heating reported in this work was achieved with two sources in the NI21 beam box. The effect of combined electron-cyclotron resonance heating (ECRH) and NBI was explored through a series of discharges varying both NBI and ECRH power. Discharges without ECRH saw a linear increase in the line-integrated plasma density, and strong peaking of the core density, over the discharge duration. The presence of 1 MW of ECRH power was found to be sufficient to control a continuous density rise during NBI operation. Simulations of fast ion wall loads were found to be consistent with experimental infrared camera images during operation. In general, NBI discharges were free from the presence of fast ion induced Alfvénic activity, consistent with low beam betas. These experiments provide data for future scenario development and initial assessment of fast-ion confinement in W7-X, a key topic of the project.

Stability of beta-induced Alfvén eigenmodes (BAE) in DIII-D

Although the stability of ellipticity, toroidal and reversed-shear Alfvén eigenmodes (EAE, TAE, RSAE) are relatively well understood, less is known about the stability of lower-frequency modes such as the beta-induced Alfvén eigenmode (BAE) but, because they are often unstable in present devices and are implicated in fast-ion transport, understanding their stability is vital. BAE stability is studied in primarily weak or reversed shear DIII-D plasmas with sub-Alfvénic deuterium beams. Modes are classified based on electron cyclotron emission, beam emission spectroscopy, magnetics, and interferometer data. The study is limited to the initial two seconds of the discharge, where the evolving q profile provides an effective scan of the dependence of stability upon q. In a dedicated experiment, BAEs are unstable at times in the discharge when the minimum of the safety factor q min is close to a rational number. The observed mode frequencies are usually close to analytic estimates of the BAE accumulation point and the eigenfunction peaks in the vicinity of q min. Unstable BAEs usually occur in bursts that chirp rapidly in frequency. To isolate the importance of thermal and beam gradients in driving the modes, the beam and electron cyclotron heating power is altered for 50–100 ms durations in reproducible discharges. As expected from the resonance condition, BAEs depend sensitively on the beam power and injection geometry. Modes only persist for ∼25 ms because the anisotropic beam population only interacts strongly with the modes over a relatively narrow range of q. A database of over 1000 beam-heated discharges shows that BAEs are more likely to be unstable when the poloidal beta exceeds 0.5.

Impact of the new TCV baffled divertor upgrade on pedestal structure and performance

Abstract A new set of carbon tiles, neutral beam heating optics and gas baffles were installed on TCV during the baffled divertor upgrade in early 2019. The installation of the baffles allows a deconvolution of the roles of main chamber and divertor neutral pressure on the H-mode pedestal structure. This physical barrier allows relatively high neutral pressures to be constrained to the divertor, thus preventing neutrals from entering the main chamber and potentially degrading core confinement. This study presents the experimentally measured and modelled pedestal heights and structure for a series of H-mode discharges prior to and after this upgrade. Increased pedestal performance at high divertor neutral pressure was observed after the baffled divertor upgrade. This was consistent across all triangularities and outer target locations investigated and is attributed to higher pedestal top temperatures being maintained at high gas injection rates. ASTRA simulations indicated beam heating power coupled to the plasma did not significantly vary after the baffled divertor upgrade or as a function of divertor neutral gas pressure. Analysis of the pedestal structure exposed a strong correlation between pedestal performance and the density pedestal position prior to and after the baffled divertor upgrade. The baffled divertor upgrade limited the outward shift of the density pedestal, thus maintaining higher pedestal performance at high divertor neutral pressures. Stability analysis indicated the majority of discharges studied were within 25% of the stability boundary. No correlation was found between the distance from the stability boundary and pedestal performance or structure. Comparison with the EPED1 model indicated that TCV discharges do not have a fixed dependence between pedestal β θ and pedestal width. A large variation in the EPED1 relating parameter was observed and found to vary with the density pedestal position.

Commissioning and initial operation of the W7-X neutral beam injection heating system

Abstract The first, of two planned, neutral beam injectors for the stellarator Wendelstein 7-X (W7-X) was commissioned for and participated in the experimental campaign (OP1.2b) from July to October 2018. The injector was equipped with two RF driven ion sources from which 90A of positive hydrogen ions were extracted at 55 kV. After neutralization, the two sources provided >3 MW of neutral beam heating power to the stellarator plasma. During the experimental campaign >300 shots were successfully performed for plasma heating or to allow for measurement of the ion temperature profile by charge exchange recombination spectroscopy (CXRS). The initial operation of the NBI system on W7-X was very successful, demonstrating both increased central plasma density and stored plasma energy, or allowing for the collection of the time resolved ion temperature over the bulk of the plasma. This paper presents, briefly, source conditioning and performance, focusing on some of the initial problems encountered. Described in more detail is the most significant challenge overcome during commissioning: failure to couple RF power into one of the sources.

High fusion performance at high Ti/Te in JET-ILW baseline plasmas with high NBI heating power and low gas puffing

This paper presents the transport analysis of high density baseline discharges in the 2016 experimental campaign of the Joint European Torus with the ITER-Like Wall (JET-ILW), where a significant increase in the deuterium–deuterium (D–D) fusion neutron rate (~2.8 × 1016 s−1) was achieved with stable high neutral beam injection (NBI) powers of up to 28 MW and low gas puffing. Increase in Ti exceeding Te were produced for the first time in baseline discharges despite the high electron density; this enabled a significant increase in the thermal fusion reaction rate. As a result, the new achieved record in fusion performance was much higher than the previous record in the same heating power baseline discharges, where Ti = Te. In addition to the decreases in collisionality and the increases in ion heating fraction in the discharges with high NBI power, Ti > Te can also be attributed to positive feedback between the high Ti/Te ratio and stabilisation of the turbulent heat flux resulting from the ion temperature gradient driven mode. The high Ti/Te ratio was correlated with high rotation frequency. Among the discharges with identical beam heating power, higher rotation frequencies were observed when particle fuelling was provided by low gas puffing and pellet injection. This reveals that particle fuelling played a key role for achieving high Ti/Te, and the improved fusion performance.

Control and data acquisition upgrades for NSTX-U

The extensive NSTX Upgrade (NSTX-U) Project includes major components which allow a doubling of the toroidal field strength to 1T, of the Neutral Beam heating power to 12 MW, and the plasma current to 2 MA, and substantial structural enhancements to withstand the increased electromagnetic loads. The maximum pulse length will go from 1.5 to 5s. The larger and more complex forces on the coils will be protected by a Digital Coil Protection System, which requires demanding real-time data input rates, calculations and responses. The amount of conventional digitized data for a given pulse is expected to increase from 2.5 to 5 GB per second of pulse. 2-D Fast Camera data is expected to go from 2.5 GB/pulse to 10, and another 2 GB/pulse is expected from new IR cameras. Our network capacity will be increased by a factor of 10, with 10 Gb/s fibers used for the major trunks. 32-core Linux systems will be used for several functions, including between-shot data processing, MDSplus data serving, between-shot EFIT analysis, real-time processing, and for a new capability, between-shot TRANSP. Improvements to the MDSplus events subsystem will be made through the use of both UDP and TCP/IP based methods and the addition of a dedicated “event server”.

Electric probe diagnostics for measuring SOL parameters, wall and divertor fluxes in KSTAR

Some components in electric probe diagnostics (EPDs) are improved in order to investigate characteristics of edge plasmas in the upstream scrape-off-layer (SOL) region and to measure wall and divertor fluxes during L-mode and H-mode plasma discharges in the Korea Superconducting Tokamak Advanced Research (KSTAR). From the upgrades in the EPDs, the measured error of the elapsed distance for the evaluation of the SOL profiles can be reduced up to 1% and the ion saturation current of up to 1.0A near an outer strike point (OSP) can be measured at the divertor region. In the SOL profile measurements during L-mode and inner wall limited plasma (BT=2.0T, Ip=0.4MA), the e-folding lengths in the main SOL region λTe and λne are evaluated as 3.5cm and 2.1cm, respectively. From particle flux measurement at the far SOL region during a diverted ELMy H-mode discharge (BT=1.8T, Ip=0.65MA), peaked heat flux toward to outboard wall during ELM bursts is estimated up to ∼20kWm−2, which may be less than 1% of the peaked divertor heat flux expected for the neutral beam (NB) heating power PNB of ∼2.66MW and stored energy WTOT of ∼0.38MJ. In addition, the movement of the OSP during a diverted H-mode plasma (BT=2.4T, Ip=0.5 MA) can be detected from the divertor probe measurement and peaked heat flux near the OSP is estimated as few MWm−2 for PNB=3.0 MW, WTOT=∼0.25MJ. In this work, the technical details of the EPDs including its upgrades and some experimental results from EPD measurements are presented.

Fast-ion transport in qmin>2, high-β steady-state scenarios on DIII-D

Results from experiments on DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] aimed at developing high β steady-state operating scenarios with high-qmin confirm that fast-ion transport is a critical issue for advanced tokamak development using neutral beam injection current drive. In DIII-D, greater than 11 MW of neutral beam heating power is applied with the intent of maximizing βN and the noninductive current drive. However, in scenarios with qmin>2 that target the typical range of q95= 5–7 used in next-step steady-state reactor models, Alfvén eigenmodes cause greater fast-ion transport than classical models predict. This enhanced transport reduces the absorbed neutral beam heating power and current drive and limits the achievable βN. In contrast, similar plasmas except with qmin just above 1 have approximately classical fast-ion transport. Experiments that take qmin>3 plasmas to higher βP with q95= 11–12 for testing long pulse operation exhibit regimes of better than expected thermal confinement. Compared to the standard high-qmin scenario, the high βP cases have shorter slowing-down time and lower ∇βfast, and this reduces the drive for Alfvénic modes, yielding nearly classical fast-ion transport, high values of normalized confinement, βN, and noninductive current fraction. These results suggest DIII-D might obtain better performance in lower-q95, high-qmin plasmas using broader neutral beam heating profiles and increased direct electron heating power to lower the drive for Alfvén eigenmodes.

Demonstration of real-time control for poloidal beta in KSTAR

Sustaining the plasma in a stable and a high performance condition is one of the important control issues for future steady state tokamaks. In the 2014 KSTAR campaign, we have developed a real-time poloidal beta (βp) control technique and carried out preliminary experiments to identify its feasibility. In the control system, the βp is calculated in real time using the measured diamagnetic loop signal, and compared with the target value leading to the change of the neutral beam (NB) heating power using a feedback PID control algorithm. To match the requested power of NB which is operated with constant voltage, the on-time periods of the intervals were adjusted as the ratio of the required power to the maximum achievable one. This paper will present the overall procedures of the βp control, the βp estimation process and NB operation scheme implemented in the plasma control system (PCS), and the analysis on the preliminary experimental results.

Toroidal ripple transport of beam ions in the mega-ampère spherical tokamak

The transport of injected beam ions due to toroidal magnetic field ripple in the mega-ampère spherical tokamak (MAST) is quantified using a full orbit particle tracking code, with collisional slowing-down and pitch-angle scattering by electrons and bulk ions taken into account. It is shown that the level of ripple losses is generally rather low, although it depends sensitively on the major radius of the outer midplane plasma edge; for typical values of this parameter in MAST plasmas, the reduction in beam heating power due specifically to ripple transport is less than 1%, and the ripple contribution to beam ion diffusivity is of the order of 0.1 m2 s–1 or less. It is concluded that ripple effects make only a small contribution to anomalous transport rates that have been invoked to account for measured neutron rates and plasma stored energies in some MAST discharges. Delayed (non-prompt) losses are shown to occur close to the outer midplane, suggesting that banana-drift diffusion is the most likely cause of the ripple-induced losses.

Neutral beam injection simulation of EAST

EAST (Experimental Advanced Superconducting Tokamak) is a fully superconducting tokamak with single/double null divertor configuration. The NBI (Neutral Beam Inject) system of EAST has two deuterium beam lines, in which beam power is 4 MW and energy is 80 keV. Neutral beam heating in EAST tokamak is simulated using NUBEAM code. The NUBEAM module is a Monte Carlo modeling of fast ion species in an axisymmetric tokamak. With the numerical simulation, the scenarios of the future neutral beam heating in EAST plasma are predicted. The influences of the injection angle on the neutral beam heating, current drive and shine-through power are analyzed by the simulation.

Characteristics of momentum transport in JT-60U H-mode plasmas

Properties of the toroidal momentum diffusivity (χϕ) and the convection velocity (Vconv) in JT-60U H-mode plasmas have been obtained by plasma parameter scans such as the plasma current, neutral beam heating power and electron density. The toroidal momentum diffusivity increases with increasing heat diffusivity (χi) over a wide range of radii (r/a = 0.2–0.6) and χϕ/χi ∼ 0.7–3 at the half radius (r/a = 0.5). The inward convection velocity (−Vconv) increases with increasing χϕ, and −Vconv/χϕ ∼ 0.5–2 (m−1) at r/a = 0.5. It is found that the ratio χϕ/χi increases with increasing ion temperature and decreases with increasing electron density. These tendencies are observed in other radial positions of r/a = 0.3, 0.4 and 0.6. Moreover, the ratio −Vconv/χϕ at r/a = 0.4, 0.5 and 0.6 increases with increasing ion and electron temperatures or temperature gradients.

New vacuum electron beam processing method based on temperature closed-loop control

This is a novel combination of vacuum electron beam (EB) processing and automatic control theory. The concept of full EB processing method was presented. The digital control system for temperature closed-loop control in EB processing based on scanning track control was realized. The method of controlling electron beam current and heating power by real-time adjusting grid bias voltage is feasible. EB processing methods consist of two stages: heating-up stage and temperature maintaining stage. In the former stage, a fuzzy controller was used to control the temperature rising velocity, the sensitivity of the control was evaluated; in the latter stage, an integral separation PID controller was used to keep the workpiece's temperature constant, ideal static and dynamic performances were obtained. The temperature rising velocity control and the maintaining control compose the full EB processing method, the transition of the two stages is bumpless.

Reheat Mode Discharges in Search of Attainable High Stored Energy and Density Limit of Compact Helical System

The results of experiments to attain high stored energy and density in the Compact Helical System (CHS) are reported. The experiments have been carried out under the maximum neutral beam heating power and highest magnetic field strength of the CHS. With the help of the reheat mode, we have so far reached a stored energy of 9.4 kJ and a density limit expressed as nc = 0.65·(Pabs·Bt/Vp)0.5 for the CHS. In the high-density regime, the confinement of CHS plasma is limited by radiation collapse. A multichannel Hα light detector system shows an asymmetric feature in the poloidal cross section and indicates that confinement degradation in the high-density regime begins at the inboard side where the CHS plasma is close to the vacuum vessel wall.

The physics of edge resonant magnetic perturbations in hot tokamak plasmas

Small edge resonant magnetic perturbations are used to control the pedestal transport and stability in low electron collisionality (νe*), ITER [ITER Physics Basis Editors et al., Nucl. Fusion 39, 2137 (1999)] relevant, poloidally diverted plasmas. The applied perturbations reduce the height of the density pedestal and increase its width while increasing the height of the electron pedestal temperature and its gradient. The effect of the perturbations on the pedestal gradients is controlled by the current in the perturbation coil, the poloidal mode spectrum of the coil, the neutral beam heating power, and the divertor deuterium fueling rate. Large pedestal instabilities, referred to as edge localized modes (ELMs), are completely eliminated with radial magnetic perturbations (δbr(m∕n)) at the q=m∕n=11∕3 surface exceeding δbr(11∕3)Bϕ−1=2.6×10−4, where Bϕ is the toroidal magnetic field on axis. The resulting ELM-free H-mode plasmas have stationary densities and radiated power, are maintained in DIII-D for up to 2550ms (17 energy confinement times), and are limited only by hardware constraints. It is found that changes caused by the magnetic perturbations in the pedestal profiles cannot be explained by a straightforward application of stochastic quasilinear diffusion theory due to the complex nature of the transport physics involved when boundary layer field lines connect regions of hot plasma directly to material surfaces.

Impact of wall saturation on particle control in long and high-power-heated discharges in JT-60U

Modification for long pulse operation up to 65 s with a neutral beam heating power of 12 MW (30 s) has provided an opportunity for studies on plasma-wall interaction on a time scale of tens of seconds. During a long pulse, ELMy H-mode discharge, the net wall-pumping rate, evaluated by a particle balance analysis, gradually decreases, and subsequently, becomes zero in ∼20 s after the neutral beam injection starts. This situation is interpreted as ‘wall saturation’. Wall saturation is observed in repetitive long pulse discharges. Although the outgas due to the increase in the divertor plate temperature and increase in the static retention with pulses also contribute to the particle balance, the dynamic retention process dominates the particle balance and determines the wall saturation. Under the condition of wall saturation, the plasma density continuously increases due to insufficient divertor-pumping rate. This density rise results in detachment of the outer divertor plasma, followed by an X-point MARFE. Results from the calculation of a 2-dimensional plasma transport code indicate that higher divertor-pumping efficiency by a factor of 1.5 is required to avoid the detachment. Note that no negative effects on impurity behaviour such as carbon bloom, or significant dilution of the core plasma has been observed even in the case of the total input energy reaching 350 MJ.

A component test facility based on the spherical tokamak

Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes Γn ≈ 8.8 × 1013 n s−1 cm−2 (‘wall loading’ WL ≈ 2 MW m−2), over size-scale >105 cm2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year (‘neutron fluence’ >0.3 MW yr−1 m−2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ∼3, Ip ∼ 12 MA, BT ∼ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere level. A systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering and technology test missions.

Measurement of the damping rate of n = 1 toroidal Alfvén eigenmodes as a function of the neutral beam heating power and plasma β on JET

The dependence of the measured damping rate upon the neutral beam heating power (PNBI) and plasma β has been experimentally investigated for n = 1 toroidal Alfvén eigenmodes (TAEs) in JET limiter plasmas with a monotonic safety factor profile and low edge magnetic shear. The mode frequency is observed to decrease with increasing β, as predicted by various theoretical models. For low PNBI<3 MW, a single n = 1 TAE splits into two modes closely spaced around the expected TAE frequency, and the mode damping rate decreases with increasing β. On the other hand, at higher PNBI>6 MW, no frequency splitting is observed and there are indications that the mode damping rate increases with increasing β.

Experimental studies of the JET NBI neutraliser plasma

Neutralisation efficiency is an important issue in the design of neutral beam injection (NBI) systems for heating and fuelling fusion plasmas. The JET Neutral Beam Injectors system does not reach the efficiency that is expected on the basis of current understanding of the neutralisation process. If this neutralisation efficiency deficit could be eliminated by re-design of the neutralisers, additional neutral beam heating power would become available. The interaction of the high-energy beam ions and the gas in the neutraliser cell results in the partial neutralisation of the beam and also in the formation of a low temperature plasma inside the neutraliser. The interaction between this plasma, the neutral gas and the walls of the cell can give rise to various phenomena capable of depleting the neutral gas molecular density hence reducing the effective neutralisation target. This paper presents the results of a number of experiments that were carried out in order to measure the plasma and gas parameters inside the neutraliser. The plasma parameters were obtained using a combination of planar Langmuir probes and a novel electrical diagnostic called a ‘plasma eater’. The gas pressure was measured using hot cathode ionisation gauges and capacitance manometers. An attempt has been made to measure the neutral hydrogen gas temperature using a spectroscopic technique that requires the measurement of the intensities of the spectral lines of the hydrogen Fulcher-α system [Plasma Sources Sci. Technol. 10 (2001) 17].