MW NBI Research Articles

Prediction of high-performance scenario with localized magnetic shear reversal on EAST tokamak

An advanced scenario with high poloidal beta, large-radius internal transport barriers (ITBs) and localized reversed-shear q-profile, has been demonstrated in DIII-D/EAST joint experiments recently on the DIII-D tokamak. This scenario is also one of the future explorations on EAST. In this article, integrated modeling has been utilized to explore the regime with localized reversed-shear q-profile and high performance based on the developed long-pulse scenario on EAST. A scenario with a strong reversed-shear q-profile has been achieved with the combination of 2 MW off-axis ECH, 2.6 MW LHW and 2 MW NBI (H98y2 ∼ 1.35, ∼ 2.56, ∼ 2.23, ). In addition, a small amount (∼0.1 MW) of centrally deposited ECH facilitates the generation of the bootstrap current on or near the axis, and hence reduces the level of the q-profile in this region; thus, a scenario with localized magnetic shear reversal and strong large-radius ITBs is obtained (H98y2 ∼ 1.4, ∼ 2.66, ∼ 2.31, ). Modeling predictions indicate that off-axis ECH/LHW and high plasma density promote the formation of large-radius ITBs. A further increase in the injected power of centrally deposited ECH (0.15 MW) or NBI (4 MW) leads to an increase in the temperature gradient, especially the electron temperature near the axis, and large-radius ITBs disappear in all channels (T e, T i, n e); also, the confinement quality decreases. The bootstrap current density profiles and LHW-driven current profiles tend to peak near the axis due to the increase in electron temperature profiles, which eventually leads to peaked total current profiles, similar to the EAST long-pulse scenario. Changes in magnetic shear result in the disappearance of large-radius ITBs. Modeling validation will be carried out in the near future on the EAST tokamak.

Accumulated 2-D neutron flux distribution during KSTAR operation

Since the successful first H-mode in 2010, duration of H-mode is extended to 1 min and the highest stored plasma energy up to 0.6 MJ is obtained by developing hybrid and internal transport barrier (ITB) mode as well as typical H-mode. Major heating system is the neutral beam (NB) of 100 keV with 5.5 MW and it is planned that the new additional 6 MW NBI is installed in 2018. So there is a lot of production of fast neutrons coming from via D(d,n)3He reaction and it is found that most of neutrons are 2.45 MeV neutrons coming from deuterium beam plasma interaction and the total neutron fluence during a campaign could be estimated to be about 1019. In particular, several nickel activation samples are installed inside/outside the vacuum vessel poloidally and toroidally and the total accumulated neutron flux is estimated from 811 keV gamma emission measurement after the campaign. The results show that the neutron activation both inside and outside the vacuum vessel has toroidally non-axisymmetric distribution and it has the highest neutron flux around poloidal limiter area. Considering the prompt ion loss at nominal KSTAR operation including the high beta-p operation, prompt loss ions from 100 keV neutral ion beams hitting on three poloidal limiters was not negligible and the non-uniform distribution is discussed in term of prompt ion loss to the poloidal limiter. In addition, this work reports on the quantitative analysis of poloidal distribution of the neutron emission based on accumulated flux measurements.

Scenario development during commissioning operations on the National Spherical Torus Experiment Upgrade

The National Spherical Torus Experiment Upgrade (NSTX-U) will advance the physics basis required for achieving steady-state, high-beta, and high-confinement conditions in a tokamak by accessing high toroidal fields (1 T) and plasma currents (1.0–2.0 MA) in a low aspect ratio geometry (A = 1.6–1.8) with flexible auxiliary heating systems (12 MW NBI, 6 MW HHFW). This paper describes the progress in the development of L- and H-mode discharge scenarios and the commissioning of operational tools in the first ten weeks of operation that enable the scientific mission of NSTX-U. Vacuum field calculations completed prior to operations supported the rapid development and optimization of inductive breakdown at different values of ohmic solenoid current. The toroidal magnetic field (BT0 = 0.65 T) exceeded the maximum values achieved on NSTX and novel long-pulse L-mode discharges with regular sawtooth activity exceeded the longest pulses produced on NSTX (tpulse > 1.8 s). The increased flux of the central solenoid facilitated the development of stationary L-mode discharges over a range of density and plasma current (Ip). H-mode discharges achieved similar levels of stored energy, confinement (H98y,2 > 1) and stability (βN/βN-nowall > 1) compared to NSTX discharges for Ip ⩽ 1 MA. High-performance H-mode scenarios require an L–H transition early in the Ip ramp-up phase in order to obtain low internal inductance (li) throughout the discharge, which is conducive to maintaining vertical stability at high elongation (κ > 2.2) and achieving long periods of MHD quiescent operations. The rapid progress in developing L- and H-mode scenarios in support of the scientific program was enabled by advances in real-time plasma control, efficient error field identification and correction, effective conditioning of the graphite wall and excellent diagnostic availability.

Assessment of operational space for long-pulse scenarios in ITER

The operational space (Ip − n) for long-pulse scenarios (Δtburn ∼ 1000 s, Q ⩾ 5) of ITER has been assessed by 1.5D core transport modelling with pedestal parameters predicted by the EPED1 code by a set of transport codes under a joint activity carried out by the Integrated Operational Scenario ITPA group. The analyses include the majority of transport models (CDBM, GLF23, Bohm/gyroBohm (BgB), MMM7.1, MMM95, Weiland, scaling-based) presently used for interpretation of experiments and ITER predictions. The EPED1 code was modified to take into account boundary conditions predicted by SOLPS4 for ITER. In contrast to standard EPED1 assumptions, EPED1 with the SOLPS boundary conditions predicts no degradation of the pedestal pressure as density is reduced. Lowering the plasma density to ne ∼ (5–6) × 1019 m−3 leads to an increased plasma temperature (similar pedestal pressure), which reduces the loop voltage and increases the duration of the burn phase to Δtburn ∼ 1000 s with Q ⩾ 5 for Ip ⩾ 13 MA at moderate normalized pressure (βN ∼ 2). These ITER plasmas require the same level of additional heating power as the reference Q = 10 inductive scenario at 15 MA (33 MW NBI and 17–20 MW EC heating and current drive power). However, unlike the ‘hybrid’ scenarios considered previously, these H-mode plasmas do not require specially shaped q profiles nor improved confinement in the core for the transport models considered in this study. Thus, these medium density H-mode plasma scenarios with Ip ⩾ 13 MA present an attractive alternative to hybrid scenarios to achieve ITER's long-pulse Q ⩾ 5 scenario and deserve further analysis and experimental demonstration in present tokamaks.

ECH effects on toroidal rotation: KSTAR experiments, intrinsic torque modelling and gyrokinetic stability analyses

Toroidal rotation profiles have been investigated in KSTAR H-mode plasma using combined auxiliary heating by co-neutral beam injection (NBI) and electron cyclotron resonance heating (ECH). The ion temperature and toroidal rotation are measured with x-ray imaging crystal spectroscopy and charge exchange recombination spectroscopy. H-mode plasma is achieved using co-current 1.3 MW NBI, and a 0.35 MW ECH pulse is added to the flat-top of H-mode. The core rotation profiles, which are centrally peaked in the pure NBI heating phase, flatten when ECH is injected, while the edge pedestal is unchanged. Dramatic decreases in the core toroidal rotation values (ΔVtor/Vtor ∼ −30%) are observed when on-axis ECH is added to H-mode. The experimental data show that the decrease of core rotation velocity and its gradient are correlated with the increase of core electron temperature and its gradient, and also with the likely steepening of the density gradient. We thus explore the viability of a hypothesized ITG (ITG ion temperature gradient instability) → TEM (trapped electron mode instability) transition as the explanation of the observed counter-current flow induced by ECH. However, the results of linear microstability analyses using inferred profiles suggest that the TEM is excited only in the deep core, so the viability of the hypothesized explanation is not yet clear.

Projected performance of ITER based on different theoretical based pedestal temperature models

Self-consistent modeling of the ITER has been carried out using the 1.5D BALDUR integrated predictive modeling code. In these simulations, the boundary is taken to be at the top of the pedestal, where the pedestal values are described using the theory-based pedestal scalings. These pedestal temperature scalings are based on three different pedestal width models: magnetic and flow shear stabilization, flow shear stabilization, and normalized poloidal pressure. The pedestal width scalings are combined with a pedestal pressure gradient scalings based on ballooning mode limit to predict the pedestal temperature. The developed pedestal temperature scalings are used together with a core transport model, which is a combination of an anomalous transport and a neoclassical transport. An anomalous transport is calculated either using the Mixed Bohm/gyro-Bohm (Mixed B/gB) core transport model or the Multimode (MMM95) core transport model, while a neoclassical transport is computed using the NCLASS model. At the reference design point (with 40 MW auxiliary heating: 33 MW NBI and 7 MW RF), it is found that the pedestal temperatures with different pedestal width scaling ranges from 2.4 keV to 2.8 keV. As a result, the performances with the same anomalous core transport model are almost similar. It is also found that the simulations using MMM95 yield better performance than those using Mixed B/gB. In addition, when the MMM95 is used, it appears that the ion temperature gradient (ITG) and trapped electron modes (TEM) are the most dominant modes. When the Mixed B/gB is used, it appears that the Bohm contribution is the most dominated term.

OEDGE modeling of the DIII-D H-mode 13CH4 puffing experiment

Use of carbon in tokamaks leads to a serious tritium retention issue due to co-deposition. To further investigate the processes involved, a detached ELMy H-mode (6.5MW NBI) experiment was performed on DIII-D in which 13CH4 was puffed into the main vessel through the toroidaly symmetric pumping plenum at the top of lower single-null discharges. Subsequently, the 13C content of tiles taken from the vessel wall was measured. The interpretive OEDGE code was used to model the results. The 13C deposition pattern could be reproduced, in general shape and magnitude, by assuming in the code the existence of a parallel flow and a radial pinch in the scrape-off layer. Parallel flows of Mach ∼0.3 toward the inner divertor and a radial pinch ∼10–20m/s (+R-direction) were found to yield 13C deposition comparable to the experiment.

06/01183 Motion and lifetime of dust grains in a tokamak plasma: Martin, J. D. et al. Journal of Nuclear Materials, 2005, 337–339, 114–118

In this paper, scrape-off layer measurements in DIII-D are presented as a function of the main discharge plasma parameters. A systematic study is under way to understand and predict the behavior of the edge and divertor plasma in DIII-D and this scaling behavior will be crucial for the design of ITER. To facilitate the studies, a fast reciprocating Langmuir probe incorporating five graphite tips was installed at the midplane of DIII-D which has the capability of performing multiple plunges 1 cm inside the separatrix during 5 MW of NBI. Density and temperature profiles in the midplane (reciprocating probe), near the top (Thomson scattering) and at the lower divertor plate (fixed Langmuir probe array) are compared by mapping the measurements into magnetic flux coordinates. Local pressure measurements are compared on different parts of a flux surface. The three different local measurements also indicate the spatial evolution of plasma conditions as plasma approaches the divertor plate. Ohmic and L-mode discharges exhibit similar (exponential) density and temperature decay in the scrape-off layer. H-mode discharges, however, display a faster spatial decay reflecting at least a factor of 3 decrease in the perpendicular diffusion coefficient. Consistency of the magnitude and scaling behavior of the edge profile parameters with models of the scrape-off layer is examined.

ICRF heating for Wendelstein 7-X

The stellarator Wendelstein 7-X is being built in Greifswald, Germany, by the Max-Planck-Institut für Plasmaphysik. Initially, the heating system will consist of 10 MW ECRH and 4 MW ICRF heating (both cw) and 5 MW NBI (for 15 s every 3 min). ECRH is the principal heating method at a magnetic field of B=2.5 T. At other magnetic fields or densities higher than the ECRH cutoff, ICRF heating will sustain the plasmas. The ICRF system is planned to consist of two radially movable antennas powered by two generators with an output power of 2 MW each. The antenna plasma load will be matched in real time to the generator impedance using ferrite loaded transmission line elements. In addition, 3 dB couplers will passively protect the generators from RF power reflections during rapid changes of the antenna plasma load that are too fast for the ferrite matching system.

Effects of non-vanishing toroidal current densities on stability in the Wendelstein 7-AS stellarator

At the W7-AS stellarator the influence of non-vanishingtoroidal current densities on stability is studied over a broad rangeof externally provided values of the rotational transform (0.301 ⩽ ιext ⩽ 0.565). By means of inductive current drive andby varying the heating scenarios, using 1 MW NBI and 0.4 MW ECRH,access is given to a variety of current density radial profiles withsignificant shear. Disruptive-like events accompanied by fast energyloss are observed as soon as low order major rational resonances appear inthe outer region of the plasma in discharges with significantcontributions of net toroidal ohmic current in the positive direction(increasing the rotational transform). To some extent the applied loopvoltage is able to prevent a total decay of the plasmacurrent. Preceding the partial or total energy loss, a perturbingeffect in the equilibrium is observed in the radial profile of theelectron temperature, as derived from ECE measurements, and the tearingmode structure is identified by tomographic reconstruction of theemission in the soft X ray wavelength region. A Δ code is usedto calculate numerically the stability of the current density radialprofiles against tearing modes, and the predictions are compared with experimental observations. ECR heated plasmas with inductive currentdrive in the negative direction are observed to be stable atιext = 0.48 and 0.52 in a reversed magnetic shearconfiguration (as defined in the tokamak sense: dq/dr < 0). No significant tearing mode activity is found inthe presence of large bootstrap and Ohkawa current densities inplasmas without ohmic current drive. However, a deterioration of confinement is observed due to other effects on rational surfaces atthe edge.

Core-edge studies with boundary island configurations on the W7-AS stellarator

Core, edge and scrape-off-layer plasma behaviour is studied principally under conditions of an a = 5/9 boundary island configuration - which is relevant for the upcoming W7-AS divertor campaign - but for now with ten inboard sector limiters. The major focus is on compatibility between good core confinement and attainment of high recycling at the limiter. At low input power Pin0.4 MW, operation at densities necessary to attain effective divertor action in the future invariably leads to a transition to the ELM-free H-mode accompanied by lower edge densities and increased core radiation until radiation collapse ensues. Thereby, enhancement factors in E of nearly two above the international stellarator confinement scaling are transiently achieved. The threshold density ethr, necessary to attain the H-mode increases with heating power, such that at 2 MW NBI heating power the H-mode is completely suppressed and peak densities at the limiter exceeding 1.5 × 1020 m-3 are realized. The efficacy of newly-installed control coils designed to manipulate the island geometry is tested. Their influence on the core plasma is verified. Due to geometrical effects associated with the mutual shadowing of the inboard limiters, statements regarding the influence on island physics must await the divertor configuration.

Compact toroid injection experiment in JFT-2M

Compact toroid (CT) injection experiments with H mode plasmas were carried out for the first time in the JFT-2M tokamak. The soft X ray emission profile shows deep CT penetration into the core region of H mode plasmas heated by 1.2 MW NBI as well as in OH plasmas, with a toroidal magnetic field of 0.8 T. The line averaged electron density rapidly increased by Δn̄e ≈ 0.2 × 1019 m-3 at a rate of 4 × 1021 m-3·s-1 in H mode and the fuelling efficiency was roughly 20%. An asymmetric radial profile in the soft X ray emission was produced for ∼50 μs by deep penetration of the CT.

The spherical tokamak programme at Culham

The spherical tokamak (ST) is the low aspect ratio limit of the conventional tokamak and appears to offer attractive physics properties in a simpler device. The START (Small Tight Aspect Ratio Tokamak) experiment provided the world's first demonstration of the properties of hot plasmas in an ST configuration and was operational at Culham from January 1991 to March 1998, obtaining plasma currents of up to 300 kA and pulse durations of ∼50 ms. Its successor, MAST, is nearing completion and is a purpose built, high vacuum machine designed to have a tenfold increase in plasma volume with plasma currents of up to 2 MA. Current drive and heating will be by a combination of induction-compression as on START, a high performance central solenoid, with 1.5 MW ECRH and 5 MW of NBI. The promising results from START are reviewed, and the many challenges posed for the next generation of purpose built STs (such as MAST) are described.

The diagnosed mobile limiters of the TJ-II stellarator for plasma boundary studies

TJ-II is a medium size (major radius R=1.5 m, average plasma radius a <0.25 m, on axis magnetic field B=1 T) helical axis stellarator. The main characteristic is its magnetic configuration flexibility, due to the separate control of the different magnetic field coils. The two diagnosed mobile limiters are installed to reduce thermal loads on the thin protection plates of the contacting plasma-chamber regions and to study the plasma edge. First diagnostics are a set of thermocouples, Langmuir probes, H α-detectors and a CCD video camera with different filters (atomic lines of HeI, H α and near IR) looking at the limiter. A method of passive spectroscopy is proposed to map the electron temperature and density over the whole limiter surface by analysing the emission of helium recycling neutrals. It is expected from previous results of other stellarators, that the boundary magnetic topology will have a strong influence on the plasma–wall interaction. The mobile limiters can control the last closed magnetic surface and diagnose the plasma boundary. A qualitative different plasma edge scenario is foreseen between the limiter and the natural island divertor configuration (rational rotational transform inside the limiter radius). Plasma–wall interaction in TJ-II shows very specific features and the optimisation of the plasma edge topology can influence strongly the core plasma parameters. In particular, impurity screening will be a challenge due to the large power density which will be available in future (up to 2 MW NBI for 0.5 s). A safe operation for future high β-plasmas is also required and the mobile limiters should help to remove a fraction of the conductive/convective power.

Development of high-power heating for large helical device

Abstract Three high-power heating mechanisms, i.e. electron cyclotron resonance heating, ion cyclotron resonance heating and neutral beam injection heating will be available to achieve operation of the Large Helical Device in the (i) high nτT mode, (ii) high T i mode and (iii) high β mode. The requirements for each heating method are (i) 10 MW of ECH at a frequency of 84 GHz with a pulse length of 10 s, (ii) 3 MW of ICRF heating in the frequency range 25–100 MHz in CW operation and (iii) 20 MW of NBI at a hydrogen injection energy of 125 keV and 10 s operation. The components of each heating mechanism which need development individually exist, i.e. (i) a high-power gyrotron with an output power of 1 MW in CW operation and a high-efficiency transmission line for ECH, (ii) a high-power co-axial transmission line with cooling and an automatic matching circuit for ICRF heating and (iii) a large negative ion source and some components of the beam line for NBI. The heating scenario and the present status of the development are described for each heating system.

Scrape-off layer measurements in DIII-D

In this paper, scrape-off layer measurements in DIII-D are presented as a function of the main discharge plasma parameters. A systematic study is under way to understand and predict the behavior of the edge and divertor plasma in DIII-D and this scaling behavior will be crucial for the design of ITER. To facilitate the studies, a fast reciprocating Langmuir probe incorporating five graphite tips was installed at the midplane of DIII-D which has the capability of performing multiple plunges 1 cm inside the separatrix during 5 MW of NBI. Density and temperature profiles in the midplane (reciprocating probe), near the top (Thomson scattering) and at the lower divertor plate (fixed Langmuir probe array) are compared by mapping the measurements into magnetic flux coordinates. Local pressure measurements are compared on different parts of a flux surface. The three different local measurements also indicate the spatial evolution of plasma conditions as plasma approaches the divertor plate. Ohmic and L-mode discharges exhibit similar (exponential) density and temperature decay in the scrape-off layer. H-mode discharges, however, display a faster spatial decay reflecting at least a factor of 3 decrease in the perpendicular diffusion coefficient. Consistency of the magnitude and scaling behavior of the edge profile parameters with models of the scrape-off layer is examined.

Helium exhaust in plasmas with strong radiative edge cooling

The compatibility of radiative edge cooling by neon injection and He exhaust with the pump limiter ALT-II is studied on TEXTOR. It is demonstrated that in plasmas with strong auxiliary heating (2 MW NBI) and with the highest average electron densities (¯ n e ≈5.5×10 19 m −3 ) the effective confinement time τ p * for He has a minimum. This good pumping performance is maintained even for cases in which up to 90% of the heating power is radiated from the plasma boundary (cold radiative edge). The processes inside the scoops of the pump limiter (neutral particle transport, re-ionization) and the variation of particle confinement in the main plasma dominate this behavior.

H-Mode Outside Divertor Free from Peripheral Disruption in Doublet III

The heating efficiency and the peripheral disruption of three distinctively different divertor configurtions were studied in Doublet III at a plasma current of 600 kA up to the neutral injection power of 7.5 MW. An outside divertor configuration with a good energy confinement (H-mode), which has a null point at the outside in the major radius and of which separatrix field lines divert to the outside wall, was observed. The discharges of the outside divertor configuration were free from the peripheral disruption characterized by Hα/Dα bursts and the stored energy was nearly proportional to the injection power up to 7.5 MW of NBI.