Non-inductive Operation Research Articles

Initial testing of Alfvén eigenmode feedback control with machine-learning observers on DIII-D

A first of its kind fully data-driven system has been developed and implemented into the DIII-D plasma control system to detect and control Alfvén eigenmodes (AE) in real-time. Susceptibility to fast ion-induced AE is a challenge in fully non-inductive tokamak operation, which significantly reduces fast-particle confinement and results in degraded fusion gain. Controlling AEs in real-time to improve fast-ion confinement is, hence, important for future advanced tokamak fusion reactors. The models were implemented and tested in experiments which showed that neural networks (NN) are highly effective in detecting 5 types of AE (BAE, EAE, LFM, RSAE, TAE) using high resolution ECE. To estimate the neutron deficit, a NN has been trained that outputs the classical neutron rate using similar inputs to NUBEAM. Also a preliminary ML-based proportional control has been designed and gone through initial testing in experiment to use feedback-control on the neutral beam power to achieve desired amplitude of AE modes and neutron deficits. The effect of AEs on fast-ion confinement is measured by analysing the gap in classical neutron rate from the proposed NN-based NUBEAM and the measured neutron rate.

Correction: The optimal values of Greenwald density limit and plasma safety factor in inductive and non-inductive operation modes

Correction: The optimal values of Greenwald density limit and plasma safety factor in inductive and non-inductive operation modes

The optimal values of Greenwald density limit and plasma safety factor in inductive and non-inductive operation modes

The optimal values of Greenwald density limit and plasma safety factor in inductive and non-inductive operation modes

Predictive modeling of Alfvén eigenmode stability in inductive scenarios in JT-60SA

The JT-60SA device offers unique conditions before ITER for the study of the interaction of energetic particles with plasma waves. With similar dimensions to JET, e.g., a major radius but with a slightly more elongated plasma volume, JT-60SA is used as a high-power device where additional heating power (including 10 MW of the 500 keV Neutral Beam Injection) of up to 41 MW and the potential for high non-inductive plasma current operation pave the path for numerous challenges in physics on MHD stability, in particular, when considering the effects of energetic particles. Several operational scenarios with ITER and DEMO-relevant plasma regimes, in terms of non-dimensional plasma parameters, are anticipated. In this work, the stability of Alfvén eigenmodes (AEs) in variants of two of the most relevant operational scenarios with single null is analyzed: a full Ip inductive scenario at high density (1.1 × 1020 m−3 on-axis electron density) and 5.48MA/2.05T toroidal plasma current and magnetic field, and an advanced (hybrid) scenario with an ion energy transport barrier (ITB) and 3.5MA/2.28T toroidal plasma current and magnetic field. The workflow included the CRONOS code to establish the scenario, the ASCOT code to calculate the slowing-down energetic particle distributions for a positive/negative ion source-based neutral beam, and the MISHKA/CASTOR-K suite to calculate the MHD spectra of AEs and the associated drive/damping contributions from the NBI energetic ions, as well as the thermal ion landau damping. The systematic analysis, over a large Fourier space of the toroidal mode number/mode frequency, provides evidence that although a significant fraction of supra-Alfvénic particles stemming from the negative ion source-based neutral beam (500 keV) can, in some cases, drive to AEs in both scenarios, it is not enough to overcome the thermal ion landau damping. In addition, the advanced scenario with ITB is shown to be stable against AEs localized in the vicinity of the barrier as well, offering good prospects of sustainability of the plasma performance and of ITB. Finally, some sensitivity scan results are shown on the influence of fast ion density and q-profile on the AE mode spectra and stability.

Application of linear electron Bernstein current drive models in reactor-relevant spherical tokamaks

Electron Bernstein current drive (EBCD) systems in spherical tokamaks are sensitive to plasma and launch conditions, and therefore require large parametric scans to optimise their design. One particular bottleneck in the simulation workflow is quasilinear modelling of current drive efficiency. Linear adjoint models are an attractive alternative, offering a ∼103 × speed-up compared to quasilinear codes. While linear models are well-tested and commonly used for electron cyclotron current drive (ECCD), they have seen little use in EBCD modelling. In this work, variants of the linear model are applied to EBCD and compared to quasilinear results in a reactor-relevant plasma, i.e. Spherical Tokamak for Energy Production (STEP). This comparison reveals it is important to accurately model the collision operator and finite Larmor radius effects in the linear model. When done properly, good agreement is found with quasilinear calculations, at least for normalised minor radii ρ < 0.7 and at low power densities. The power density threshold for quasilinear effects during EBCD is found to be significantly lower than that of ECCD. This is attributed to the much lower group velocity of the electron Bernstein wave (EBW). Thus, the linear model is only valid for EBCD modelling at low power densities (e.g. ≲ 1 MW launched EBW power in STEP). This may be satisfied in present-day experimental devices, but certainly not in reactors targeting non-inductive operation.

Long-pulse high-performance H-mode plasmas achieved on EAST

A record duration of a 310 s H-mode plasma (H98y2 ∼ 1.3, ne/nGW ∼ 0.7, fBS &amp;gt; 50%) has been recently achieved on experimental advanced superconducting tokamak (EAST) with metal walls, exploiting the device's improved long-pulse capabilities. The experiment demonstrates good control of tungsten concentration, core/edge MHD stability, and particle and heat exhaust with an ITER-like tungsten divertor and zero injected torque, establishing a milestone on the path to steady-state long-pulse high-performance scenarios in support of ITER and CFETR. Important synergistic effects are leveraged toward this result, which relies purely on radio frequency (RF) powers for heating and current drive (H&amp;CD). On-axis electron cyclotron heating enhances the H&amp;CD efficiency from lower hybrid wave injection, increasing confinement quality and enabling fully non-inductive operation at high density (ne/nGW ∼ 70%) and high poloidal beta (βP ∼ 2.5). A small-amplitude grassy edge localized mode regime facilitates the RF power coupling to the H-mode edge and reduces divertor sputtering/erosion. The high energy confinement quality (H98y2 ∼ 1.3) is achieved with the experimental and simulated results pointing to the strong effect of Shafranov shift on turbulence. Transport analysis suggests that trapped electron modes dominate in the core region during the record discharge. The detailed physics processes (RF synergy, core-edge integration, confinement properties, etc.) of the steady-state operation will be illustrated in the content. In the future, EAST will aim at accessing more relevant dimensionless parameters to develop long-pulse high-performance plasma toward ITER and CFETR steady-state advanced operation.

Core-Pedestal Plasma Configurations in Advanced Tokamaks

Several configurations for the core and pedestal plasma are examined for a predefined tokamak design by implementing multiple heating/current drive (H/CD) sources to achieve an optimum configuration of high fusion power in a noninductive operation while maintaining an ideally magnetohydrodynamic (MHD) stable core plasma using the IPS-FASTRAN framework. IPS-FASTRAN is a component-based lightweight coupled simulation framework that is used to simulate magnetically confined plasma by integrating a set of high-fidelity codes to construct the plasma equilibrium (EFIT, TOQ, and CHEASE), calculate the turbulent heat and particle transport fluxes (TGLF), model various H/CD systems (TORIC, TORAY, GENRAY, and NUBEAM), model the pedestal pressure and width (EPED), and estimate the ideal MHD stability (DCON). The TGLF core transport model and EPED pedestal model are used to self-consistently predict plasma profiles consistent with ideal MHD stability and H/CD (and bootstrap) current sources. In order to evaluate the achievable and sustainable plasma beta, varying configurations are produced ranging from the no-wall stability to with-wall stability regimes, simultaneously subject to the self-consistent TGLF, EPED, and H/CD source profile predictions that optimize configuration performance. The pedestal density, plasma current, and total injected power are scanned to explore their impact on the target plasma configuration, fusion power, and confinement quality. A set of fully noninductive scenarios are achieved by employing ion-cyclotron, neutral beam injection, helicon, and lower-hybrid H/CDs to provide a broad profile for the total current drive in the core region for a predefined tokamak design. These noninductive scenarios are characterized by high fusion gain (Q ~ 4) and power (Pfus ~ 600 MW), optimum confinement quality (H98 ~ 1.1), and high bootstrap current fraction (fBS ~ 0.7) for Greenwald fraction below unity. The broad current profile configurations identified are stable to low-n kink modes either because the normalized pressure is below the no-wall limit or a wall is present.

Improvement of zero-dimensional system model and its analysis and prediction of steady-state operating regime on EAST

The zero-dimensional system model has been widely used for predicting and analyzing plasma performance in fusion reactors and designing next-generation tokamaks. These models can quickly scan and calculate various parameter, and can be used for the design of device reference operation point and preparation for more accurate one-dimensional numerical simulations. They can also be used to predict device operational parameters and heating/ current drive conditions, providing a quick reference for experimental design. However, relying on physical approximations and empirical formulas can lead to significant systematic errors. In this work we introduce a plasma equilibrium program to obtain the main plasma profile parameters and their calculations based on magnetic surface information. The bootstrap current calculation is improved by introducing the relationship between the bootstrap current coefficient of the Sauter model and the collision rate change. The improved model is validated by using experimental results from EAST, and the results of the zero-dimensional system model calculations are found to be consistent with the results of kinetic equilibrium analysis. Based on the improved model and existing experimental results, the required heating/current drive power and achievable normalized beta for steady-state, long-pulse operation of the 500 kA plasma current on EAST are analyzed and predicted. The calculation results show that EAST can achieve steady-state operation at the 500 kA plasma current with bootstrap current fraction over 50% in the parameter range of 7.0–9.5 MW heating/driving power, &lt;inline-formula&gt;&lt;tex-math id="M5"&gt;\begin{document}$ {H}_{98} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M5.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M5.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;is 1.25–1.35, and &lt;inline-formula&gt;&lt;tex-math id="M6"&gt;\begin{document}$ {f}_{{\rm{n}}{\rm{G}}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M6.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M6.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;~0.9. Additionally, to maintain the total non-inductive current, the total heating/current drive power needs to be highly sensitive to plasma confinement and density, which is the most effective way to increase the bootstrap current fraction and reduce the peak heat loads on the divertor. Improving plasma confinement is the most effective way to achieve high bootstrap current fraction and reduce the peak heat load on the divertor. In this work, we also analyze the effect of heating power ratio on the bootstrap current, showing that adjusting the power ratio can change the bootstrap current fraction, and we further analyze the long-pulse operating region of EAST with a plasma current of 500 kA. In the range of 9.5 MW total heating/current driving power, &lt;inline-formula&gt;&lt;tex-math id="M7"&gt;\begin{document}$ {H}_{98} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M7.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M7.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; is 1.0–1.4, and normalized electron density &lt;inline-formula&gt;&lt;tex-math id="M8"&gt;\begin{document}$ {f}_{{\rm{n}}{\rm{G}}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M8.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="11-20230364_M8.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt; is 0.8–1.0, high-performance long-pulse or fully non-inductive steady-state operation can be achieved, supporting the research on the physics of ITER and CFETR steady-state operation modes. In general, improving the plasma confinement performance can achieve fully non-inductive operation at lower heating/driving power while maintaining the same plasma parameters, and expand the plasma operating regime, which is the most effective way to achieve high-parameter steady-state operation of the plasma.

Optimizations of CFETR steady state H-mode scenario with localized reversed shear enhanced internal transport barrier

Both a fully noninductive steady state operation scenario and a hybrid scenario with fusion power ∼ 1 GW and fusion gain &amp;gt;10 are being considered to fulfill the mission of a Chinese fusion engineering testing reactor. Compared to the hybrid scenario, plasma current is generally lower in steady state operation, so that better confinement and stabilization of MHD instability introduced by higher normalized beta (possibly beyond the ideal MHD limit without a wall) are required to achieve the same fusion performance. Integrated modeling is used to find candidate scenarios to match both these requirements at the same time. By creating a localized strong reversed magnetic shear using radio frequency wave driven current, a strong off-axis internal transport barrier is formed, so that the target fusion power and fusion gain are achieved for Chinese fusion engineering testing reactor steady state operation. Further optimizing the location of the reversed magnetic shear by modifying radio frequency wave launch parameters can keep the normalized beta below the ideal MHD no-wall limit while the fusion power remains beyond 1 GW. Based on this finding, several combinations of heating and current drives are proposed with fusion gain close to 12.5.

Physics design of new lower tungsten divertor for long-pulse high-power operations in EAST

A new lower tungsten divertor has been developed and installed in the EAST superconducting tokamak to replace the previous graphite divertor with power handling capability increasing from <2 MW m−2 to ∼10 MW m−2, aiming at achieving long-pulse H-mode operations in a full metal wall environment with the steady-state divertor heat flux of ∼10 MW m−2. A new divertor concept, ‘corner slot’ (CS) divertor, has been employed. By using the ‘corner effect’, a strongly dissipative divertor with the local buildup of high neutral pressure near the corner can be achieved, so that stable detachment can be maintained across the entire outer target plate with a relatively lower impurity seeding rate, at a separatrix density compatible with advanced steady-state core scenarios. These are essential for achieving efficient current drive with low-hybrid waves, a low core impurity concentration and thus a low loop voltage for fully non-inductive long-pulse operations. Compared with the highly closed small-angle-slot divertor in DIII-D, the new divertor in EAST exhibits the following merits: (1) a much simpler geometry with integral cassette body structure, combining vertical and horizontal target plates, which are more suitable for actively water-cooled W/Cu plasma facing components, facilitating installation precision control for minimizing surface misalignment, achieving high engineering reliability and lowering the capital cost as well; (2) it has much greater flexibility in magnetic configurations, allowing for the position of the outer strike point on either vertical or horizontal target plates to accommodate a relatively wide triangularity range, δ l = 0.4–0.6, thus enabling to explore various advanced scenarios. A water-cooled copper in-vessel coil has been installed under the dome. Five supersonic molecular beam injection systems have been mounted in the divertor to achieve faster and more precise feedback control of the gas injection rate. Furthermore, this new divertor allows for double null divertor operation and slowly sweeping the outer strike point across the horizontal and vertical target plates to spread the heat flux for long-pulse operations. Preliminary experimental results demonstrate the ‘corner effect’ and are in good agreement with simulations using SOLPS-ITER code including drifts. The EAST new divertor provides a test-bed for the closed divertor concept to achieve steady-state detachment operation at high power. Next step, a more closed divertor, ‘sharp-cornered slot’ divertor, building upon the current CS divertor concept, has been proposed as a candidate for the EAST upper divertor upgrade.

Mathematical modeling using batch scheduling approach for optimal tritium inventory in the ISS of the ITER fuel cycle

An optimal model to schedule fuel regeneration is developed for the Tokamak Cryopumps and Cryogenic Viscous Compressors (CVCs) of the fuel cycle of ITER. The objective is to minimize the tritium inventory inside the Isotope Separation Systems (ISS) while supplying the tokamak’s required amount of fuel considering fueling stages: ramp-up, flat-top, and ramp-down for each tokamak operation mode (Inductive, Hybrid, and Non-Inductive). Case studies of CVC regeneration times are presented to illustrate how CVC regeneration affects the ISS tritium working inventory. Results showed that nine iterations of CVC regeneration are enough to maintain ISS working tritium inventory management during Non-Inductive tokamak operation mode. The developed model uses mixed integer linear programming of the state-task network.

Neutral beam injection for DEMO alternative scenarios

Within the current EU pre-conceptual design activities, the reference demonstrative fusion power plant concept relies on ITER-like, pulsed H-mode plasmas (“DEMO1”). Simultaneously, alternative design options are being considered. An example is “Flexi-DEMO”, which is designed for steady-state plasma discharges, but can also be run pulsed at the same fusion power with the same auxiliary heating systems if the plasma confinement time turns out to be not sufficient for non-inductive operations. Preliminary investigations have started also for ELM-free regimes, such as I-mode or QH-mode, due to the issues encountered in ELMs active control at reactor scales. Neutral beam injection (NBI) is one of the candidates for a DEMO heating and current-drive system. Depending on the system design, NBI can be optimized for e.g. central heating (as for DEMO1), current-drive (as for Flexi-DEMO) or even to provide torque to the plasma (which may favour the access to ELM-free QH-mode regime). In the present contribution we describe the main NBI features in terms of design parameters (NBI energy, injection geometry) that would optimize these alternative DEMO plasma scenarios, in particular regarding Flexi-DEMO and QH-mode regime. Wide-range parametric scans through METIS numerical simulations of DEMO plasmas indicate the optimal design windows for NBI system in each plasma configuration. These results need however further validations against technical constraints, available technology limits and detailed plasma transport simulations.

Dynamic modeling of pellet production process for pellet injection system in ITER

Abstract This article proposes a dynamic simulation model of the pellet production process for the pellet injection system (PIS) in the fusion fuel cycle. The integrated devices in the pellet production process should be designed and operated to meet safety limitations while ensuring the required quantity of fuel supply. The presented model is used to investigate the variation in fuel inventory of the system under time-varying tokamak fuel demand. The model applies to the T2 pellet production in three different operation scenarios: inductive, hybrid and non-inductive operation. The simulation results suggest the reliable storage capacity of the main devices under a wide range of changing operating conditions. Preliminary analysis of design and operation options is expected to provide useful insights into the development of PIS.

Mathematical Modeling to Minimize Tritium Inventory of the Fuel Cycle of ITER

This paper proposes a mathematical model for the optimal operation of the fuel cycle of ITER. The developed model aims to minimize the tritium inventory in the fuel cycle by adopting a two-phased scheduling approach. To consider multiple equipment in the fuel cycle, the proposed solving algorithm is designed as sequential scheduling models: (Phase I) to minimize tritium inventory in the vacuum roughing system and (Phase II) to minimize tritium working inventory in the isotope separation system. The scheduling models are developed based on a state-task-network method. Given a required amount of tritium for fueling scenarios considering ramp-up, flat-top, and ramp-down, the proposed model provides the optimal operation plan for deuterium-tritium plasma operation including information on fueling rate, duration, and timing between each unit. Among six case studies, the noninductive tokamak operation mode with high tritium demands showed the highest tritium working inventory during one burn-and-dwell cycle.

Exploring fusion-reactor physics with high-power electron cyclotron resonance heating on ASDEX Upgrade

The electron cyclotron resonance heating (ECRH) system of the ASDEX Upgrade tokomak has been upgraded over the last 15 years from a 2 MW, 2 s, 140 GHz system to an 8 MW, 10 s, dual frequency system (105/140 GHz). The power exceeds the L/H power threshold by at least a factor of two, even for high densities, and roughly equals the installed ion cyclotron range of frequencies power. The power of both wave heating systems together (>10 MW in the plasma) is about half of the available neutral beam injection (NBI) power, allowing significant variations of torque input, of the shape of the heating profile and of Qe/Qi, even at high heating power. For applications at a low magnetic field an X3-heating scheme is routinely in use. Such a scenario is now also forseen for ITER to study the first H-modes at one third of the full field. This versatile system allows one to address important issues fundamental to a fusion reactor: H-mode operation with dominant electron heating, accessing low collisionalities in full metal devices (also related to suppression of edge localized modes with resonant magnetic perturbations), influence of Te/Ti and rotational shear on transport, and dependence of impurity accumulation on heating profiles. Experiments on all these subjects have been carried out over the last few years and will be presented in this contribution. The adjustable localized current drive capability of ECRH allows dedicated variations of the shape of the q-profile and the study of their influence on non-inductive tokamak operation (so far at q95 > 5.3). The ultimate goal of these experiments is to use the experimental findings to refine theoretical models such that they allow a reliable design of operational schemes for reactor size devices. In this respect, recent studies comparing a quasi-linear approach (TGLF) with fully non-linear modeling (GENE) of non-inductive high-beta plasmas will be reported.

Progress of physics understanding for long pulse high-performance plasmas on EAST towards the steady-state operation of ITER and CFETR

Recently, the first ever 100 s long, steady-state H-mode discharge with good control of impurities, core and edge MHD stabilities, and heat exhaust was demonstrated in the Experimental Advanced Superconducting Tokamak (EAST) using the ITER-like (International Tokamak Experimental Reactor) tungsten upper divertor. Using both radio frequency (RF) power and neutral beam injection (NBI) heating, EAST has demonstrated fully non-inductive scenarios with an extension of fusion performance at high density and low rotation: β P ∼ 2.5, β N ∼ 2.0, H98,y2 ∼ 1.2, bootstrap current fraction fBS ∼50% at q95 ∼ 6.8. With pure RF power heating, plasmas have been maintained for up to 21 s (over 40 times the current relaxation time) with zero loop voltage and small edge localized modes (ELMs) at high density (ne/nGW ∼ 0.6–0.8), β P ∼ 2.0, β N ∼ 1.6, and ƒBS ∼47%. Experimental investigations show how plasma current profiles, turbulent transport and radiation properties self-consistently evolve toward fusion relevant steady state conditions. Modeling and physics experiments have confirmed the synergistic effects between electron cyclotron heating (ECH) and low hybrid wave (LHW), where ECH enhances the heating and current drive from LHW injection, enabling fully non-inductive operation at higher density. Small/no ELMs facilitate the RF power coupling in the H-mode phase and reduce divertor erosion. A low tungsten concentration was observed at high β P with a hollow profile in the core. Reduction of the peak divertor heat flux with f rad of up to 40% was compatible with the high β P scenario by using active radiation feedback control. With features such as dominant electron heating, zero/low NBI torque and an ITER-like tungsten divertor, fully non-inductive high-performance experiments on EAST offer unique contributions towards the succesful operation of ITER and CFETR (the Chinese Fusion Engineering Testing Reactor).

Derivation of bootstrap current fraction scaling formula for 0-D system code analysis

We constructed new scaling formulas for the bootstrap current faction fBS for tokamaks with full non-inductive operation with positive shear safety factor q profiles. A fitting database was built by the ACCOME code, which contains 12,150 cases of self-consistent equilibria with the q profiles satisfying qρ≥1.0 and q95≥3.0 in 0≤ρ<1, where ρ is the normalized minor radius and q95 is q at the 95% poloidal flux surface. From the fitting database, we derived the formulas by the multiple regression analysis. The best formula contained the inverse aspect ratio ε, the poloidal beta, the internal inductance, q95/q0, and the profile shape indices for density and temperature as input. The Root Mean Squared Error for the fitting database of this formula was 0.025. We tested the derived formulas to three kinds of datasets and compared with the K. Gi models and the ITER Physics Design Guidelines model. The first test is for the compact neutron source with the same plasma size with the fitting database, the second (including reversed shear) and the third ones (positive shear only) are for the conventional reactor size. The best formula seems to be applicable to tokamak plasmas with a wide range of parameters.

Overview of physics studies on ASDEX Upgrade

The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q95 = 5.5, ) at low density. Higher installed electron cyclotron resonance heating power 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG.Stable high-density H-modes with MW m−1 with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently . This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated.Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of and Er allow for inter ELM transport analysis confirming that Er is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of ‘natural’ no ELM regimes have been extended. Stable I-modes up to have been characterised using -feedback.Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle—measured for the first time—or the cross-phase angle of and fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvénic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO.

Extension of the operating space of high-bN fully non-inductive scenarios on TCV using neutral beam injection

The fully non-inductive sustainment of high normalized beta plasmas () is a crucial challenge for the steady-state operation of a tokamak reactor. In order to assess the difficulties facing such scenarios, steady-state regimes have been explored on the tokamak à configuration variable (TCV) using the newly available 1 MW neutral beam injection (NBI) system. The operating space is extended towards plasmas that are closer to those expected in JT-60SA and ITER, i.e. with significant NBI and electron cyclotron resonance heating and current drive (ECRH/CD), bootstrap current and fast ion (FI) fraction. values up to 1.4 and 1.7 are obtained in lower single null L-mode () and H-mode () plasmas, respectively, at zero time averaged loop voltage and . Fully non-inductive operation is not achieved with NBI alone, whose injection can even increase the loop voltage in the presence of EC waves. A strong contribution to the total plasma pressure of thermal and FIs from NBI is experimentally evidenced and confirmed by interpretative ASTRA and NUBEAM modeling, which further predicts that FI charge-exchange reactions are the main loss channel for NBH/CD efficiency. Internal transport barriers, which are expected to maximize the bootstrap current fraction, are not formed in either the electron or the ion channel in the plasmas explored to date, despite a significant increase in the toroidal rotation and FI fraction with NBI, which are known to reduce turbulence. First results on scenario development of high- fully non-inductive H-mode plasmas are also presented.

Observation of a fully non-inductive H-mode regime dominated by the sporadic-small edge-localized modes in EAST with a tungsten divertor

Here we report a fully non-inductive H-mode operation with a sporadic-small-edge-localized mode(ELM)-dominant pedestal realized in EAST with a tungsten divertor, where the normalized electron collisionality at the pedestal top is estimated below unity. This is regarded as the expansion and/or supplement of our previous research having been preliminarily reported (section 4.2 in (B N Wan et al 2017 Nucl. Fusion 57 102019)) in the EAST overview article contributed to the special issue on FEC 2016 summaries and overviews. In this operation, spontaneous avoidance of large ELMs is realized reproducibly, for the first time on EAST, at a lower collisionality (below unity) at the pedestal top, profiting from high heating power. Besides, this regime shows the well controlled bulk plasma density and light impurity concentration, good energy confinement with H98, (y2) ≃ 1, together with a loop voltage very close to zero. All of these features are considered critical to a steady-state operation with high performance, which EAST is on target to demonstrate in the near future in support to the International Thermonuclear Experimental Reactor (ITER). However, this operation regime suffers a continuous accumulation of medium-/high-Z impurities, which may seriously limit its sustainability. The resonant magnetic perturbation (RMP) with a dominant toroidal mode number n = 1 has been employed for assisting exhaust of heavy impurities, which is, unfortunately, not successful since the RMP with an effective poloidal spectrum always brings back large ELMs. This scenario transits to an ELM H-mode regime as plasma current is decreased, and elimination of large ELMs is well reproduced by ELITE simulation. Finally, we present a discussion on the potential mechanism eliminating large ELMs, which seems to highlight the LHW-induced profound change in edge magnetic topology.