Advanced Tokamak Research Articles

Effect of Specific Pressure of Fast Particles on Double-Tearing Mode of Resistance Magnetic Fluid

In the advanced Tokamak magnetic field configuration, the instability of the double tear mode excited by the diamagnetic shear configuration plays an important role. At the same time, due to the fusion reaction and auxiliary heating means, a large number of fast particles will be generated. The existence of fast particles will affect the growth of double tear mode. In the context of resistive magnetic fluids, the role of fast particle specific pressure in the instability of the double tear mode was analyzed, and the CLT-K program was used to simulate the instability of the fast particle specific pressure to the double-tearing mode to explore the physical mechanism of its impact.

Discharges with Reduced-Volume Plasma Filament in T-15 Upgrade Tokamak

The formation in the T-15 Upgrade tokamak of plasma filaments with a smaller volume than in the base discharge scenario, keeping the total plasma current constant, has been studied by numerical simulation. The purpose of such operating regimes is to raise the specific power of additional plasma heating. These regimes are of interest for attaining highest possible temperatures under limited additional heating power, e.g., for attaining AT (Advanced Tokamak) regimes. Additionally, these operating regimes can be investigated in the early stage of the T-15 Upgrade experiments when the total available heating power is low. Calculations with the Tokameq code have produced stationary equilibrium configurations for reduced plasma volume regimes. Three variants of plasma-volume reduction have been considered: displacing the plasma filament to the strong magnetic field region near the inner chamber wall while maintaining the filament center in the equatorial plane, and also displacing the plasma filament up or down. Numerical calculations have shown that such regimes may reduce the plasma-filament volume by a factor of 2 to 5, thus significantly raising the specific power of additional heating.

Stability nature of ohmic discharges with hollow current profiles on the HL-2A tokamak

The equilibrium configurations with hollow current density profiles can effectively improve plasma confinement. The stability of the ohmic discharges with hollow current density profiles on the HL-2A tokamak was studied for various parameters, such as plasma poloidal beta βp, the axial and edge safety factors q0 and qs, as well as the minimum safety factor . The results suggest that the restriction to the parameters of discharges with hollow current density profiles is stricter than that of conventional discharges with peaked current density profiles. In the case of q profile with reverse shear, the instability of the configuration is significantly improved with higher βp. As a result, the configuration with a hollow toroidal current profile is stable for high βp operations. Only if q0, qs and satisfy the specific limitations, i.e. on HL-2A, only if both and hold, together with and , may the configurations with the reverse shears be stable for hollow-current-type Ohmic discharge. The results may give hints for the role of the discharges with hollow toroidal current densities in the coming advanced tokamak operations.

Increased core ion temperatures in high-beta advanced scenarios in ASDEX upgrade

Non-inductive advanced tokamak scenarios are a possible way for future nuclear fusion power plants to run in non-pulsed operation. In these scenarios, the ohmic current is replaced on the one hand with a current driven by external sources such as NBI and ECRH and on the other hand with a substantial bootstrap-current. The bootstrap current is produced in the presence of pressure gradients. This means, to increase the bootstrap-current-fraction, it is advantageous to have regions where ITG turbulence is reduced. To be able to extrapolate from non-inductive scenarios done in smaller present day devices, it is important to have transport models that allow the user to reproduce these experiments and accurately capture the physics behind the reduction of turbulent transport. One commonly used model is the quasi-linear gyrofluid transport model TGLF, which is well tested in standard scenarios. However, in the past, TGLF has failed to reproduce the peaked ion temperature profiles of certain advanced scenario shots.In this article, progress in overcoming this issue of TGLF will be discussed. Results of a recent publication are reproduced in which TGLF was able to match the experimentally measured peaked ion temperature profiles. In these simulations, we find the E × B-shear to play an important role in the reduction of ITG turbulence. In contrast to that, experimentally we find the E × B-shear to have no effect on the formation of such regions of decreased turbulent transport. This finding is in line with simulations using the gyrokinetic code GENE. A new approach of modelling these advanced scenarios in TGLF is introduced, allowing us to match the peaked ion temperature profiles without depending on the E × B-shear.

First Evidence of Local E×B Drift in the Divertor Influencing the Structure and Stability of Confined Plasma near the Edge of Fusion Devices.

The structure of the edge plasma in a magnetic confinement system has a strong impact on the overall plasma performance. We uncover for the first time a magnetic-field-direction dependent density shelf, i.e., local flattening of the density radial profile near the magnetic separatrix, in high confinement plasmas with low edge collisionality in the DIII-D tokamak. The density shelf is correlated with a doubly peaked density profile near the divertor target plate, which tends to occur for operation with the ion B×∇B drift direction away from the X-point, as currently employed for DIII-D advanced tokamak scenarios. This double-peaked divertor plasma profile is connected via the E×B drifts, arising from a strong radial electric field induced by the radial electron temperature gradient near the divertor target. The drifts lead to the reversal of the poloidal flow above the divertor target, resulting in the formation of the density shelf. The edge density shelf can be further enhanced at higher heating power, preventing large, periodic bursts of the plasma, i.e., edge-localized modes, in the edge region, consistent with ideal magnetohydrodynamics calculations.

An efficient MHD equilibrium solver for control oriented transport models

In tokamak plasmas, the current density profile is a strong determiner of both MHD stability and confinement. The design of active control techniques for this profile relies on control oriented transport models that capture the essential dynamics of its evolution. The physics of transport is coupled to the MHD equilibrium that determines the spatial structure of the magnetic configuration. However, since this coupling is difficult to solve, the common practice is not to update the equilibrium condition during profile control design or current profile tracking. In this work, a simplified coupling scheme for transport and equilibrium suitable for internal profile control design is proposed. The scheme relies on a new MHD equilibrium solver that efficiently updates the magnetic configuration using the safety factor and pressure profiles computed by the control oriented transport code. The new solver is described in detail and validated in advanced tokamak conditions including challenging features like reversed magnetic shear in the plasma core, edge transport barrier and high plasma beta.

Design of ECH launcher for KSTAR advanced Tokamak operation

For the advanced tokamak operation in Korea Superconducting Tokamak Advanced Research (KSTAR), up to 6 MW of electron cyclotron heating (ECH) is planned. The ECH launcher should be able to operate in steady state and serve as a central/off-axis heating and current drive, q-profile control, and control of the neoclassical tearing mode (NTM) and sawtooth. The water cooled launchers have been upgraded to include continuous operation with precise, high-speed real-time control of the beams by integration with the plasma control system (PCS) for supporting the experiments, in particular for suppressing NTM. In this study, the launcher was successfully operated for 78 s with about 650 kW EC power, where the temperature of the coolant, with a flow rate of 9.6 lpm, is saturated and approximately 2 °C. The launcher control system had a maximum control cycle of 5 kHz and a motor latency of about 10 ms, and the poloidal scanning speed of the beam was 20°/s. Additionally, for the higher localization of power deposition, the curvature and size of the mirrors in the launcher were optimized. This paper presents the ECH launcher design and its control system with the operation results.

Influence of the safety factor profile on the particle and heat transport in the Globus-M spherical tokamak

The advanced tokamak scenario is a promising operation scenario for ITER and fusion neutron sources. In this scenario the minimum value of the safety factor in the center of the plasma exceeds unity. In the compact spherical tokamak Globus-M, the formation of such conditions is possible with neutral beam injection at the current ramp-up phase. Due to the slower diffusion of current inside the plasma, a zone is formed with reduced heat and particle transport across the magnetic field, which affects the temperature and density profiles of the plasma. This leads to the peaked density profile formation and improvement of the energy confinement time. To achieve a high fraction of the bootstrap current, it is necessary to increase the plasma pressure. At the same time, the maximum allowable pressure is limited to the normalized beta limit.

Overview of HL-2A recent experiments

The mission of HL-2A is to explore the key physical topics relevant to ITER and to the advanced tokamak operation (e.g. the operation of future HL-2M), such as the access of H-mode, energetic particle physics, edge-localized mode (ELM) mitigation/suppression and disruption mitigation. Since the 2016 Fusion Energy Conference, the HL-2A team has focused on the investigations on the following areas: (i) pedestal dynamics and L–H transition, (ii) techniques of ELM control, (iii) turbulence and transport, (iv) energetic particle physics. The HL-2A results demonstrated that the increase of mean shear flow plays a key role in triggering L–I and I–H transitions, while the change of flow is mainly induced by the ion pressure gradient. Both mitigation and suppression of ELMs were realized by laser blow-off seeded impurity (Al, Fe, W). The 30% Ne mixture supersonic molecular beam injection seeding also robustly induced ELM mitigation. The ELMs were mitigated by low-hybrid current drive. The stabilization of m/n = 1/1 ion fishbone activities by electron cyclotron resonance heating was found on the HL-2A. A new m/n = 2/1 ion fishbone activity was observed recently, and the modelling indicated that passing fast ions dominantly contribute to the driving of 2/1 fishbone. The non-linear coupling between the toroidal Alfven eigenmode and tearing mode (TM) led to the generation of a high frequency mode with the toroidal mode number n = 0. The turbulence was modulated by TM when the island width exceeds a threshold and the modulation is localized merely in the inner area of the islands. Meanwhile, turbulence radial spread took place across the island region.

A high field side multijunction launcher with aperture impedance matching for lower hybrid current drive in DIII-D advanced tokamak plasmas

A high field side (HFS) lower hybrid current drive launch scenario improves wave accessibility, has single pass damping at r/a ~ 0.6–0.8 and good current drive efficiency in DIII-D advanced tokamak discharges. The DIII-D experiment is an opportunity to validate HFS wave propagation, absorption and scrape-off layer benefits. A HFS multi-junction launcher is designed and simulated in COMSOL over a range of plasma edge conditions to evaluate n|| spectrum, directivity, and return loss. The COMSOL model utilizes a lossy anisotropic dielectric modeled with the cold plasma dispersion relation cross validated against ALOHA and Petra-M codes. The COMSOL model seamlessly includes a realistic plasma model and coupler that allows for rapid optimization of a single launcher module, while Petra-M allows more complex simulation of an eight-module array including curvature, and warm plasma effects. The resulting design utilizes a traveling wave poloidal power divider to minimize peak electric fields in the vacuum region of the coupler, and an internal aperture matching structure provides an impedance match to the plasma over a wide range of plasma density and density gradient edge conditions.

First experimental tests of a new small angle slot divertor on DIII-D

A new small angle slot (SAS) divertor concept has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Initial tests on DIII-D find a strong interplay between such anticipated ‘SAS’ effects and cross-field drifts, favouring operation with the ion B × ∇B drift away from the X-point, as currently employed for advanced tokamaks. This offers the following key improvements relative to DIII-D’s open lower divertor or partially-closed upper divertor: (i) SAS allows for transition to low temperature moderately detached divertor conditions with Te ≲ 10 eV at very low main plasma densities, lower than are usually attainable at all in DIII-D high confinement (H-mode) plasmas as used in these tests; (ii) Pedestal performance and core confinement are significantly improved with SAS. The final confinement collapse associated with the onset of X-point MARFE (multifaceted asymmetric radiation from the edge) following deep detachment occurs at significantly higher pedestal densities, thus widening the window of H-mode operation compatible with a dissipative divertor. For operation with the ion B × ∇B drift toward the X-point, the divertor plasma transitions to a bifurcative detached state at much higher densities, similar to other divertor configurations in DIII-D. These results highlight the strong interplay between divertor closure and drifts, and point to an interesting divertor optimization path to explore that offers potential for future fusion reactors.

Experimental and numerical study of the different quench acceleration phenomena in the JT-60SA Toroidal Field coils

JT-60SA is an advanced superconducting fusion Tokamak jointly constructed by Japan and Europe. It takes the main missions of addressing key physics issues and providing direct operation experiences for ITER and DEMO reactors. In the framework of the JT-60SA project, the 18 NbTi superconducting Toroidal Field (TF) coils have been tested in quench conditions at CEA Paris-Saclay. In our previous study of these coils tests, a typical quench behavior has been identified and analyzed. This behavior shows for the majority of the coils (13 tests over 19), four dynamic phases including a quench initiation phase with a velocity around 3 m/s which is rapidly (several hundreds of milliseconds) followed by a quench acceleration one at around 30 m/s near the start of the current discharge. This is called "early" quench acceleration. Nevertheless, a few number of coils showed a quench acceleration with a certain delay of about 0:5 s to 2 s after the current discharge. This paper will propose a study of this "delayed" quench acceleration phenomenon in two steps: Firstly, a physical analysis of the experimental data of these few tests; Secondly, a numerical study with the code THEA analyzing the testing conditions impact on the beginning delay of the quench acceleration phase.

Status, scientific results and technical improvements of the NBH on TCV tokamak

The TCV tokamak contributes to physics understanding in fusion reactor research by a wide set of experimental tools, like flexible shaping and high power ECRH. A 1MW, 25 keV deuterium heating neutral beam (NB) has been installed in 2015 and it was operated from 2016 in SPC-TCV domestic and EUROfusion MST1 experimental campaigns (˜50/50%). The rate of failures of the beam is less than 5%.Ion temperatures up to 3.5 keV have been achieved in ELMy H-mode, with a good agreement with ASTRA predictive simulations. The NB enables TCV to access ITER-like βN values (1.8) and Te/Ti ˜1, allowing investigations of innovative plasma features in ITER relevant ELMy H-mode. The advanced Tokamak route was also pursued, with stationary, fully non-inductive discharges sustained by ECCD and NBCD reaching βN˜1.4–1.7.Real-time control of the NB power has been implemented in 2018 and presented together with the statistics of NB operation on the TCV. During commissioning, the NB showed unacceptable heating of the TCV beam duct, indicating a higher power deposition than expected on duct walls. A high beam divergence has been found by dedicated measurement of 3-D beam power density distribution with an expressly designed device (IR measurement on tungsten target).

Top Launch for Higher Off-axis Electron Cyclotron Current Drive Efficiency

Efficient off-axis current drive is crucial for economic, steady-state tokamak fusion power plants. “Top Launch” electron cyclotron current drive (ECCD) is one promising method for driving strong off-axis current to achieve the desired broad current profile for the Advanced Tokamak regime. New simulations show that by launching the electron cyclotron (EC) waves downwards (or upwards) nearly parallel to the resonance plane with a large toroidal steering, higher current drive efficiency can be obtained at large radii owing to the large Doppler shift, wave-particle interactions on HFS of the plasma, and the long absorption path. Implementations on CFETR and DIII-D are simulated identifying clear benefits and optimal configurations. The design of a prototype test being installed on DIII-D is presented, which will experimentally validate whether top launch ECCD can be an improved efficiency current drive technique for future fusion reactors.

Study of off-axis current drive using EC top-launch in KSTAR

The off-axis current drive using electron cyclotron (EC) wave is typically very low due to the low density and temperature and electron trapping effect when it is launched from outside midplane. However, the heating and current drive by EC wave is being regarded as a essential element as an off-axis current drive source together with lower hybrid current drive (LHCD) for advanced tokamak operation research in KSTAR in future. Therefore, the reliable and high efficient ECCD using top launch has been studied for two different launch schemes of down-shift and up-shift resonance with various EC frequencies which will be available in the future KSTAR ECRH system. The ray tracing (GENRAY) simulation studies show that a broad ECCD profile peaked off axis is obtained with a high current drive efficiency for both schemes. It is almost twice as high as that of the outside midplane launch for the fundamental O-mode EC wave.

DIII-D Research to Prepare for Steady State Advanced Tokamak Power Plants

We review progress made on the advanced tokamak path to fusion energy by the DIII-D National Fusion Facility (Luxon et al. in Nucl Fusion 42:614, 2002). The advanced tokamak represents a highly attractive approach for a future steady state fusion power plant. In this concept, there is a natural alignment between high pressure operation, favorable stability and transport properties, and a highly self-driven (‘bootstrap’) plasma current to sustain operation efficiently and without disruptions. Research on DIII-D has identified several promising plasma configurations for fully non-inductive operation with potential applications to a range of future devices, from ITER to nuclear science facilities, to compact or large scale fusion power plants. Significant progress has been made toward realizing these scenarios, with the demonstration of high β access, off-axis current drive techniques, model based profile control, and stability and ELM control in reactor relevant physics regimes. Radiative techniques have also been pioneered to develop improved compatibility with divertor requirements, and simultaneous access to high performance pedestals. Research has also developed major advances in physics understanding, validating concepts of kinetic damping of ideal MHD instabilities that enable high β operation, identifying how current profile and β influence plasma turbulence in order to validate and improve turbulent transport models, and understanding the physics of energetic particle redistribution due to Alfvénic and other instabilities. These advances have been partnered with development of a rigorous integrated modeling framework used to interpret and validate individual physics models of the various aspects of plasma behavior, and to guide development of improved regimes and upgrades. These tools are also being used to develop and validate concepts for future reactors directly. Having established these foundations, DIII-D is now undergoing a substantial upgrade to raise power, current drive, electron heating and 3-D field capabilities in order to validate this physics and test conceptual solutions in reactor-relevant physics regimes, with a goal to resolve the key scientific and technology questions to enable a decision on a future steady state fusion power plant.

Considerations of the q-profile control in KSTAR for advanced tokamak operation scenarios.

The q-profile control is essential for tokamaks exploring the advanced tokamak scenarios, which is expected to be able to provide a possible route toward a steady-state high performance operation in a fully non-inductive current drive state. This is because the pressure and current profiles must remain optimal for the scenario during the injection of large amounts of heating and current drive. Here, essential tools for the q-profile control are the motional Stark effect diagnostic for measuring the radial magnetic pitch angle profile and a state-of-the-art plasma control system. The pulse duration of the H-mode discharge at KSTAR has been extended year by year with improved control performance, and the experiment of internal transport barrier (ITB) formation in a weakly reversed q-profile with a marginal neutral beam injection majority heating successfully demonstrated that the ITB is an alternative candidate to achieve a high performance regime in KSTAR. These recent achievements are attributed to reliable profile measurement, which means that profile feedback control has become a necessary step to ensure a robust and reliable approach to advanced scenarios as the next step of research in KSTAR. In this paper, we discuss the technical and conceptual requirements for q-profile control according to the upgrade plan for heating and current drive systems in the coming years.

Drift effects and up–down asymmetry in balanced double‐null DIII‐D divertor configurations

In a preliminary design study for a small‐angle slot (SAS) divertor configuration suitable for DIII‐D double‐null advanced tokamak (AT) plasmas, a conceptual “SAS 2” slot divertor configuration has been modelled with SOLPS‐ITER. A balanced double‐null equilibrium, representing an H‐mode AT plasma, is employed, with 5 MW neutral beam input power. In simulations with fully recycling walls and no pumping, an up–down‐symmetric open divertor geometry is compared with the SAS 2 configuration. In density scans without electric or magnetic drifts, the upstream separatrix electron density needed to achieve detachment at both target strike points (ne, sep, det.) is ∼35% lower in SAS 2 than in the open configuration, with ne, sep, det. = 1.66 and 2.59 × 1019 m−3, respectively. With drifts, the same ∼35% difference remains, but detachment density for each configuration is shifted up by ∼15%, giving ne, sep, det. = 1.94 and 2.99 × 1019 m−3, respectively. Cryopumping is modelled for the SAS 2 divertor, and with a pump duct deep in the upper slot on the common‐flux side, detachment density is shifted up by ∼15–20%. Strong pumping in the upper slot delays detachment there compared with the lower slot, regardless of drift direction. This research sets the stage for future modelling that would explore more sophisticated cryopumping configurations, and sensitivity of divertor performance to changes in magnetic balance and input power.

Three Confinement Systems—Spherical Tokamak, Standard Tokamak, and Stellarator: A Comparison of Key Component Cost Elements

Since the 1950s “next-step” fusion devices and power plant studies have been developed for a number of magnetic confinement systems but an open question remains: can a magnetic fusion device be simplified to the point where it will be cost competitive and operate with high availability? Concept designs based on the ARIES advanced tokamak, spherical tokamak (ST), and the quasi-axisymmetric (QAS) stellarator option have progressed in recent years through a series of Princeton Plasma Physics Laboratory (PPPL) studies with an underlying intent to improve the engineering feasibility of each, giving special attention to concepts that simplify the device configuration and improve maintenance features. For the ST option, design details centered on a 3-m Fusion Nuclear Science Facility that evolved to incorporate vertical maintenance, high temperature superconductor magnets, a small inboard duel coolant lead lithium blanket, and a liquid metal divertor. In collaboration with the Korean fusion demonstration reactor (K-DEMO) and CFETR concept study teams the tokamak design has evolved to increase plasma component access within a vertical maintenance approach using enlarged toroidal field coils incorporating a low- and high-field Nb3Sn winding pack that provide a peak field of 16 T. A recent PPPL stellarator study focused on simplifying the stellarator winding topology to improve access to in-vessel components; combining coil optimization with winding surfaces that incorporated geometry constraints specified by engineering. This paper centered on a 1000-MW power plant design with a tokamak like vertical maintenance scheme that allows access to remove large internal blanket sectors. Results of three confinement studies (PPPL developed ST, K-DEMO, and QAS stellarator) will be presented to highlight concepts that simplify each device configuration and improved their maintenance features. Scaling each option to a common 1000-MW net electric power plant mission allows comparisons to be made of key cost elements such as the size of major core components, sizing of the reactor hall, or external facilities needed to handle and store activated in-vessel components.

Exploring new directions for advanced tokamaks reveals promising potential

An all-metal ASDEX Upgrade facility opens new pathways for upcoming fusion demonstration plants and explores new scenarios for long-pulse operation.