Core Plasma Density Research Articles

Recent progress in improvement in ion cyclotron range of frequencies coupling and power absorption with new antennas of Experimental Advanced Superconducting Tokamak (EAST)

Efficient ion cyclotron range of frequencies (ICRF) wave heating requires good wave coupling at the plasma edge and good radio frequency power absorption in the plasma core. This study reviews recent progress in improving these two aspects of ICRF heating with the new two-strap antennas through various experiments and simulations on the Experimental Advanced Superconducting Tokamak (EAST). Our study shows that the ICRF coupling can be significantly improved by decreasing the parallel wave number, increasing the local scrape-off layer (SOL) density by midplane gas puffing, and increasing the global SOL density by decreasing the separatrix–antenna distance. It can also be improved by increasing the core plasma density, changing the divertor strike point position, and optimizing the antenna phasing. The core ICRF power absorption can be increased by optimizing the cyclotron resonance position and minority ion concentration and by applying new heating schemes such as three-ion heating. Although some of the methods have been previously studied on other devices, improving ICRF coupling by shifting the divertor strike point was tested on EAST for the first time. Quantitative characterization of these methods and the conclusions drawn from this study can provide important insights for achieving more efficient ICRF heating in current and future fusion machines.

Suppression of equilibrium magnetic islands by density profile effect in quasi-axisymmetric stellarator plasmas

In a quasi-axisymmetric stellarator, a significant bootstrap current will result in the generation of low-order rational surfaces and three-dimensional (3D) magnetic islands. In this paper, the influence of plasma density profiles on the equilibrium magnetic islands for the Chinese first quasi-axisymmetric stellarator (CFQS) is investigated using the HINT code. It is found that the flattening of the core plasma density profile leads to a significant suppression of magnetic islands. When the peaking factor of plasma density is 1.19, complete suppression of magnetic islands occurs while maintaining excellent integrity of the magnetic surface even with the volume-averaged plasma beta <β> increase up to 2%. On the other hand, during the transition of a plasma density profile from flat to hollow, there is a reversal in the core bootstrap current, resulting in reduction of rotational transform values to pass through the rational surface. Hence, formation of magnetic islands in the core region. Therefore, effective inhibition of CFQS’s magnetic islands can be achieved by appropriately controlling density profiles through methods like gas injection.

Magnetic reconnection during sawteeth crashes

Sawteeth oscillations are periodic relaxations of the core plasma density and temperature in tokamaks. The rise of the temperature due to external heating is terminated by the crash phase, which involves magnetic reconnection. This is the case of fast magnetic reconnection in collisionless plasmas (Lundquist number S≥108) with a strong guide field. (The toroidal magnetic field in a tokamak is a few orders larger compared to the reconnected helical field.) Experimental measurements show non-linear behavior before and during the crash phase. Simplified single-fluid models are not able to explain the reconnection dynamics during the crash, and two-fluid effects have to be considered. In this case, numerical simulations give good agreement with the observations for the crash duration. At the same time, the present simulations explain experimentally observed phenomena only partially, and several questions remain an area of active research: evolution during the crash, the onset of the fast phase, the existence of the post-cursors, the degree of stochasticity, and others. This overview paper summarizes the current understanding of the crash process, highlights remaining problems, and shows connections to magnetic reconnection research in other plasmas.

Preparation of a beam emission spectroscopy diagnostic based on a Lyman-alpha line to diagnose core and edge-plasma density fluctuation on the HL-2A tokamak

In this article, the design of a Lyman-alpha-based beam emission spectroscopy (LAB) diagnostic on the HL-2A tokamak has been proposed for the first time. The purpose of this novel diagnostic is to measure density fluctuations of tokamak plasma. The light-collection system of LAB, which consists of the first mirror and two groups of coaxial double-mirror telescopes, can realize a two-segmented viewing field of ρ = 0‒0.2 and ρ = 0.75‒1, which is optimized to measure plasma density fluctuation, not only in the edge transport barrier region but also in the internal transport barrier region, to investigate the underlying physics of turbulence in tokamaks. Spectrometers are developed to separate out the Doppler-shifted target line (122.03 and 122.17 nm) from the background Lyman-alpha line (121.53 nm). Here, 30 Core-LAB channels and 30 Edge-LAB channels are under development on the HL-2A tokamak. It has high radial spatial resolutions of about 2.7 mm and 3.3 mm for the core and edge channels, respectively. Taking the high light intensity of this Lyman-alpha line into account, temporal resolution of 200 kHz can be ensured by broad bandwidth amplifiers. This high spatio-temporal resolution makes LAB a potential keen tool to experimentally investigate tokamak plasma physics.

Characterisation of divertor detachment onset in JET-ILW hydrogen, deuterium, tritium and deuterium–tritium low-confinement mode plasmas

Measurements of the ion currents to and plasma conditions at the low-field side (LFS) divertor target plate in low-confinement mode plasmas in the JET ITER-like wall materials configuration show that the core plasma density required to detach the LFS divertor plasma is independent of the hydrogenic species protium, deuterium and tritium, and a 40%/60% deuterium-tritium mixture. This observation applies to a divertor plasma configuration with the LFS strike line connected to the horizontal part of the LFS divertor chosen because of its superior diagnostic coverage. The finding is independent of the operational status of the JET cryogenic pump. The electron temperature (Te) at the LFS strike line was markedly reduced from 25 eV to 5 eV over a narrow range of increasing core plasma density, and observed to be between 2 eV and 3 eV at the onset of detachment. The electron density (ne) peaks across the LFS plasma when Te at the target plate is 1 eV, and spatially moves to the X-point for higher core densities. The density limit was found approximately 20% higher in protium than in tritium and deuterium-tritium plasmas.

H-mode dithering phase studies on ST40.

The dithering H-mode phase, characterized by oscillations, is generally observed at input power values close to the L-H transition power threshold and low plasma collisionalities (low electron density and/or high plasma temperature). Measurements to characterize the dithering phase are presented for the low aspect ratio, high magnetic field tokamak, ST40. The dithering phase oscillation frequency is observed between 400 and 800 Hz and demonstrates an inverse relationship with core plasma density. Dithering phase H-modes are documented across a nonlinear, low-density power threshold operational space, with signature low- and high-density branches. The minimum power threshold for dithering H-mode access is measured at a core, line average electron density of 4.7(±0.5) × 1019 m-3, close to a predicted value of 4.1(±0.4) × 1019 m-3 from multi-machine studies. ASTRA calculated values of power coupled to the ion species, at the dithering H-mode transition, exhibit a similar nonlinear dependence on density. This analysis points to the important contribution of the ion thermal channel to the L-H phase transition. The low-frequency plasma density and D-alpha dithers appear to be accompanied by sudden bursts of magnetohydrodynamic (MHD) activity. A simple model is tested to demonstrate a possible scenario of self-regulation among turbulence, zonal flows, pressure (density) gradient and MHD activities. This article is part of a discussion meeting issue 'H-mode transition and pedestal studies in fusion plasmas'.

Characteristics of electron temperature profile stiffness in electron-heated plasmas on EAST

A very high core electron temperature (T e0 ∼ 10 keV) plasma has been established and stably sustained by applying both lower hybrid wave (LHW) and on-axis electron cyclotron resonance heating (ECRH) in the Experimental Advanced Superconducting Tokamak (EAST). In this work, power balance analysis shows that the increase of ECRH power can increase the normalized T e gradient significantly at the plasma core region (ρ < 0.6), but does not change the T e profile stiffness in the low-density L-mode plasmas. This has been considered to be due to a strong synergistic effect between ECRH and LHW. Furthermore, three distinguishable stages characterized by different T e profile stiffnesses can be identified from the density ramp-up in the electron-heated plasma on EAST. A stronger T e profile stiffness at ρ = 0.3 has been observed in the Stage-II, where the LHW power deposition gradually moves away from the plasma core region, following the electron density increases. Furthermore, the formation of an internal plasma density transport barrier inside ρ ∼ 0.6, accompanied by a sudden drop in core T e and a rise in both core plasma density and ion temperature, has been observed for the first time during the transition from the Stage-II to the Stage-III when the central line-averaged plasma density reaches a threshold of 2.2 × 1019 m−3. This finding strongly affects further development of high-performance gas-fueled electron-heated plasma scenarios in EAST and suggests an advanced operational regime with a wide internal plasma density transport barrier.

Choice of Gas Isotope Composition for Neutral Beam Injectors of the FNS-ST Compact Fusion Neutron Source

The dependence of the neutron yield of the FNS-ST (spherical tokamak) fusion neutron source on the fraction of tritium in the core D+T plasma is analyzed for the case of using tritium neutral beam injectors with 200-keV energy and 6-MW power. The FNS-ST operating regimes are explored using the SOLPS4.3 and ASTRA codes for different values of core plasma density ne, T fraction in the plasma, and particle diffusivity. The FC-FNS code is used to estimate the fluxes of the fuel components in the fuel cycle (FC), which are produced by different injection systems: gas puffing, pellet injection, and neutral beam (T) injection. It is shown that in the case of the Т beam injection, in the operating range of parameters, the neutron yield can reach 6.0 × 1017 s−1, which is the value comparable to that obtained for the scenario of D-beam injection into the balanced D+T plasma. In the case of the T-beam injection, in the range of parameters, for which the neutron yield is close to its maximum, the amount of tritium in the FC is lower than in the case of the D-beam injection. The neutron yield can be increased to 6.5 × 1017 n/s if full separation of the D and T is introduced for the gas pumped out from the divertor and puffed back into the torus. With this approach, in the case of the tritium beam, the amount of tritium in the FC is Tinv of ~170 g. If this approach is used in the case of the deuterium beam, the neutron yield can reach 7.0 × 1017 n/s. However, in this case, the amount of tritium contained in the FC increases to 215 g. The results of the analysis performed are used for optimizing the FC of the FNS-C (compact) fusion neutron source, which is planned for construction in the framework of the comprehensive program of the State Corporation Rosatom “Development of Engineering, Technology and Scientific Research in the Field of Using Atomic Energy in the Russian Federation for the Time Period up to 2030.”

Anisotropic Surface Broadening and Core Depletion during the Evolution of a Strong-Field Induced Nanoplasma

Strong-field ionization of nanoscale clusters provides excellent opportunities to study the complex correlated electronic and nuclear dynamics of near-solid density plasmas. Yet, monitoring ultrafast, nanoscopic dynamics in real-time is challenging, which often complicates a direct comparison between theory and experiment. Here, near-infrared laser-induced plasma dynamics in $\ensuremath{\sim}600\text{ }\text{ }\mathrm{nm}$ diameter helium droplets are studied by femtosecond time-resolved x-ray coherent diffractive imaging. An anisotropic, $\ensuremath{\sim}20\text{ }\text{ }\mathrm{nm}$ wide surface region, defined as the range where the density lies between 10% and 90% of the core value, is established within $\ensuremath{\sim}100\text{ }\text{ }\mathrm{fs}$, in qualitative agreement with theoretical predictions. At longer timescales, however, the width of this region remains largely constant while the radius of the dense plasma core shrinks at average rates of $\ensuremath{\approx}71\text{ }\text{ }\mathrm{nm}/\mathrm{ps}$ along and $\ensuremath{\approx}33\text{ }\text{ }\mathrm{nm}/\mathrm{ps}$ perpendicular to the laser polarization. These dynamics are not captured by previous plasma expansion models. The observations are phenomenologically described within a numerical simulation; details of the underlying physics, however, remain to be explored.

Comparison of disruption mitigation from shattered pellet injection with massive gas injection on J-TEXT

The mitigation of disruption damage is essential to the safe operation of a large-scale tokamak. In order to achieve the safe operation of ITER, the shattered pellet injection (SPI) has been considered as a primary measure of disruption mitigation. A dedicated argon SPI system, focusing on disruption mitigation has been designed for the J-TEXT tokamak. In the J-TEXT SPI system, a pure argon pellet can be formed in the freezing tube, then separated from the tube and accelerated by a punch mechanism. The pellet can be injected with a speed of 150–300 m s−1. The performance of disruption mitigation by Ar SPI has been compared with Ar massive gas injection (MGI). The cooling process observed from the ECE indicates that the SPI has deeper deposition, with the cold front that can reach the q = 1 rational surface in case of SPI, but stops at the q = 2 profile in MGI. The increase of core plasma density during a fast shutdown is higher than that with Ar MGI, which proves a deeper penetration of SPI. In disruption, the magnetohydrodynamic (MHD) activities, measured at the Mirnov coils, have similar behavior in both MGI and SPI shots. The m/n = 2/1 mode dominates the MHD activities from the penetration to the end of the thermal quench (TQ). Subsequently, in the current quench (CQ) phase, the m/n = 3/1 mode grows to become the dominant mode. The radiation power is much stronger in the SPI shot. The radiation asymmetry is compared in the two plasma disruption mitigation methods. Changing the pellet velocity can effectively adjust the TQ process and CQ rate, which can achieve a higher impurity assimilation rate when the pellet velocity increases in a certain range.

Assessment of the burning-plasma operational space in ITER by using a control-oriented core-SOL-divertor model

In future tokamaks, the control of burning plasmas will require careful regulation of the plasma density and temperature. Along with the design of effective burn-control systems, understanding how the fusion power varies in the density-temperature space is vital for the operation of fusion power plants. In this work, the steady-state operational space of ITER is studied using a control-oriented core-plasma model coupled to a two-point model of the scrape-off-layer (SOL) and divertor regions. The two models are coupled through the exchange of input-output parameters. The deuterium and tritium recycling from the wall are output parameters of the SOL-divertor model that are used as input parameters in the core-plasma density balance. Furthermore, the separatrix temperature, which is an output parameter of the SOL-divertor model, is incorporated into the radial core-plasma temperature profiles. Therefore, the temperature-dependent power balance of the plasma core is intimately linked to the SOL-divertor model. Both the power entering the SOL from the core, as determined by the core-plasma power balance, and the separatrix density, as dictated by the core-plasma density balance, are input parameters to the SOL-divertor model. They are control knobs in the SOL-divertor model that can be regulated using the core-plasma actuators: auxiliary power and pellet injection. There are various operational limitations, such as the saturation of the aforementioned actuators, that will prevent ITER from accessing certain high-fusion plasma regimes. The achievable tritium concentration in the fueling lines and the maximum sustainable heat load on the divertor will impose further restrictions. By accounting for these limitations, the ITER operational space is computed based on the coupled core-SOL-divertor model and visualized using Plasma Operation Contour (POPCON) plots that map performance metrics, such as the fusion to auxiliary power ratio, over the density-temperature space. Comparisons are drawn between plasmas with different recycling, confinement, and SOL-divertor conditions.

Helicon wave heating and current drive in toroidal plasmas

Investigation of helicon wave heating and current drive (CD) in toroidal plasma has been carried out with the ray-tracing code GENRAY. The wave trajectories, profiles of power deposition and driven-currents are presented. It is shown that increasing the poloidal launch angle does not change the absorption rate of the wave, but moves the peak of the wave deposition toward the plasma core. Wave frequencies in the range of our interest (460 MHz ∼ 480 MHz) do not change the absorption of the wave. Under low parameter plasmas, we find that both the peak position and magnitude of the wave power deposition are dependent on the launched parallel refractive index n||. With increasing the refractive index, the driving efficiency first increases and then decreases. The parallel refractive index figure of merit for mid-plane launching in HL-2M configuration is 3.2. With high plasma parameters, both the position and magnitude of the driving current are weakly dependent on n||. The magnitude of the driven current is determined by both the absorption rate and the value of ξe, which is related to n||. It is also demonstrated that driven current increases with increase of the core plasma temperature and decrease of the core plasma density. The plasma ohmic current seems do not significantly affect the helicon CD efficiency, but moves the driven current toward the plasma core region and narrows the width of the driven current profile. In addition, the preliminary calculations including scrape-off-layer (SOL) suggest that the power absorption rate loss in SOL is about 5%–20%. These results provide a reference and theoretical basis for the design and construction of HL-2M helicon wave system.

H-mode confinement in the pellet-enforced high-density regime of the all-metal-wall tokamak ASDEX Upgrade

During recent years a pellet-based system for feedback controlled core density fuelling has been developed at the ASDEX Upgrade tokamak, making the device well suited for the investigation of the reactor relevant high-density regime in an appropriate configuration. Efforts have proven that an adequately applied pellet actuator enables safe and reversible access to core plasma densities, as intended in current reactor designs and even beyond. From the analysis of a wide range of different plasma scenarios it becomes apparent that the beneficial increase of confinement with density saturates when approaching the Greenwald density. When taking this into account properly, it has emerged that for standard configurations the energy confinement can be sustained when the density is increased. However, the surplus confinement by different means, such impurity seeding or strong shaping, cannot be fully preserved. This is presumed to be attributed to a separatrix density which is not yet adequately controlled.

Impacts of sawtooth crashes on tokamak plasmas in DEMOs

Self-consistent simulations of sawtooth oscillations on four DEMO (demonstration fusion power plant) designs, proposed by European, Indian, Chinese, and Korean teams, are carried out using the Porcelli sawtooth triggering model and the modified version of the Kadomtsev magnetic reconnection model coupled with the BALDUR integrated code. The simulation results suggest that all sawtooth crashes are triggered by the driving force for the internal kink instability overcoming the fast ion stabilization found in all DEMOs. Conversely, different expansions of the pre-crash helical flux are found, causing different mixing radii on each design. The mixing radius of the European DEMO is found to be the largest, resulting in the largest reduction of the core plasma temperature after the sawtooth crash. It is also observed that for all DEMOs except the Chinese DEMO, regular sawtooth crashes lead to an increase in the core plasma density and radiative power density in different levels, which are due to the steepness of the pre-crash hollow density profile. Furthermore, while helium is found to be outwardly transported by sawtooth crashes for all DEMOs, the sawteeth yield a penetration of the seeded impurities, remarkably found in the European design. Interestingly, based on the European design, a decrease in the atomic mass of the plasma fuels and seeded impurities are found to yield greater flatness of the pre-crash density profile and thus a smaller impact of sawtooth oscillations on central density. Additionally, the sawteeth degrade the plasma performance in the Chinese DEMO, but demonstrate positive impacts on the performance in the European, Indian, and Korean DEMOs, which results from the increase of the central ion density caused by the sawteeth. Furthermore, variation in neutral beam injection power exhibits both positive and negative correlations of sawtooth oscillation impacts on central ion density. The sensitivity of the mixing region of DEMOs is also investigated for scans where the main plasma parameters are varied.

Assessment of particle and heat loads to the upper open divertor in ASDEX Upgrade in favourable and unfavourable toroidal magnetic field directions

Abstract Pairs of ASDEX Upgrade L-mode discharges with the toroidal magnetic field, BT, in the forward and reverse directions have been used to study the impact of neoclassical drifts on the divertor plasma conditions and detachment. The evolution of the peak heat flux and the total power loads onto both the outer and the inner targets depends significantly on the toroidal field direction: increasing the core plasma density affects mainly the heat loads in the BT 0 (favourable). Ion saturation current measurements show similar trends to those of the IR heat flux data. These discrepancies are not only caused by drifts but also by different levels of radiated power in the core, thus the power across the separatrix, Psep. Tomographic reconstructions show that Psep is not constant within the entire dataset. Finally, at I p = 0.8 MA , a significant reduction of the peak heat flux is observed at both targets for both field directions. On the other hand, at I p = 0.6 MA , a reduction of the peak heat flux is only observed for BT I p = 0.8 MA .

Efficient Fast Heating of Dense Core Plasma by Laser-Driven Strong Magnetic Field

Efficient Fast Heating of Dense Core Plasma by Laser-Driven Strong Magnetic Field

Statistical properties of the plasma fluctuations and turbulent cross-field fluxes in the outboard mid-plane scrape-off layer of Alcator C-Mod

The Alcator C-Mod mirror Langmuir probe diagnostic is used to characterize plasma fluctuations in the outboard mid-plane scrape-off layer (SOL). Ohmically heated L-mode plasmas with Greenwald fractions from 0.1 to 0.45 are analyzed. In plasmas with a low line-averaged core density, the radial profiles of the average electron density and temperature feature a two-scale structure with a strong gradient in the near-SOL and a weak gradient in the far-SOL. The relative fluctuation levels of the electron density are below 25% throughout the entire SOL and the fluctuation driven heat flux is dominated by plasma density fluctuations. In the plasma with the highest line-averaged core density, the radial profiles of the electron density and temperature feature only a single weak gradient length scale across the SOL. The plasma density at the limiter radius is approximately 10 times larger than in low-density discharges. The average electron temperature at the separatrix falls to approximately 30 eV while the magnitude of the relative temperature fluctuations increases with the line-averaged core plasma density. With increasing plasma density, the fluctuation-driven heat fluxes increase by more than one order of magnitude while retaining a relative fluctuation level of approximately 3. Even though the SOL is on average cooler than in low density discharges, large electron temperature fluctuations govern the turbulence driven heat flux towards the wall while contributions from density fluctuations and triple correlations contribute approximately 35% and 15%, respectively.

Imaging the Global Distribution of Plasmaspheric Oxygen

AbstractThis paper investigates the potential for 83.4 nm imaging of the plasmaspheric dense oxygen torus, using simple models for core plasma density and composition to constrain a simulated image code. We derive the requirements for plasmaspheric O+imaging, and the expected performance of an imager based on a slightly modified version of the IMAGE extreme ultraviolet camera. We find that such an imager can achieve a sensitivity of 0.69(s R pixel)−1, sufficient to capture the dense torus 83.4 nm signal with 25 min integration time. The background rejection ratios for this design are 1.5 × 10−4at 58.4 nm and 7.4 × 10−8for Lyman‐α. We discuss the effects of ion temperature and motion, and O++glow. We compute simulated O+images of the formation and global distribution of the dense torus. We also examine the possibility of direct observation of oxygen outflow from the ionosphere.

Numerical study of core formation of asymmetrically driven cone-guided targets

Compression of a directly driven fast ignition cone-sphere target with a finite number of laser beams is numerically studied using a three-dimensional hydrodynamics code IMPACT-3D. The formation of a dense plasma core is simulated for 12-, 9-, 6-, and 4-beam configurations of the GEKKO XII laser. The complex 3D shapes of the cores are analyzed by elucidating synthetic 2D x-ray radiographic images in two orthogonal directions. The simulated x-ray images show significant differences in the core shape between the two viewing directions and rotation of the stagnating core axis in the top view for the axisymmetric 9- and 6-beam configurations.

Compatibility of separatrix density scaling for divertor detachment with H-mode pedestal operation in DIII-D

The midplane separatrix density is characterized in response to variations in upstream parallel heat flux density and central density through deuterium gas injection. The midplane density is determined from a high spatial resolution Thomson scattering diagnostic at the midplane with power balance analysis to determine the separatrix location. The heat flux density is varied by scans of three parameters, auxiliary heating, toroidal field with fixed plasma current, and plasma current with fixed safety factor, q95. The separatrix density just before divertor detachment onset is found to scale consistent with the two-point model when radiative dissipation is taken into account. The ratio of separatrix to pedestal density, ne,sep/ne,ped varies from ⩽30% to ⩾60% over the dataset, helping to resolve the conflicting scaling of core plasma density limit and divertor detachment onset. The scaling of the separatrix density at detachment onset is combined with H-mode power threshold scaling to obtain a scaling ratio of minimum ne,sep/ne,ped expected in future devices.