Toroidal Direction Research Articles

Startup experiment with the newly installed lower tungsten divertor of KSTAR

Plasma startup experiments after the KSTAR upgrade of a lower divertor from carbon to tungsten have been studied during the KSTAR 2023 plasma campaign. In-vessel eddy currents, especially those induced on the divertor supporting structure in the toroidal direction, are significantly altered by the upgrade of the lower divertor configuration. These changes could affect the poloidal field configuration of the plasma startup phase, which is essential for the reliable breakdown and burn-through of deuterium neutral gas and the stable rise of the plasma current. Here, we present the development process of a new startup scenario that considers these upgrades. The vacuum magnetic field design code was utilized to estimate the changes in the poloidal field due to the new divertor configuration. Modifications to the startup scenario were prepared to compensate for the different effects of eddy currents, and experimental validation and optimization were performed during the initial phase of the KSTAR 2023 campaign. The final version of the startup scenario and plasma startup data are presented, and the reliability of the new startup scenario was also confirmed throughout the remainder of the KSTAR campaign.

Post-mortem analysis of the deposit layers on the lower divertor after the 2023 high particle fluence campaign of WEST

We analysed deposits collected on the lower divertor of WEST after the first high fluence campaign performed in 2023. Deposits were collected on the high field side (thick deposition area) of 2 ITER-grade plasma facing units (PFUs), located on different toroidal positions. Using focused ion beam cross sectioning, the deposits were found to be very thick (12–55 µm). A significant difference was observed in the toroidal direction with thicker deposits for the PFU located at the maximal heat load in the inner side, showing a deposition pattern due to the toroidal magnetic field modulation. Deposits present a complex layer-by-layer structure with dense layers, some melted parts and porosities within the layers. The deposition is mainly composed of tungsten with oxygen, boron, carbon and traces of nitrogen whose composition vary along the radial direction. We identified W-rich deposits in the high plasma flux area near the inner strike point and deposits rich in O, B, C and N in the low plasma flux area further away from this strike point. W dense layers of about 5 up to 40 µm thick in the thick deposit area near the strike point were attributed to the high fluence campaign. Pure boron layers resulting from wall conditioning and processed by plasma wall interactions were also observed.

Toroidal distribution of heat load on castellated plasma facing components for lower divertor target in EAST

In tokamak devices, the performance of plasma facing components (PFCs) under high heat loads is a key challenge for achieving long-pulse and high-performance plasma operations. The lower divertor of EAST adopts an ITER-like cassette modules (CMs) structure. However, regular distribution of damage along the toroidal direction have been observed on the divertor target plates, which not only affects the lifespan of PFCs but also limits the long-pulse and high-performance plasma operations. Therefore, investigating the toroidal distribution of the heat load on the divertor target in EAST is crucial for understanding the mechanism of damage on the target plate surface. The toroidal characteristics of heat loads on CMs of the lower divertor in EAST was analyzed using a three-dimensional magnetic field line tracing program (PFCFlux). For the inner vertical target (IVT) and outer vertical target (OVT) within a single CM, the heat load varies monotonically along the toroidal direction. In contrast, the heat load distribution on the outer horizontal target (OHT) is uniform, except at the chamfers, which is directly correlated with the target’s geometric structure and the magnetic field configuration. The surface heat load distribution exhibits toroidal non-uniformity, with the heat deposition pattern being generally consistent and displaying periodic variations. The toroidally uneven heat load distribution is attributed to the change in the magnetic field’s tilt angle. Additionally, the peak heat load is observed at the chamfer positions between cassettes, and an analysis is conducted on the relationship between peak heat load and the size of the chamfer. Smaller tilt angles result in lower surface heat load. Furthermore, a comparative analysis of peak heat loads at the chamfer under different plasma current conditions reveals that higher plasma currents generally result in increased peak heat loads. This investigation into the toroidal heat load distribution on the divertor surface contributes to understanding of the damage mechanism of W PFC as the target plate and provides valuable data for divertor design of future fusion devices.

Full-wave simulations on helicon and parasitic excitation of slow waves near the edge plasma

Helicon waves are thought to be promising in various tokamaks, such as DIII-D, because they can penetrate reactor-grade high-density cores and drive the off-axis current with higher efficiency. In the frequency regime ∼ 476 MHz, both slow electrostatic and fast electromagnetic helicon waves can coexist in DIII-D. If the antenna parasitically excites the slow mode, these waves can propagate along the magnetic field line into the scrape-off layer (SOL). Although the importance of the misalignment of the Faraday screen and the electron density in the SOL on the excitation and propagation of slow modes is well known, the conditions for minimizing slow mode excitation have yet to be optimized. Using the Petra-M simulation code in the 2D domain, we analyze the effects of the misalignment of the antenna in the poloidal direction, the misalignment of the Faraday screen in the toroidal direction, and the density in front of the antenna on slow mode generation. Our results suggest that the misalignment of the Faraday screen is a critical factor in reducing the slow mode and that the misalignment angle should be below ∼ 5° to minimize the slow wave excitation. When the electron density is higher than 3.5×1018 m−3 in the SOL, the generation of the slow mode from the antenna is minimized and unaffected by the misalignment of the Faraday screen.

X-ray imaging crystal spectrometer (XICS) diagnostic on the HL-3 tokamak

The construction of an X-ray imaging crystal spectrometer (XICS) diagnostic system on the HL-3 tokamak plays a crucial role in measuring core plasma parameter profiles, including ion temperature, electron temperature, rotational velocity, and impurity radiation profiles. This diagnostic system has been specifically designed to provide detailed and accurate data essential for understanding the behavior and characteristics of tokamak plasma. The resonance spectral line (w line of Ar XVII at 3.9494 Å) and its satellites of helium-like argon ions were chosen on the basis of the HL-3 parameter range. A spherically bent quartz crystal (1012) with a lattice constant of 2d = 4.562 Å, a curvature radius of Rc = 3.0 m, and a size of 10 cm (high) × 5 cm (wide) was mounted on an adjustable displacement platform in the XICS system. The position was adjusted in three dimensions (vertical, inclined, rotation) to effectively record the spectra of the helium-like argon ions. The XICS has a tangential angle of 49° for the toroidal direction in the magnetic axis. This layout accounts for 65.6 % of the toroidal rotation velocity component. The XICS system was set at an 11° poloidal angle in the mid-plane to cover a plasma range of 10 cm above to 50 cm below the mid-plane, which corresponds to the q = 1 surface of the HL-3 plasma (elongation κ > 1.8). This XICS layout produced a spectrum with poloidal and toroidal rotational contributions, requiring decoupling through data processing. In general, because the contribution of the poloidal rotation velocity is much smaller than that of the toroidal rotation velocity, it can be ignored. The XICS system offers a spatial resolution of ∼1.5 cm and a temporal resolution of 5–10 ms on the basis of a high-performance PILATUS3 × 900 K detector. The spectral resolution is λ/Δλ∼4×104, which satisfies the experimental measurement requirements. Preliminary results of ion and electron temperature profiles were obtained for the HL-3 tokamak.

Assessment of the runaway electron load distribution in ITER during 3D MHD induced beam termination

In ITER, disruption-born runaway electrons (REs), unless mitigated, are expected to form a several Mega-Ampere beam that ultimately intercepts the first wall leading to melting of the plasma facing components. Developing a successful mitigation strategy therefore requires modeling that takes into account the coupling between REs and the MHD during the formation and termination of the beam, along with estimates for the wall loads that can be compared to design values. Using the JOREK code, this work aims to provide the latter by presenting a novel model for collisions between REs and the wall and applying it to a beam termination scenario in ITER. To this end, the transport of REs is modeled by tracing particles in the fields calculated by preceding simulations of the disruption event where REs were treated in the fluid picture and self-consistently coupled to the MHD. The resulting heat loads are found to be highly localized in both the poloidal and toroidal directions, with 3D features also playing an important role in the appearance of hot spots. Peak loads were on the order of 107−108Jm2 . The load distribution was found to only be weakly sensitive to the initial phase space distribution of the REs when decoupled from the fields, while modifying properties of the underlying MHD yielded notable differences in wetted area and peak loads, in particular for cases with higher resistivity.

X-ray tomography system concept for stellarator TJ-II

A method based on the principle of x-ray tomography is a good way of assessing the parameters of the radiation emission distribution in the plasma. In this method, a set of radiation detectors is usually placed around a selected section of the plasma or in two groups of detectors arranged perpendicularly to each other in the top and side positions. Such an arrangement is intended to provide a sufficient amount of data to reconstruct the distribution of radiation emission intensities across the plasma cross section. For the plasma produced in the TJ-II stellarator, a two-dimensional (2D) imaging detector is proposed with the ability to discriminate photon energy in only one specifically selected position in the toroidal direction. The proposed diagnostics could come with advantages of a 2D toroidal camera, providing access to 3D information, and would complement standard poloidal tomography. It could be used either for better constraining of the 2D tomography in simple cases with axisymmetric SXR emissivity, for simulation validation of 3D effects in a synthetic diagnostic approach, or to attempt direct 3D tomography. The subject of this contribution is to present the idea of a 3D x-ray tomography system based on the use of a square matrix of GEM detector and the obtained preliminary results of the conducted simulations. The developed diagnostics is to contain 33 462 detection pixels associated with lines of sight. The analysis was carried out on the basis of the numerical simulations based on the GEM_MP.exe and Tomography.exe computer codes (both created at IPPLM) enabling the consideration of free-free, free-bound, and bound-bound x-ray radiation reaching each of the pixels of the GEM diagnostics (GEM_MP.exe code) and testing algorithms proposed for tomographic reconstruction of the plasma emissivity (Tomography.exe code). During calculations, real profiles of electron density and electron temperature, as well as real geometry of the measuring system were taken into account. Additionally, the influence of factors such as filters used in the detector and the efficiency of the used active medium on the quality of tomographic reconstruction was examined. The summary evaluates the proposed tomographic reconstruction method.

Reconstruction and interpretation of ionization asymmetry in magnetic confinement via synthetic diagnostics

Strong poloidal refueling asymmetry in the DIII-D tokamak is inferred from line radiation measurements. Synthetic diagnostics in neutral transport modeling coupled to gyrokinetic simulations illuminate implications for the plasma flow profile in the scrape-off layer of single-null beam-driven discharges. Recycling occurs primarily either on the inner or outer divertor legs, depending on the toroidal magnetic field direction. By reversing the toroidal magnetic field, the observed line radiation asymmetry is nearly eliminated or reversed. It is determined that, while relatively simple physics can describe the observed ionization asymmetry, predicting the overall brightness of the hydrogenic Lyman-α signal requires detailed simulation of the plasma and resulting turbulence. To this end, kinetic plasma simulations fully coupled to comprehensive neutral transport calculations—a novel capability—provide first-principles reproduction of Lyman-α observations on DIII-D.

Migration and balance of 13C in the stellarator Wendelstein 7-X after 13CH[formula omitted] local injection into the island divertor

A 13C tracer experiment was carried out to investigate impurity transport and migration in the stellarator Wendelstein 7-X (W7-X). About 4.0 × 1022 molecules of 13C marked methane were injected through the divertor gas injection system during plasma operation. After the experiment, the divertor target elements were extracted to be analyzed by post-mortem methods. In particular, the results of the Nuclear Reaction Analysis are presented, and the 13C balance and predominating contributors to migration are discussed.A two-dimensional 13C deposition pattern on 12C graphite modules was obtained with over 6000 single measurements. The post-mortem analysis results extrapolations show a high local deposition of around 25% of the 13C within a 5 cm radius around the injection holes embedded in a graphite target module of one of the ten divertor units. Additionally, a striation is detected in the toroidal direction along the outer strike line and the net carbon deposition zone along the intersecting magnetic island containing about 23% of the 13C within the same divertor unit where injections were made. The residual 52% of 13C are globally migrated across the machine. In total, ∼ 90% of the injected 13C could be extrapolated by post-mortem analysis. Thus, the 13C balance is closed. The distribution indicates a non-negligible amount of global transport and the strong influence of the magnetic topology on migration. In contrast to the strike line pathway, the transport along the magnetic island only takes place counterclockwise in toroidal direction. These results are used to provide benchmark data for ERO2.0 impurity tracing simulations and to be a reference for impurity migration processes in a stellarator.

Validation of SOLPS-ITER simulations against the TCV-X21 reference case

This paper presents a quantitative validation of Scrape-Off Layer Plasma Simulation-ITER (SOLPS-ITER) simulations against the TCV-X21 reference case and provides insights into the neutral dynamics and ionization source distribution in this scenario. TCV-X21 is a well-diagnosed diverted L-mode sheath-limited plasma scenario in both toroidal field directions, designed specifically for the validation of turbulence codes (Oliveira, Body et al 2022 Nucl. Fusion 62 096001). Five new, neutrals-related observables are added here to the extensive, publicly available TCV-X21 dataset. These are three deuterium Balmer lines in the divertor and neutral pressure measurements in the common and private flux regions. The quantitative agreement metric used in the validation is combined with the conjugate gradient method to approach the SOLPS-ITER input parameters that return the best overall agreement with the experiment. A proof-of-principle test of this method results in a modest improvement in the level-of-agreement; the shortcomings impacting the result and how to improve the methodology are discussed. Alternatively, a scan of the particle and heat diffusion coefficients shows an improvement of 10.4% in the level-of-agreement, approximately twice as high as that achieved by the gradient method. This result is found for an increased transport coefficient compared to what is usually used for TCV L-mode plasmas. The simulations further indicate that ∼65% of the total ionization occurs in the SOL, from which ∼70% in the divertor regions, despite being a sheath-limited regime, motivating the inclusion of self-consistent neutral models in future turbulence simulations on the path towards improved agreement with the experiment.

Development of a tokamak magnetohydrodynamic code with the discontinuous Galerkin and Weighted Essentially Non-Oscillatory methods

In this study, we present the development of a new initial-value magnetohydrodynamic (MHD) code for toroidal geometry using discontinuous Galerkin (DG) and Weighted Essentially Non-Oscillatory (WENO) methods. The code utilizes a triangular mesh based on the flux of the fixed boundary equilibrium in the poloidal plane, which is uniformly divided in the toroidal direction. By solving the conservative perturbed generalized Lagrange multiplier (GLM) MHD model, the code can simulate both ideal and resistive toroidal plasmas. We validate the code by performing the linear calculations of the internal kink mode and tearing mode, and our results show good agreement with calculations conducted using an eigen-value MHD code. Furthermore, nonlinear simulations of the resistive internal kink mode and tearing mode are also carried out. We discuss the challenges and limitations of the code, as well as the future efforts to improve its performance. The development of this new initial-value MHD code opens up new possibilities for studying and understanding the dynamics of toroidal plasmas, contributing to the advancement of fusion research.

Evaluation of Gamma-Ray Dose Distribution in the Toroidal Direction of LHD Vacuum Vessel from Radionuclides Generated in Deuterium Plasma Experiments

The evaluation of activities from radionuclides inside a vacuum vessel is required for the safe maintenance and estimation of the activation level of fusion devices. The Large Helical Device (LHD), Toki, Japan, has been performing deuterium plasma experiments since 2017, and radionuclides have accumulated in components such as vacuum vessels. In this work, the gamma-ray spectrum and activity of each detected radionuclide in an LHD vacuum vessel were evaluated using portable high-purity germanium (P-HPGe) and numerical calculations of photon transport for comparison with the removed armor tiles to evaluate the occupational exposure in the maintenance of LHD experiments. Measurements using the P-HPGe detector were performed at 20 positions along the toroidal direction of the LHD, at each O port, and between each pair of adjacent O ports. Using the estimated efficiency of the photoelectronic effects inside the P-HPGe detector calculated by Monte Carlo N-particle code version 6.1, the activities of the radionuclides in the armor tiles attached to the vacuum vessel were evaluated. The estimated activity concentrations were compared with the measured activities of the armor tiles removed from the vacuum vessel. Gamma rays from 58Co, 54Mn, and 60Co were detected, and the measured activities of 58Co and 54Mn generated by fast neutrons in the armor tile were almost consistent with those obtained from the removed armor tiles. The radioactivity of 60Co in the armor tile was underestimated compared with the activity obtained from the removed armor tiles. The measurement of activation by fast neutrons is feasible, whereas activation by thermal neutrons requires more accurate calculations and surveys to be measured using this method.

Analysis of the neutral fluxes in the divertor region of Wendelstein 7-X under attached and detached conditions using EMC3-EIRENE

This paper analyzes the neutral fluxes in the divertor region of the W7-X standard configuration for different input powers, both under attached and detached conditions. The performed analysis is conducted through EMC3-EIRENE simulations. They show the importance of the horizontal divertor to generate neutrals, and resolve the neutral plugging in the divertor region. Simulations of detached cases show a decrease in the number of generated neutrals compared to the attached simulations, in addition to a higher fraction of the ion flux arriving on the baffles during detachment. As the ionization takes place further inside the plasma during detachment, a larger percentage of the generated neutral particles leave the divertor as neutrals. The leakage in the poloidal and toroidal direction increases, just as the fraction of collected particles at the pumping gap. The fraction of pumped particles increases with a factor two, but stays below one percent. This demonstrates that detachment with the current target geometry, although it improves the power exhaust, is not yet leading to an increased particle exhaust.

The Hodge decomposition of shell current on the Keda Torus eXperiment device

The Hodge decomposition is a valuable tool for uniquely decomposing total currents on the composite shell into three types: inductive current, halo current, and harmonic current, each with its specific physical meaning. During plasma disruptions, halo currents appear, essential for studying the wall’s thermal load and electromagnetic force. Furthermore, understanding halo currents is crucial for improving the existing methodologies by removing their effects on equilibrium reconstructions and instability analyses based on boundary magnetic probe data. On the Keda Torus eXperiment (KTX) device, radial and tangent halo currents can be simultaneously provided to locate the contact region during a minor disruption experimentally. Additionally, experimental results demonstrate that, in addition to the occurrence of halo current during minor disruption events, halo current is already present simultaneously with the generation of inductive current when a resistive wall mode exists. For devices that lack the capability to measure the two-dimensional shell current distribution on the entire shell, we propose a method to estimate inductive and halo currents only using a set of shell currents along the toroidal direction. This technique is demonstrated on the KTX device and provides an overall good approximation of the inductive and halo current distribution.

Effects of toroidally-distributed-divertor biasing on scrape-off-layer (SOL) current drive, divertor particle flux and fast electron confinement in the QUEST spherical tokamak

A novel divertor biasing by four biasing plates distributed toroidally (TDDB) on the upper divertor target plate is applied to low density tokamak plasmas started-up by 28 GHz 2nd harmonic electron cyclotron current drive (ECCD) in a quasi-double null configuration of the QUEST spherical tokamak (ST). In the ST plasmas of line averaged electron density < ne > ∼0.7–1 × 1018 m−3, about 20%–40% of the current Ibias driven by ∼85 V sawtooth bias voltage reaches the lower divertor plate along the biased scrape-off-layer (SOL) flux tube as the SOL current ISOL . The fact ISOL is noticeably lower than Ibias indicates an appreciable leakage of parallel current from the biased SOL flux tube. The leakage currents in the toroidal and radial directions are confirmed by detection of them using the unbiased plates in the TDDB experiments. From the ion saturation current density profile obtained by a divertor Langmuir probe array, the fall-off lengths of divertor particle flux are estimated together with the strike line position. Total particle flux to the upper divertor, evaluated by the integrated ion saturation current density profile is reduced by up to 45% during positive biasing of the TDDB, depending on the position of the strike line to the biased plate. In addition, the TDDB also induces a noticeable loss of fast electrons produced by ECCD, leading to an ∼2% reduction in the maximum toroidal current of the ST plasma compared to a shot without the TDDB. Reduction of the divertor particle flux and enhancement of the fast electron losses are thought to be dominantly caused by E ×B drift induced by the TDDB. In the present experimental conditions, the effects of magnetic perturbations produced by the SOL currents on the fast electron losses can be neglected because of a too small SOL current.

SOLPS-ITER drift modeling of neon impurity seeded plasmas in EAST with favorable and unfavorable toroidal magnetic field direction

To better understand divertor detachment and asymmetry in the Experimental Advanced Superconducting Tokamak (EAST), drift modeling via the comprehensive edge plasma code SOLPS-ITER of neon impurity seeded plasmas in favorable/unfavorable toroidal magnetic field (B T) has been performed. Firstly, electrostatic potential/field (/E) distribution has been analyzed, to make sure that and E are correctly described and to better understand drift-driven processes. After that, drift effects on divertor detachment and asymmetry have been focused on. In accordance with the corresponding experimental observations, simulation results demonstrate that in favorable B T the onset of detachment is highly asymmetric between the inner and outer divertors; and reversing B T can significantly decrease the magnitude of in-out asymmetry in the onset of detachment, physics reasons for which have been explored. It is found that, apart from the well-known E × B drift particle flow from one divertor to the other through the private flux region, scrape-off layer (SOL) heat flow, which is much more asymmetrically distributed between the high field side and low field side for favorable B T than that for unfavorable B T, is also a critical parameter affecting divertor detachment and asymmetry. During detachment, upstream pressure (P u) reduction occurs and tends to be more dramatical in the colder side than that in the hotter side. The convective SOL heat flow, emerging due to in-out asymmetry in P u reduction, is found to be critical for understanding divertor detachment and asymmetry observed in EAST. To better understand the calculated drastic power radiation in the core and upstream SOL, drift effects on divertor leakage/retention of neon in EAST with both B T directions have been addressed for the first time, by analyzing profile of poloidal neon velocity and that of neon ionization source from atoms. This work can be a reference for future numeric simulations performed more closely related to experimental regimes.

Modeling study of divertor particle flow pattern and in–out plasma density asymmetry due to drifts with SOLPS and BOUT++

A study of the effects of drifts on the particle flow pattern and in–out divertor plasma density asymmetry for L-mode and H-mode plasmas is carried out for EAST discharges by the edge plasma transport codes SOLPS and BOUT++ . The simulation of L-mode plasmas is done by SOLPS while the simulation of H-mode plasmas is done by BOUT++ . The toroidal magnetic field direction for the simulated discharge is artificially reversed in the codes to study the effects of different drift directions on the divertor particle flow pattern and the in–out asymmetry of divertor plasma density. The divertor particle flows induced by diamagnetic and E × B drifts are found to have similar directions in the divertor region for the same discharge. The directions of the flows induced by drifts would be reversed with the reverse of toroidal magnetic field direction. The diamagnetic drift seems to have no effect on the in–out asymmetry of divertor plasma density due to its divergence-free nature. However, the E × B drift could result in a pronounced asymmetry of plasma density between the inner and outer divertor targets. The density in–out asymmetry caused by E × B drift is reversed with the reverse of E × B drift flow direction. Detailed analysis shows that the radial component of the E × B drift flow is the main cause of density asymmetry. The results from the simulation of H-mode plasmas with BOUT++ are similar to those of the L-mode plasmas with SOLPS except that the drift effects seem to be slightly larger in the H-mode plasmas compared to the L-mode plasmas.

Front face shaping of the inner wall tiles in the COMPASS Upgrade tokamak

The central column of the COMPASS Upgrade tokamak, where the plasma discharges will be started and shut down, is covered by a continuous first wall composed of 8 bulk tungsten guard limiters with recessed Inconel tiles in between. In diverted configurations, the plasma will be limited for a short duration (∼0.2 s–0.4 s) during the plasma current ramp-up and ramp-down phases but high heat fluxes due to short power decay lengths are foreseen. An adapted shaping is therefore needed to spread the heat loads over a maximum area. A novel method taking into account the different magnetic equilibria and the incident angle of the field lines was used to design a roof shape able to ensure a constant heat flux on the toroidal direction, accounting for the λqnear−SOL and λqfar−SOL. Due to the diversity of plasma scenarios considered on COMPASS-U, the optimized shape of the inner wall tiles strongly differs from one scenario to another, thus a compromise has to be found among all foreseen scenarios. The tile shaping also needs to take into account possible short wave misalignments. Calculations with the 3D magnetic field line tracing code PFCFlux validates the design, confirming the constant toroidal profile of deposited heat flux for the COMPASS-U workhorse scenarios, or slightly decreasing or increasing heat flux profiles (from the center to the sides of the tile) for scenarios with smaller or higher λqnear−SOL, respectively, within the expected large range (1.7 mm<λqnear−SOL<5 mm), but still providing acceptable loads. Estimated heat fluxes are in the range of 5 MW/m2 to 35 MW/m2 on the 8 guard limiters, receiving 70%–80% of the power deposited to the inner wall. In case of shortwave misalignments up to 0.5 mm in the radial direction, calculations show that leading edges are still protected, confirming the robustness of the design.

Modeling of plasma beta effects on the island divertor transport in the standard configuration of W7-X

The influence of plasma beta effects on the edge plasma transport in the Wendelstein 7-X standard configuration is studied systematically by using EMC3-EIRENE combined with a 3D equilibrium code named HINT. The magnetic topology changes induced by plasma beta effects are significantly reflected in plasma transport behaviors and heat flux patterns on divertor targets. After validating the modeling strategy by comparisons with experimental data, the extended simulations for high performance plasmas show that the threshold separatrix density for accessing the power detachment is reduced in higher beta plasmas. Compared with the vacuum field case, the impurity radiation distributions with finite beta effects are modified in the magnetic island region. The divertor heat flux is distributed more evenly along the toroidal direction on the strike line at the vertical target. The strike line on the horizontal target moves towards the pumping gap with an increase in the plasma beta. In addition, the different pressure profiles with the same central beta also result in a modified heat flux pattern on the divertor targets.

Analytic MHD equilibria for rotating plasmas with X-points and elongation, triangularity and squareness control

The MHD equilibrium equation that describes plasmas rotating in the toroidal direction is solved using the Solov’ev ansatz. The equilibria have 2 X-points and the shape of the plasma can be controlled by fixing the values of the aspect ratio, elongation, triangularity, squareness and curvature at the equatorial points. The plasma current, total β and Mach number can be also specified. Examples of equilibria with positive and negative triangularity and different values of the squareness and Mach number are presented.