Poloidal Plasma Research Articles

Response of the poloidal rotation to resonant magnetic perturbations in the EAST tokamak

Abstract In this paper the response of the plasma poloidal rotation to resonant magnetic perturbations (RMPs) is investigated in EAST Tokamak using the multi-channel Doppler backscattering (DBS) system. It shows that the poloidal rotation spins up towards the ion-diamagnetic drift direction with increasing external perturbation field, which will reduce the edge shear. In ohmically heated discharges, the n = 1 RMP can only affect the edge poloidal rotation when the RMP coil current is small, and the influence will gradually reach the inner regions with increasing RMP coil current. At the moment of the n = 1 RMP penetration, all the poloidal rotations measured by the DBS will increase significantly, and then they will keep almost unchanged with the increase of the RMP coil current. In H-mode discharge, the poloidal rotation is significantly influenced by the n = 2 RMP, and the edge velocity well even reverses, along with edge-localized modes (ELMs) mitigation. However, in the same shot, the n = 4 RMP with the same coil current amplitude can hardly affect the poloidal rotations and the behavior of ELMs.

First results of turbulence investigation on Globus-M2 using radial correlation Doppler reflectometry

The first results of investigation of the turbulence structure using Doppler backscattering (DBS) on the Globus-M2 tokamak are presented. A one-channel DBS system with a variable probing frequency within the 18–26 GHz range was installed to investigate the edge plasma at normalized minor radii 0.9–1.1. Radial correlation Doppler reflectometry was used to study the changes in turbulence eddies after the LH transition. Correlation analysis was applied to the phase derivative of complex in-phase and quadrature (IQ) signals of the DBS diagnostic as it contains information about the poloidal plasma rotation velocity. In L-mode, the radial correlation length L r is estimated to be 3 cm and after transition to H-mode reduces to approximately 2 cm. Gyrokinetic modelling in a linear local approximation using code GENE indicates that the instability with positive growth rate at the normalized minor radius 0.75 in L-mode and H-mode on Globus-M2 was the ion temperature gradient (ITG) mode.

Retraction Note: Measurement of Plasma Poloidal Beta in Presence of Resonant Helical Field in IR-T1 Tokamak

Retraction Note: Measurement of Plasma Poloidal Beta in Presence of Resonant Helical Field in IR-T1 Tokamak

Retraction Note Investigation of Effects of Toroidal Field Ripple on Plasma Poloidal Beta in IR-T1 Tokamak

Retraction Note Investigation of Effects of Toroidal Field Ripple on Plasma Poloidal Beta in IR-T1 Tokamak

Roles of non-axisymmetric perturbations in free drift vertical displacement events on EAST

The safe operation of most tokamaks, especially the large ones, relies on the feedback control of vertical displacement events (VDEs). However, most of these feedback control systems are based on axisymmetric VDE models. In this study, we use NIMROD simulations to study the role of non-axisymmetric perturbations in free drift vertical displacement events on EAST. The high-n modes in the non-axisymmetric VDE grow first, which drives the formation of high-n magnetic island chains. Subsequently, the magnetic island chains grow and overlap with each other, leading to the destruction of the magnetic flux surface, which induces a minor disruption and accelerates the start of the following major disruption. The magnetic island and the stochastic magnetic field allow the toroidally non-axisymmetric poloidal plasma current to jet towards the hoop force direction, forming finger-like and filamentary structures. Such a plasma current non-axisymmetry strongly depends on the anisotropy in the thermal transport coefficients.

FLIPEC, an ideal MHD free-boundary axisymmetric equilibrium solver in the presence of macroscopic flows

The most relevant features of FLIPEC (Free fLow Iterative Plasma Equilibrium Code) are presented. This new code iteratively calculates free-boundary, axisymmetric ideal MHD equilibria with arbitrary poloidal and toroidal plasma flows. FLIPEC is a mature code that has emerged from a complete overhaul of a previous version (F-Torija Daza 2022 et al Nucl. Fusion 62 126044). It uses a (inverse) curvilinear coordinate representation for the Grad–Shafranov–Bernoulli equation system, which allows FLIPEC to extend its free-boundary capabilities to arbitrary plasma shapes and removes many limitations with regards to the distance between plasma and external coils. Run-time stabilization of vertical modes has also been implemented by means of artificial feedback coils. Finally, active targeting schemes have also been included. These capabilities are illustrated on two very different cases: the ITER tokamak baseline configuration and a NSTX spherical tokamak equilibrium.

Mechanical design of ITER radial neutron camera Ex-Port system

The Radial Neutron Camera is an ITER diagnostic designed to measure the un-collided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions, through an array of detectors covering a poloidal plasma section along collimated Lines Of Sight (LOS). It is composed by two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module of ITER Equatorial Port 1: the In-Port RNC, devoted to plasma edge coverage, and the Ex-Port RNC, devoted to the plasma core coverage. This paper presents an overview of the mechanical design of the Ex-Port RNC at the Preliminary Design Review (PDR) stage. The Ex-Port RNC is located in the Port Interspace and consists of a massive shielding structure hosting the detector units and two sets of collimators lying on different toroidal planes. The Ex-Port RNC design is presented both from the point of view of functional requirements (e.g. LOS positions and angles, radiation shielding, weight limitations) and of manufacturability. Finally, the Ex-port RNC structural integrity is assessed, and its design validated against the main loads and load combinations.

A Poloidal CO2 Laser Collective Scattering System for Measuring Electron-Scale Turbulence in the EAST

Anomalous electron thermal transport is an important issue that restricts the development of magnetic confinement thermonuclear fusion, and it is closely related to electron-scale turbulence. This paper introduces the poloidal CO2 laser collective scattering diagnostic system installed on the Experimental Advanced Superconducting Tokamak (EAST) for electron-scale turbulence measurement. The system can measure density fluctuations with four distinct wavenumbers simultaneously ranging from 10 to 30 cm−1 (correspondingly kθρs∼1to5) in two regions (the core region ρ≈0 to 0.4 and the outer region ρ≈0.4to0.8), which realizes the spatial resolution for turbulence measurement. And, the plasma poloidal rotation velocity in these two regions can be calculated using the measured density fluctuation frequency. In addition, the characteristics of small scattering angle and negligible wave refraction effects reduce the size of the ports required for this diagnostic system. These advantages make the diagnostic system an effective tool for measuring electron-scale turbulence and may play an important role in future burning plasma experiments.

Design & manufacturing of cooled immersion tubes for divertor bolometry in Wendelstein 7-X

After the successful Operation phase 1.2b experimental campaign Wendelstein 7-X was upgraded for steady-state operation with pulse duration of 1800s and up to 18 GJ energy input. Heat management in steady-state requires active water-cooling of the in-vessel components, in particular if temperature-stable environments are required for diagnostics such as bolometry. Two new, actively cooled bolometer immersion tubes (IT) with two camera housings (5–7 detector heads each) have been designed and manufactured for W7-X. By installing the two IT in the same module viewing the poloidal plasma cross-section from opposite sides provides a complete coverage with lines of sight. This allows tomographic reconstruction of the plasma radiation, but make accurate knowledge of the diagnostic geometry essential. This makes accurate manufacturing and assembly essential.An additional design challenge in W7-X is the suppression of Electron Cyclotron Resonance Heating (ECRH) stray-radiation as well as undesired particle deposition from plasma deposition, dust and boronization in the interior of the IT. This was realized with a welded sheet metal shell and a shutter system for each IT. In the design of the cooling channel, a high degree of temperature stability in particular for the detector heads was required. This resulted in complex components that were challenging for conventional manufacturing approaches. Selective laser melting might be an improvement for future projects.

Theoretical investigation of braking of tearing mode rotation by resistive walls in ITER

The locking of the 2/1 tearing mode to the resistive wall in the ITER tokamak (15 MA inductive scenario 2) is investigated theoretically using a cylindrical asymptotic matching model. The model takes into account the fact that ITER plasmas will effectively be surrounded by two walls; the inner blanket module layer with a time constant of about 23 ms, and the outer vacuum vessel with a time constant of about 380 ms. The model also takes cognizance of the fact that neither the blanket module layer nor the vacuum vessel can be accurately described as “thin” walls (in the ordinarily accepted sense). The model incorporates changes in both the plasma poloidal and the toroidal angular velocity profiles, in response to the electromagnetic braking torque that develops at the rational surface, because it turns out that neoclassical poloidal flow-damping is not strong enough to completely suppress changes in the poloidal velocity. Finally, the model accurately calculates changes in the poloidal and toroidal plasma angular velocity profiles by evolving the full angular equations of motion, taking the electromagnetic braking torque, plasma inertia, plasma viscosity, and poloidal flow-damping into account. The time required for the 2/1 tearing mode to grow from a small amplitude to a sufficient one to lock to the walls is found to be about 3.5 s. The critical full radial island width at which wall locking is triggered is found to be about 9% of the plasma minor radius.

Application of integrated simulation environment SIEMNED to the analysis of the MEPHIST-0 tokamak operation

This paper presents the results of numerical simulation of plasma equilibrium and stability in the MEPHIST-0 tokamak with SIEMNED software and comparison of simulation results with experiments. The determined characteristics of the vacuum chamber show that it significantly affects the entire discharge. For various scenarios of the inductor operation, a comparison of experimental data and simulated currents and magnetic fields induced in the chamber was carried out. For steady-state tokamak operation, a numerical study of equilibrium plasma configurations was carried out depending on the currents in the poloidal magnetic field coils and plasma current. The vertical plasma instability was investigated. The limiting values of plasma ellipticity preventing the vertical plasma instability were numerically determined. Numerical simulations show that plasma equilibrium is supported by induced currents. It was shown numerically that magnetic configuration with ‘zero of higher order’ were obtained before the plasma shot, suggesting consistency between the simulation results and observations.

Rational surfaces, flows and radial structure in the TJ-II stellarator

In this work, we report on the results obtained by measuring several turbulent quantities well inside the plasma edge by means of a Langmuir probe during dynamical rotational transform scans in the TJ-II stellarator, while applying a radial electric field to the edge plasma using a biasing probe. By calculating the intermittence parameter from floating potential measurements, we are able to identify a major low order rational surface and hence relate the probe measurements to the local value of the rotational transform. Based on the former, we are able to show that the poloidal plasma velocity (and hence radial electric field) has a significant radial structure that is clearly related to the rotational transform profile and in particular the lowest order rational surfaces in the range studied. The poloidal velocity is also affected by the edge biasing. The particle flux Γ was also found to exhibit a radial pattern, as did the flow shear suppression term , but the relation of the former to the low-order rational surfaces was less clear. We surmise that this lack of direct correspondence is due to an unknown term in the turbulence evolution equation: the instability growth rate, γ. We make use of a reduced Magnetohydrodynamic turbulence model to interpret the results. Overall, a picture is obtained in which the plasma self-organizes towards a state with a clear radial pattern of the radial electric field, in line with expectations from some numerical studies describing the spontaneous formation of an ‘E × B staircase’, consisting of alternating layers with fast and slow radial transport. In this state, the radial profiles of various quantities (density, temperature, pressure) will not be smooth.

Nuclear Analyses for the Assessment of the Loads on the ITER Radial Neutron Camera In-Port System and Evaluation of Its Measurement Performances

The radial neutron camera (RNC) is a key ITER diagnostic system designed to measure the uncollided 14- and 2.5-MeV neutrons from deuterium–tritium (DT) and deuterium–deuterium (DD) fusion reactions, through an array of detectors covering a full poloidal plasma section along collimated lines of sight (LoS). Its main objective is the assessment of the neutron emissivity/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\alpha $ </tex-math></inline-formula> source profile and the total neutron source strength, providing spatially resolved measurements of several parameters needed for fusion power estimation, plasma control, and plasma physics studies. The present RNC layout is composed of two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module (DSM) of ITER Equatorial Port 1 (EP01): the ex-port subsystem and the in-port one. The ex-port subsystem, devoted to the plasma core coverage, extends from the Port Interspace to the Bioshield Plug: it consists of a massive shielding unit hosting two sets of collimators lying on different toroidal planes, leading to a total of 16 interleaved LoS. The in-port system consists of a cassette, integrated inside the port plug DSM, containing two detectors per each of the six LoS looking at the plasma edges. The in-port system must guarantee the required measurement performances in critical operating conditions in terms of high radiation levels, given its proximity to the plasma neutron source. This article presents an updated neutronic analysis based on the latest design of the in-port system and port plug. It has been performed by means of the Monte Carlo MCNP code and provides nuclear loads on the in-port RNC during normal operating conditions (NOC) and inputs for the measurement performance analysis.

Fusion neutron emissivity tomography for ITER Vertical Neutron Camera

ITER Vertical Neutron Camera (VNC) is a diagnostic intended to measure time- and space-resolved neutron emission profile in the poloidal plasma section. It is composed of the two fan-shaped collimating structures: lower (5 collimators) and upper (6 collimators). Every collimator contains 2 fission chambers and 2 diamond detectors.We developed an algorithm of simultaneous reconstruction of DD and DT neutron emissivity profiles from measurements of VNC neutron flux monitors. The algorithm is based on maximum likelihood principle. We consider operation of diamond detectors as threshold counters. By setting deposited energy threshold equal to 3.5 MeV diamond detectors become DT neutron flux monitors. Relying on the different energetic sensitivities of VNC neutron flux monitors (fission chambers and diamond detectors) it is possible to distinguish fluxes of DD and DT neutrons.We evaluated sensitivity of neutron emissivity reconstruction to the statistical errors of measurements, errors of contribution matrix, and errors of magnetic surface shapes. The algorithm has been demonstrated to be robust against errors of the input data.

The Heavy-Ion Beam Diagnostic of the ISTTOK Tokamak-Highlights and Recent Developments.

The unique arrangement of the heavy-ion beam diagnostic in ISTTOK enables one to measure the evolution of temperature, density and pressure-like profiles in normal and AC discharges. The fast chopping beam technique provided the possibility to reduce the noise on the measurements of the plasma pressure-like profile and for the precise control of the plasma column position in real time. The consequent improvements in S/N levels allowed the observation of the effects of runaway beam magnetic energy conversion into plasma local heating. In addition, it made it possible to follow the evolution of the quiescent plasma maintained during AC transitions when the plasma current is null. The use of a new operation mode in the cylindrical energy analyzer provided an improved resolution up to five times in determining the fluctuations of the plasma potential as compared to the normal operation mode. Such analyzer is extremely compact (250 mm × 250 mm × 120 mm) and provides a unique geometry in order to cover the whole plasma diameter. The detector configuration choice gives the possibility for the simultaneous measurements of plasma poloidal magnetic field, plasma pressure-like and plasma potential profiles together with their fluctuations.

High-βp scenario realized by the integration of internal and external transport barriers in the HL-2A tokamak

High-βp scenario addresses needs for an attractive tokamak fusion reactor design, where βp is the normalized plasma poloidal beta. High-βp experiments have been performed during recent years in the HL-2A tokamak. The high-performance region is realized by the integration of internal and external transport barriers which are dubbed as ITBs and ETBs, namely, double transport barriers (DTBs), with the high-power NBI heating. Generally, the ITB forms and becomes strong after the NBI injection on HL-2A. Subsequently, the edge ion temperature and toroidal rotation increase, as a result that the ETB or pedestal creates and L-H transition occurs. A high-βp (βp∼3) scenario is obtained with Ip∼110 kA and q95∼5, and another high-βp (βp∼2.7) scenario is also realized with Ip∼160 kA and q95∼4.2. The two high-βp scenarios are both characterized by a large-radius ITB and an ETB. However, the two scenarios are different. For the former there are strong large-radius ITB and weak ETB on radial ion temperature profiles, but for the later there are weak large-radius ITB and strong ETB. Although two high-βp scenarios on HL-2A have been accessed, however they are both transient. In addition, MHD instabilities in high-βp scenarios are also present for a high fusion gain and reactor relevant high fusion-density plasma.

Prevention of electron cyclotron current drive triggering explosive bursts in reversed magnetic shear tokamak plasmas for disruption avoidance

An explosive burst excited by a neoclassical tearing mode (NTM) is one of the possible candidates for disruptive terminations in reversed magnetic shear (RMS) tokamak plasmas. For the purpose of disruption avoidance, numerical investigations have been implemented on the prevention of explosive bursts triggered by the ill-advised application of electron cyclotron current drive (ECCD) in RMS configuration. Under the situation of controlling NTMs by ECCD in RMS tokamak plasmas, a threshold in electron cyclotron driven current has been found. Below the threshold, not only are the NTM islands not effectively suppressed but a deleterious explosive burst could also be triggered, which might contribute to major disruption to tokamak plasmas. In order to prevent this ECCD from triggering explosive bursts, three control strategies have been attempted in this work and two of them have been recognized to be effective. One is to apply differential poloidal plasma rotation in the proximity of outer rational surface during the ECCD control process; the other is to apply two ECCDs to control NTM islands on both rational surfaces at the same time. In the former strategy, the threshold is diminished due to the modification of the classical tearing mode index. In the latter strategy, the prevention is accomplished as a consequence of the reduction of the coupling strength between the two rational surfaces via the stabilization of inner islands. Moreover, the physical mechanism behind the excitation of the explosive burst and the control processes by different control strategies have all been discussed in detail.

Upgrade of the APEC code: Implementation of halo currents and multi-turn coils analyses

The APEC code, implemented for the simulation of the plasma equilibrium and its evolution in presence of 3D passive structures [1][2], has been upgraded to simulate the effect of the eddy currents in the coils and the halo current in the plasma. The effect of the eddy currents in the coils is implemented considering each coil voltage driven and characterized by its number of turns with its corresponding resistivity. An additional degree of freedom in the finite element regions of the coils is in this case needed corresponding to the electrical current in each turn. The effect of the halo current is instead taken into account by applying both toroidal and poloidal plasma current also outside the last closed surface. The halo width is proportional to a fraction of the plasma core flux. The part of the plasma poloidal halo current intersecting the first wall is injected to the passive structures. Simulation with the upgraded version of APEC have been run for the ITER machine to show the soundness of the new APEC features.

Contribution to the angular momentum transport paradigm for accretion disks

We analyze the stationary configuration of a thin axisymmetric stellar accretion disk, neglecting non-linear terms in the plasma poloidal velocity components. We set up the Grad-Shafranov equation for the system, including the plasma differential rotation (according to the so-called co-rotation theorem). Then, we study the small scale backreaction of the disk to the central body magnetic field and we calculate the resulting radial infalling velocity. We show that the small scale radial oscillation of the perturbed magnetic surface is associated to the emergence of relevant toroidal current densities, able to balance the Ohm law even in the presence of quasi-ideal values of the plasma resistivity. The contribution to the infalling velocity of the averaged backreaction contrasts accretion, but it remains negligible as far as the induced magnetic field is small compared to that of the central body.

Controlling the rotation of drift tearing modes by biased electrode in ADITYA-U tokamak

The influence of background plasma poloidal rotation on the rotation frequency of the m/n = 2/1 drift tearing mode (DTM) has been studied in ADITYA-U tokamak. The poloidal rotation velocity of the background plasma in the ion diamagnetic direction is increased or decreased by inducing an outward or inward radial electric field, respectively, through a biased-electrode placed in the edge region of the plasma. The rotation frequency of the preexisting drift tearing mode, rotating in the electron diamagnetic direction, concomitantly decreased or increased with the application of bias depending on its polarity. The positive-bias increases the background plasma rotation in the ion-diamagnetic direction from its pre-bias value, hence decreasing the DTM rotation frequency, whereas the negative bias reduces the plasma rotation velocity in the ion-diamagnetic direction, hence increasing the mode rotation. In addition to that, a short gas puff introduced during the positive and negative bias pulse further reduces the mode frequency, however, with different amplitudes in different bias-polarities. These observations suggest that the background plasma rotation contributes significantly toward the rotation of DTMs, and the rotation frequency of the magnetohydrodynamic modes can be modified by varying the poloidal rotation of the background plasma and/or the diamagnetic drift frequency.