E. J. Strait

Magnetic diagnostic system of the DIII-D tokamak

External measurements of the magnetic field surrounding a hot, magnetically confined plasma yield important information about the state of the plasma, since the external field is generated in part by electric currents within the plasma itself. Therefore, magnetic diagnostics are an essential part of both the operation and the physics experiments in tokamaks and other magnetic confinement devices. The magnetic diagnostic system of the DIII-D tokamak includes approximately 250 inductive sensors of various types: axisymmetric poloidal flux loops, diamagnetic-flux loops, magnetic probes and saddle loops for the measurement of local magnetic field, and Rogowski loops for the measurement of coil currents and plasma current. The primary uses of the data include plasma shape and position control with a real-time digital control system, postdischarge equilibrium reconstruction, spectrum analysis in time and space of plasma instabilities, and direct feedback control of slowly growing instabilities. The sensors, instrumentation, calibration, applications, and operating experience are described.

MHD stability of optimized shear discharges in JET

The main limitation to the performance of JET optimized shear (OS) discharges is due to MHD instabilities, mostly in the form of a disruptive limit. The structure of the MHD mode observed as a precursor to the disruption, as measured from SXR and ECE diagnostics, shows a global ideal MHD mode. The measured mode structure is in good agreement with the calculated mode structure of the pressure driven kink mode. The disruptions occur at relatively low normalized beta (1 < βN < 2), in good agreement with calculated ideal MHD stability limits for the n = 1 pressure driven kink mode. These low limits are mainly due to the extreme peaking factor of the pressure profiles. Other MHD instabilities observed in JET OS discharges include, usually benign, chirping modes. These modes, which occur in bursts during which the frequency changes, have the same mode structure as the disruption precursor but are driven unstable by fast particles.

An upgrade of the magnetic diagnostic system of the DIII-D tokamak for non-axisymmetric measurements

The DIII-D tokamak magnetic diagnostic system [E. J. Strait, Rev. Sci. Instrum. 77, 023502 (2006)] has been upgraded to significantly expand the measurement of the plasma response to intrinsic and applied non-axisymmetric 3D fields. The placement and design of 101 additional sensors allow resolution of toroidal mode numbers 1 ≤ n ≤ 3, and poloidal wavelengths smaller than MARS-F, IPEC, and VMEC magnetohydrodynamic model predictions. Small 3D perturbations, relative to the equilibrium field (10(-5) < δB/B0 < 10(-4)), require sub-millimeter fabrication and installation tolerances. This high precision is achieved using electrical discharge machined components, and alignment techniques employing rotary laser levels and a coordinate measurement machine. A 16-bit data acquisition system is used in conjunction with analog signal-processing to recover non-axisymmetric perturbations. Co-located radial and poloidal field measurements allow up to 14.2 cm spatial resolution of poloidal structures (plasma poloidal circumference is ~500 cm). The function of the new system is verified by comparing the rotating tearing mode structure, measured by 14 BP fluctuation sensors, with that measured by the upgraded B(R) saddle loop sensors after the mode locks to the vessel wall. The result is a nearly identical 2/1 helical eigenstructure in both cases.

Experimental tests of linear and nonlinear three-dimensional equilibrium models in DIII-D

DIII-D experiments using new detailed magnetic diagnostics show that linear, ideal magnetohydrodynamics (MHD) theory quantitatively describes the magnetic structure (as measured externally) of three-dimensional (3D) equilibria resulting from applied fields with toroidal mode number n = 1, while a nonlinear solution to ideal MHD force balance, using the VMEC code, requires the inclusion of n ≥ 1 to achieve similar agreement. These tests are carried out near ITER baseline parameters, providing a validated basis on which to exploit 3D fields for plasma control development. Scans of the applied poloidal spectrum and edge safety factor confirm that low-pressure, n = 1 non-axisymmetric tokamak equilibria are determined by a single, dominant, stable eigenmode. However, at higher beta, near the ideal kink mode stability limit in the absence of a conducting wall, the qualitative features of the 3D structure are observed to vary in a way that is not captured by ideal MHD.

Frequency response of metal-clad inductive magnetic field probes

The frequency response of an inductive magnetic field probe enclosed in a metal shell is calculated, including the important effect of inductive coupling between the probe coil and the shell. With an appropriate passive external circuit, the bandwidth of the system can be extended significantly beyond the cutoff frequency of the shell alone. A simple, noninvasive test shows that the model accurately predicts the behavior of magnetic probes in the DIII‐D tokamak.

Higher fusion power gain with profile control in DIII-D tokamak plasmas

Strong shaping, favourable for stability and improved energy confinement, together with a significant expansion of the central region of improved confinement in negative central magnetic shear target plasmas, increased the maximum fusion power produced in DIII-D by a factor of 3. Using deuterium plasmas, the highest fusion power gain, the ratio of fusion power to input power, Q, was 0.0015, corresponding to an equivalent Q of 0.32 in a deuterium-tritium plasma, which is similar to values achieved in tokamaks of larger size and magnetic field. A simple transformation relating Q to the stability parameters is presented

Magnetic control of magnetohydrodynamic instabilities in tokamaks

Externally applied, non-axisymmetric magnetic fields form the basis of several relatively simple and direct methods to control magnetohydrodynamic (MHD) instabilities in a tokamak, and most present and planned tokamaks now include a set of non-axisymmetric control coils for application of fields with low toroidal mode numbers. Non-axisymmetric applied fields are routinely used to compensate small asymmetries ( δB/B∼10−3 to 10−4) of the nominally axisymmetric field, which otherwise can lead to instabilities through braking of plasma rotation and through direct stimulus of tearing modes or kink modes. This compensation may be feedback-controlled, based on the magnetic response of the plasma to the external fields. Non-axisymmetric fields are used for direct magnetic stabilization of the resistive wall mode—a kink instability with a growth rate slow enough that feedback control is practical. Saturated magnetic islands are also manipulated directly with non-axisymmetric fields, in order to unlock them from th...

Real time analysis of tokamak discharge parameters

The techniques used in implementing two applications of real time digital analysis of data from the DIII‐D tokamak are described. These tasks, which are demanding in both the speed of data acquisition and the speed of computation, execute on hardware capable of acquiring 40 million data samples per second and executing 80 million floating point operations per second. In the first case, a feedback control algorithm executing at a 10 kHz cycle frequency is used to specify the current in the poloidal field coils in order to control the discharge shape. In the second, fast Fourier transforms of Mirnov probe data are used to find the amplitude and frequency of each of eight toroidal mode numbers as a function of time during the discharge. Data sampled continuously at 500 kHz are used to produce results at 2 ms intervals.

Damping of toroidal Alfvén modes in DIII‐D

The previously developed single‐gap kinetic theory for toroidicity‐induced Alfven eigenmodes (TAE) is extended and applied to Doublet III‐D [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] experimental data. It is found that the theory gives reasonable agreement with the data when an appropriate magnetohydrodynamic boundary condition is accounted for. As is shown, this boundary condition is equivalent to an appropriate real frequency shift relative to the continuum gap. The correct eigenfunction near the gap, and thus the correct damping, is obtained by using the gap structure calculated from an equilibrium reconstruction that includes low aspect ratio, noncircularity, and finite beta effects, combined with an experimentally measured frequency. In the considered experimental cases, the damping is well into the nonideal regime.

Approach to steady state high performance in DD and DT plasmas with optimized shear in JET

Steady state high performance with improved core confinement and sustainable plasma edge conditions has been approached on JET in a double barrier (DB) mode. The DB mode combines an internal transport barrier of the optimized shear regime with an edge transport barrier of an ELMy H?mode regime. Improved confinement with an H?factor HITER-89 ? 2 has been maintained for four energy confinement times. Ion and electron temperature profiles remain peaked in the DB mode, while the density profile is broad and similar in shape to the conventional ELMy H?mode profile. The energy confinement improves across the whole plasma cross-section, and the ion heat conductivity falls to the neoclassical level in the core. Particle transport studies show that impurity accumulation can be avoided in the DB mode. In DT discharges the DB mode has attained a fusion gain of Q ? 0.4, producing 6.8?MW of fusion power, compared with Q ? 0.2 in the conventional sawtoothing ELMy H?mode. The ELMs are more benign, with an amplitude an order of magnitude smaller in the DB mode. The DB mode has a high potential to improve performance in reactor relevant conditions.