Institute of plasma physics › Structure of IPP › Fusion Plasma Division › COMPASS Tokamak › Diagnostics › Magnetic diagnostics
Magnetic Diagnostics
| Measured quantities: |
plasma current density, total plasma current, plasma position, plasma shape, plasma conductivity, |
| Spatial resolution: |
~< 2 mm for plasma position reconstruction ~< 5 mm MHD instabilities, shape and plasma current density distribution |
| Temporal resolution: |
< 10 µs |
| Responsible person: | |
| Collaboration: |
Culham Centre for Fusion Energy, Abingdon, United Kingdom |
The sources of magnetic field in tokamaks are of various kinds. Different type of currents generated by the power supplies, vessel currents generated by induced voltages and the plasma current give birth to magnetic fields in the poloidal and toroidal directions inside a tokamak. By measurement of these fields with magnetic diagnostics we are able to obtain plasma current density, total plasma current, plasma position, plasma shape, plasma conductivity, total energy content and to have an information on MHD instabilities. The COMPASS tokamak is equipped by more then 400 magnetic diagnostic coils covering the vacuum vessel in both the poloidal and toroidal directions (Fig. 1, Tab. 1) and enabling measurement of above mentioned quantities.
Another purpose of magnetic measurement is so called Equilibrium Reconstruction. The principle is based on the solution of the equation of equilibrium force balance (Grad-Shafranov equation):
(where Δ* is the elliptic operator, Ψ is the poloidal flux function, μ0 is the magnetic permeability, R is major radius, p is plasma pressure, F(Ψ) = RBTOR and JT is the toroidal current density) using the available measurements as constraints on the toroidal current density. The aim is to obtain a solution, which is consistent with the experimental magnetic signals, then a complete mapping of magnetic surfaces of the entire plasma can be done using numerical codes, for example code EFIT. EFIT solves the equilibrium problem, taking into account the internal measurements and the actual plasma current distribution. As a result, EFIT provides a complete magnetic description of the plasma, which includes both the plasma current density distribution and the flux distribution (Fig. 2).
Fig. 1: Top view of the magnetic diagnostics on tokamak COMPASS
|
Name of the coils |
External or internal location with respect to the vessel |
Number of coils |
Purpose of measurement |
|
Full toroidal loops (Flux loops) |
ext |
8 |
loop voltage and poloidal flux (used for real time cotrol and EFIT) |
|
Saddle loops |
ext |
22x4 and 2x8 |
the difference in poloidal flux (used for real time cotrol and EFIT) |
|
Remote loops |
ext |
5 |
loop voltage and poloidal flux |
|
Diamagnetic loops |
int |
2 |
perpendicular beta |
|
Diamagnetic compensation loops |
int |
2 |
toroidal field |
|
FCA coils |
ext |
16x3 |
horizontal, vertical and toroidal magnetic field |
|
Discrete Mirnov coils |
int |
3x24x3 |
local poloidal, radial and toroidal fields (hence 3 times), 24 at one cross section (used for halo currents study) |
|
High n coils |
int |
4 |
n-number of MHD instabilities |
|
Divertor Mirnov coils |
int |
2x8 |
coils imbedded in divertor plates (used for ELMs study) |
|
Internal Partial Rogowski coils |
int |
16 |
local magnetic field parallel to a vacuum vessel (used for poloidal current density distribution, real time cotrol and EFIT) |
|
External Partial Rogowski coils |
ext |
16 |
local magnetic field parallel to a vacuum vessel, eddy currents |
|
Full Rogowski coils |
1 ext and 1 int |
2 |
plasma and vacuum vessel current |
Table 1: The description of the magnetic diagnostics on COMPASS
Fig. 2: Example of EFIT reconstruction of Flux surfaces (tokamak COMPASS-D, UKAEA, Culham, shot 30866)
