The Academy of Sciences of the Czech rebublic - ASCR


Magnetic Diagnostics

 

Measured quantities: plasma current density, total plasma current, plasma position, plasma shape, plasma conductivity, total energy content, MHD instabilities
Spatial resolution:

 

~< 2 mm for plasma position reconstruction
~< 5 mm MHD instabilities, shape and plasma current density distribution

 

Temporal resolution: < 10 µs
Responsible person: J. Havlíček, O. Hronová
Collaboration: Culham Centre for Fusion Energy, Abingdon, United Kingdom
Consorcio RFX, Padova, Italy

 

 

Diagnostic description:

 

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):
 
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).
 

 

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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

 

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Fig. 2: Example of EFIT reconstruction of Flux surfaces (tokamak COMPASS-D, UKAEA, Culham, shot 30866)