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14.5 Comparison of major tokamaks in the world

The size of EAST is similar to that of DIII-D tokamak. The main parameters are summarized in Table. 1. A significant difference between EAST and DIII-D is that DIII-D has a larger minor radius, which makes DIII-D able to operate with a larger toroidal current than that EAST can do for the same current density. Another significant difference between EAST and DIII-D is that the coils of EAST are supper-conducting while the coils of DIII-D are not. The supper-conducting coils enable EAST to operate at longer pulse.

EAST DIII-D[19] KSTAR[17] SPARCWEST JET
Major radius R0 1.85m 1.67m 1.8m 1.85m 2.5m 2.96m
minor radius a 0.45m 0.67m 0.5m 0.57m 0.5m 0.9m
elongation
Plasma volume 20m3
No. of TF coils, turns, current 16, 130, 14.5kA24, 6, 126kA16, 56, 35.2kA
Bt at R0 3.26T 2.17T 3.5T 12.2T 3.7T 3.45T
CS coil module×turn×current 6×120×14.5kA
No. of independent PF coils 6+6 ?+18
Available solenoid magnetic flux 12Vs 10.5Vs 17Vs
Maximum plasma current 1.0MA 3.0MA 2MA 8.7MA 1MA 5MA
Pulse length 400s 10s 300s
superconducting? Yes No Yes Yes No

 

Table 1: Comparison of main parameters of EAST, DIII-D, JET, and ITER tokamaks. The Bt of EAST is calculated at R = 1.85m by using Eq. (489) with Is = 14.5kA (the maximal current allowed). The currents listed in the table are maximum plasma current whereas typical plasma currents for EAST are 0.5MA and for DIII-D are 2MA or less.

ITER[1] CFETR(old version)CFETR (new) BEST
Major radius R0 6.2m 5.7m 7.2m 3.6m
minor radius a 2.0m 1.6m 2.2m
elongation
No. of TF coils, turns, current 18, 134, 68kA 16, 132, 67.5kA 16, ?, ? 16, 152, 45.6kA
Bt at R0 5.29T 5.00T 6.5T
CS coil module×turn×current
No. of independent PF coils
Available solenoid magnetic flux
Maximum plasma current 15MA 10MA 14MA
Pulse length 400s
superconducting? Yes Yes Yes Yes

 

Table 2: Continued from Table 1.

 

 

ASDEX-U HL-2M NSTXMAST
Major radius R 0 1.65m 1.78m 0.85m 0.9m
minor radius a 0.7m 0.65m 0.68m 0.6m
elongation
No. of TF coils, turns, current 16,?,? 20,7,190kA
Bt at R0 2.99T 0.3T 0.55T
CS coil module×turn×current 1,?,?
No. of independent PF coils 1+16
Available solenoid magnetic flux 14Vs
typical plasma current 1.6MA 2.5MA
Pulse length 5s
superconducting? No No No

 

Table 3: Continued from Table 2.

DIII-D has 24 groups of TF coils with 6turns/coil, i.e., total turns are 24 × 6 = 144, with a maximum current of Is = 126kA in a single turn[19]. Using formula (488), the toroidal filed at R = 1.67m can be calculated, giving 2.17T.

DIII-D is special in that its poloidal field (PF) coils are located inside of the TF-coils, which makes the PF-coils more close to the plasma and thus more efficient in shaping the plasma. However, this nested structure is difficult to assemble. In superconducting tokamaks (e.g., EAST, KSTAR, ITER), PF coils are all placed outside of the TF-coils.

I noticed that HL-2M also has the PF coils located within the TF-coils, similar to DIII-D. This remind me that this layout may apply to all non-superconducting tokamaks (to be confirmed, No, non-superconducting tokamak ASDEX-U has PF coils outside of TF coils).

KSTAR has 16 TF coils and 14 PF coils. Both of the TF and PF coil system use internally cooled superconductors. The nominal current in TF coils is 35.2kA∕turn with 56turns∕coil and all coils connected in series[23]. Using these information and formula (488), the toroidal filed at R = 1.8m can be calculated, giving 3.5T. The PF coil system consists of 8 Central Solenoide (CS) coils and 6 outer PF coils and can provide 17 V-sec.

ITER has 18 TF coils with number of turns in one coil being 134 and current per turn 68kA[5]. Using these information and formula (488), the toroidal filed at R = 6.2m can be calculated, giving 5.29T

 

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