Fusion Engineering and Design 58–59 (2001) 277–280

Design of the poloidal field system and plasma equilibriumof HT-7U tokamak

Wu Weiyue *, Li BaozengInstitute of Plasma Physics, Chinese Academy of Sciences, PO Box 1126, Hefei, Anhui 230031, People’s Republic of China

Abstract

HT-7U superconducting tokamak is an advanced steady-state plasma physics experimental device will be built inHefei at the Chinese Academy of Sciences, Institute of Plasma Physics (CASIPP). The HT-7U device is a fullsuperconducting Tokamak which means that the toroidal field (TF) coils system and poloidal field coils (PF) systemsare used with superconducting conductor. There are fourteen superconducting PF coils, including central selenoid,located symmetrically about the equator plane. The selenoid is supported by the case of TF coils. The NbTicable-in-conduit conduct (CICC) cooled by supercritical helium at 4.5 K is chosen as superconductor for all of thePF and TF coils. The plasma current is �1 MA and the duration of the plasma is �10 s for ohmic heatingdischarge. The preliminary engineering design of PF and the results of the calculating for plasma equilibrium will bereported in this paper. The maximum capacity of the magnetic flux for PF is �10 Web and the stray field in plasmainitiate region is quite low. The peak magnetic field in PF coils is �4.5 T. Three type shapes of plasma can be chosenduring the operating which are shown in the result of our calculations. © 2001 Elsevier Science B.V. All rightsreserved.

Keywords: Tokamak; Poloidal; Magnetic; Equilibrium

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1. Main parameters of HT-7U device

The scientific mission of HT-7U (Hefei Toka-mak-7 Upgrade), which is a next generation ad-vanced tokamak, is to study physical issues on thesteady-state operation of a non-burning plasma.The engineering mission of HT-7U tokamak is toestablish the technology basis of full supercon-ducting tokamaks to investigate the future fusion

device for demo reactor. HT-7U will have a longpulse (60–1000 s) capability, a flexible PF system,auxiliary heating and current drive systems, andwill be able to accommodate the divertor heatloads that make it an attractive test for the devel-opment of advanced tokamak operating modes.

In the global design of the magnets system ofHT-7U device [1], a PF system consists of four-teen coils and a TF system consists of 16 coils. PFand TF coils are located within an evacuatedcryostat that allows these magnets systems andtheir supporting mechanical structure to be cooledto liquid-helium temperatures, 4.5 or 3.8 K. At

* Corresponding author. Tel.: +86-551-559-1855; fax: +86-551-559-1310.

E-mail address: wuwy@mail.ipp.ac.cn (W. Weiyue).

0920-3796/01/$ - see front matter © 2001 Elsevier Science B.V. All rights reserved.

PII: S0920 -3796 (01 )00461 -6

W. Weiyue, L. Baozeng / Fusion Engineering and Design 58–59 (2001) 277–280278

Fig. 1. The section of the HT-7U.

2. Design on ohmic heating parameters of PFcoils system

For the HT-7U PF coils system there are threerequirements which are the capability of ohmicheating magnetic flux, the stray field in region ofplasma initial and the equilibrium field forplasma. The concept of the PF system design is toreduce the total ampere-turns and the magneticenergy stored in the system to be used for bothbuild-up plasma current and equilibrium for it.The first step in defining each PF coil location isto calculate the maximum radial position and sizeof the solenoid. The overall layout of HT-7U isshown in Fig. 1 and Table 1. In this table, Ioh isthe current in each coil at discharge moment ofohmic heating system, R1 and Z1 are the first turncenter position in R and Z directions, Nr and Nz

are the turns number for each coil in R and Zdirection, Bw and Bh are the distance between twoneighbour turns in R and Z direction.

In nominal cause of operating, the plasma cur-rent of HT-7U is 1 MA. The machine is capableof achieving this plasma current for at least 10 sby utilizing inductive drive alone. This is accom-plished with 10 Web capable of PF coil systemduring operating. The major radius of the plasmais 1.78 m and the minor radius is 0.4 m. A slightchange was made to accommodate better condi-tions for the mechanical design of divertor andvacuum vessel.

The ohmic heating system for HT-7U tokamakdevice was designed to satisfy the capability offlux swing and stray field, even the plasma initialcan be helped by lower hybrid wave or electroncyclotron wave. The ohmic heating system should

these temperatures, the magnet windings are capa-ble of producing intense magnetic fields, externalto the TF coils and associated structures. Fig. 1shows the section of the HT-7U superconductingtokamak.

The plasma geometry and the position aregiven from EQT (EQuilibrium of Tokamak) codeand PFFIFFPS code. The PF coils system shouldbe satisfied with the requirement of physics design[2] and engineering constraints, in which the PFcoils system should be satisfied with: (1) the ca-pacity of the magnetic flux is �10 Web; (2) thestray field of the ohmic heating system in theregion of plasma initial is lower than 5×10−3 T;(3) maximum field is limited lower than 5 T at thesuperconducting magnets; and (4) the current ofthe CICC conductor cannot be more the 15 kAon the field of magnets of 4.5 T [3].

Table 1Coil current and PF coils parameters

Z1 (cm)No. coil NrIoh (kA) Nz Bw (cm) Bh (cm)R1 (cm)

73.08556.311.4 2.27PF1 2.172012.7 56.3 52.385 7 20 2.17 2.27PF3

2.272.1720756.3 101.685PF5 11.697.2 172.02 11 4PF7 2.175.73 2.27

5.73PF9 97.2 182.1 17 12 2.17 2.27290.54 170.5PF11 63.15 10 2.17 2.27

2.272.178483.35.694 325.PF13

W. Weiyue, L. Baozeng / Fusion Engineering and Design 58–59 (2001) 277–280 279

Fig. 2. Ohmic heating stray field and flux contour.

Table 3Parameters of three type plasma shapes

Rx (m)dxkxa (m)R (m) Zx (m)

0.418 0.8041.92 0.516 1.5831.7981.95 1.00.545

0.448 1.79 0.62 0.8011.6321.91

netic flux capability and to satisfy the constraintsfor structural and superconductor performance.The peak field on PF coils is �4.5 Tesla in thefinal design [4].

3. Calculating results of plasma equilibrium

HT-7U will be operated in a circle section of theplasma, but a non-circle cross section with a highelongation is a preferred in the future of theoperating. The PF system should support theinductive operation under the plasma burn pulseslasting �10 s without EC wave or other condi-tions. Extension of the burn pulse will be towardsup to 1000 s as the steady-state. HT-7U operationbaseline will keep the plasma current of 1 MA withthe elongation kx is �1.8 to 2.0 and the triangu-larility dx is �0.6. The HT-7U plasma will beshaped and controlled by PF coils system only ina pulse length of 10 s. There are two candidatemodes of operation of the HT-7U plasma. In thebasic mode, the plasma minor radius is 0.4 m andthe major radius is 1.78 m. In other mode, theplasma minor radius can be 0.45 m and the majorradius can be 1.91 m, where the required currents

Table 2Ohmic heating stray field

Z (cm)R (cm) Bz (T)Br (T)

138.0 0 −0.001180170.0 0 0.00040

0.0005300178.0194.2 00 0.00045

−0.00291250.5 00

be provided with enough flux swing and very lowstray field in the region of the plasma breakdown.An �10 V/s and quite lower stray field wasdesigned in PF coils system. The ohmic heatingstray field is shown in Fig. 2 and Table 2. The PFcoils system design has been made as compact aspossible. To minimize ampere-turn of the PF coilsand power requirements and the overall deviceradius, the parameters of the ohmic heating systemhave been optimized to achieve maximum mag-

Fig. 3. Plasma shapes of HT-7U.

W. Weiyue, L. Baozeng / Fusion Engineering and Design 58–59 (2001) 277–280280

Fig. 4. Typical inductive discharge scenario (R=1.79 m, a=0.41 m, kx=1.84, dx=0.58).

Fig. 5. PF coils field in operation.for plasma control are moderate and the corre-sponding out-of-plane loads on the TF coils arestructurally tractable. The value of the poloidalbeta is �1.6 for two modes. The calculationresults are shown in Table 3 and the plasmashapes are shown in Fig. 3. In Table 3, Rx is thedistance from X-point to the center point of thetokamak device in horizontal and Zx is the dis-tance from X-point to the equator plane of thedevice.

4. The current waveforms of poloidal coils

Fig. 4 is the inductive discharge waveform.After a fixed OH current and plasma shape, thenthe current waveforms for PF coils are fixed. Thefield on coils is quite different in different timepoints from the current change during operating.The peak field on the PF coil occurs near the endsof the solenoid on the inboard sides and theallowable maximum field at the superconductingPF coil is 4.5 T. Fig. 5 shows the field on each PFcoil in the inductive discharge case. The field

maximum varying rate for PF coil is �7 T/s andtime duration is �60 ms. A well-controlled start-up was obtained with a loop voltage as low as 8V, which among tokamaks is quite a challengingtechnology in experiment.

Acknowledgements

The author gratefully acknowledges the supportof Wang Kuan Chen Education Foundation,Hong Kong.

References

[1] Wu Songtao et al., Fus. Technol. (1998) 1679–1682.[2] Yu Qingquan, Some Issues of HT-7U Physics Design, June

4, 1997.[3] Wu Songtao, The Modifications and Requirements of the

HT-7U Tokamak Device Design, Sept. 1999.[4] W. Weiyue, et al., Plasma Sci. Technol. 2 (2000) 119–126.