Thermohydraulics for cooling of Tokamak Superconducting Magnets by pptfiles

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									      Thermohydraulics for cooling of Tokamak Superconducting Magnets
                                 S. Nicollet1, B. Lacroix, J.L. Duchateau


1
    CEA, DRFC/STEP, CE Cadarache, France


For the ITER (International Thermonuclear Experimental Reactor) project, very high magnetic fields
will be produced by different superconducting coils (Figure 1): the Toroidal Field (TF) coils are
necessary to confine the plasma (9 T); the Central Solenoid and the 6 equilibrium coils constituting
the PF (Poloidal Field) system provide the magnetic fields which induce, shape and control the 15
MA plasma current during the 1800 s of a typical plasma scenario.
In Fusion domain, the size of the machines requires high current conductors and high voltages during
protection phases when the magnet must be rapidly deenergised. In ITER, the choice of
superconducting coils has been made like in Tore Supra (Cadarache) to keep the electricity
consumption at low level The present solutions for ITER consist in Cable-In-Conduit Conductor
(CICC, Figure 2) characterised by a steel jacket, an external electrical insulation, and forced flow
cryogenic cooling of supercritical helium at few bars and a temperature at about 4.5 K. The CICC
comprises two regions in parallel: the bundle region where the superconducting strands are located,
and the central hole delimited by a so called central spiral.
The Toroidal Field type magnets include Nb3Sn CICC (68 kA) which are inserted into stainless steel
radial plates (Figure 3) and constitute the winding pack which is itself comprised in a stainless steel
case. The superconducting joints permits to wound the pancake’s CICC electrically in series whereas
they are hydraulically in parallel. During a safety discharge of the magnet, eddy currents and
associated heat generation are induced in these plates. This power is transferred into the conductor
helium channels by a diffusion process through the conductor steel jacket and insulation. The PF
Coils are wound with square Jacket niobium-titanium CICC (45 kA). These coils will experience
severe heat loads specially during the 400 s of the plasma burn: nuclear heating due to the 400 MW
of fusion power, thermal radiation and AC losses (due to variation of magnetic field).
In order to evaluate the conductor margins as well as to assess one possible quench, thermal and
hydraulic analysis are performed with two type of codes:
- the 1-D CRYOSOFT codes Gandalf and Flower, which model one CICC coupled with one
cryogenic loop
- the VINCENTA code developed by Efremov Institute which permits some quasi-3D model of a
system of coils coupled with cryogenic circuit.
The previously described heat loads are taken into account as well as the conductor parameters, the
magnetic field and the external cooling circuit. A series of parameters are determined such as friction
factor in CICC Channels, heat exchange coefficient from one region to the other, mass flow
distribution, etc…
In preparation of the ITER project, two model coils at reduced schedule have been produced with
Nb3Sn CICC and tested: the Central Solenoid Model Coil (CSMC) in JAERI (Japan) and the
European Toroidal Field Model Coil (TFMC) with 10 parallel pancakes (Figure 3) in FZK-Germany in
2001 and 2002. Numerical flow model and thermohydraulic analyses have been performed on
steady state or transient operation to determine the performances of the coils [REF1]. The results are
here presented and discussed.
Figure 1 : ITER Magnetic System Figure 2 : Cable In Conduit Conductor Figure 3 TFMC outer leg
equatorial cross section

								
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