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SST-1 Tokamak
A steady state superconducting tokamak SST-1 is under design and fabrication at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation (k) and triangularity (d).
The specific objective of the SST-1 project is to produce 1000 s elongated double null divertor plasma. There are several conventional questions in tokamak physics, which will be addressed again in steady state scenario. Some of these are related to the energy, particle and impurity confinement, the effect of impurities and edge localized modes (ELM) in steady on energy confinement, the stability limits and their dependence on current drive methods, the resistive tearing activities in presence of RF fields, disruptions and vertical displacement events (VDE), and thermal instability. In steady state operations non-inductive current drive will sustain the plasma current. Different aspects of current drive such as different current drive methods and their combinations, current drive efficiency, profile control and bootstrap current, will be studied. An efficient divertor is required for the steady state operations with elongated plasma. Various aspects of divertor operation such as steady state heat and particle removal, erosion and particle recycling, radiative divertors and pumped divertors will be studied. Advance tokamak regimes are of prime interest in fusion research. These regimes are characterized with high bN and high bootstrap current, and are generally obtained in high (H-mode) and very high (VH-mode) confinement modes in plasma with high triangularity, elongation and large negative shear. Although SST-1 is not optimized for advanced tokamak regimes, we propose to attempt some experiments in this direction with in the limitations of the machine.
The choice of the parameters is dictated by the technological and physics goals. NbTi superconductor at 4.5K is used for the superconducting magnets and maximum field at the conductor is restricted to 5.1 T. Low aspect ratio machines are difficult to design using superconducting coils due to space restrictions. Furthermore, higher aspect ratios have the advantages such as high bootstrap current, better confinement etc. We have, therefore, opted for large aspect ratio (~5) in SST-1. Other Tokamaks have observed substantial improvement in confinement (VH mode) and bN with higher triangularity (d~0.4-0.8). Elongation improves the current carrying capacity of the plasma. With elongation in the range of k~1.6-2.0 improvement in bN has been observed. We have, therefore, chosen a range of k and d similar to these ranges. The double null configuration allows for the distribution of power between larger number of divertor plates thus reducing the heat load per plate. We have, therefore, selected double null configuration, with a provision to go to single null operations in future.
The machine has a major radius of 1.1 m, minor radius of 0.20 m, a toroidal field of 3.0 T at plasma center and a plasma current of 220 kA. Elongated plasma with elongation in the range of 1.7 to 1.9 and triangularity in the range of 0.4 to 0.7 can be produced. Hydrogen gas will be used and plasma discharge duration will be 1000 s. Auxiliary current drive will be based mainly on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Heating (ICRH) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating (ECRH) at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV.
Superconducting (SC) coils for both toroidal field (TF) and Poloidal Field (PF) are to be deployed in the SST-1 tokamak. An ultra high vacuum (UHV) compatible vacuum vessel, placed in the bore of the TF coils, houses the plasma facing components (PFC). A high vacuum cryostat encloses all the SC coils and the vacuum vessel. Liquid Nitrogen (LN2) cooled thermal shield between the vacuum vessel & SC coils as well as between cryostat and the SC coils reduce the radiation heat load on the SC coils. Normal conductor ohmic transformer system is provided to initiate the plasma and sustain the current for initial period. A pair of vertical field coils is provided for circular plasma equilibrium at the startup stage of the plasma. A set of saddle coils placed inside the vacuum vessel provide fast vertical control of the plasma while PF coils are to be used for shape control. Other subsystems include radio frequency (RF) systems for pre-ionization, auxiliary current drive and heating, neutral beam injection (NBI) system for supplementary heating, cryogenic systems at liquid helium (LHe) and LN2 temperatures, chilled water system for heat removal from various subsystems. A large number of diagnostics for plasma and machine monitoring will be deployed along with a distributed data acquisition and control system.
References:
[1] The SST Team, ²Conceptual Design of SST-1 Tokamak², 16th IEEE/NPSS Symposiumon Fusion Engineering, University of Illinois, Urbana-Champaign, 1, (1995) 481.
[2] Y. C. Saxena and SST-1 Team, "Present Status of SST-1", Nucl . Fus. 40 (2000) 1069
[3] Pradhan, S,., et al., ²SST-1 Poloidal Field Magnets², Proc. 17th IEEE/NPSS Symposium on Fusion Engineering, San Diego Vol. 2 (1997) 665.
[4] Bahl, R., et al., ²Design of Ohmic System for SST-1², Proc. 17th IEEE/NPSS Symposium on Fusion Engineering, San Diego Vol.2, (1997) 661.
[5] Pradhan, S., et al., "Superconducting Cable-in-Conduit-Conductor for SST-1 magnets", FURUKU Report 99-06(67), Vol. II,(1999) 482
[6] Bedakihale, V.M., et al., ²Support structure of TF magnet System², Proc.17th IEEE/NPSS Symposium on Fusion Engineering, San Diego Vol. 2, (1997) 657.
[7] Ranga Nath T., et al, "Design analysis of vacuum vessel and cryostat of SST-1 Tokamak", Proc. 17th IEEE/NPSS Symposium on Fusion Engineering, San Diego Vol. 2 (1997) 1025.
[8] D. Chenna Reddy et al., " Design of plasma facing components of SST-1", FURUKU Report 99-06(67), Vol. III, (1999), 572.
[9] M. Warrier et al, Jl. of Nucl. Mater. Vol. 266-269 (1999) 726.
[10] Chakraborty, A. K, et al., "Neutral Beam injector for Steady state super conducting tokamak" (Proc. Symp. on Fusion Technology, 1996), Lisbon.
[11] P. Ranjan et al, "Central control system for SST-1 Tokamak", FURUKU Report 99-06(67), Vol. III, (1999), 722.
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