transformer system will be provided to initiate the plasma and ..... nitrogen storage tanks of 105 m3 have been provided. ..... Steady state neutral beam injector.
Present status of the SST-1 project Y.C. Saxena, SST-1 Team Institute for Plasma Research, Bhat, Gandhinagar, India Abstract. SST-1 is a steady state superconducting tokamak used to study the physics of plasma processes in tokamaks under steady state conditions and to gain knowledge about technologies related to steady state tokamak operation. Major subsystems of SST-1 are described and the present status of the SST-1 project is presented.
1.
Introduction
The steady state superconducting (SC) tokamak SST-1 is under design and fabrication at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of plasma processes in tokamaks under steady state conditions and learning the technologies related to steady state tokamak operation. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 [1, 2] tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation κ and triangularity δ. In the following we give a brief description of the SST-1 tokamak and discuss the present status of the project.
2. 2.1.
The SST-1 machine Physics issues for the SST-1 tokamak
The specific objective of the SST-1 project is to produce 1000 s elongated double null divertor plasmas. There are several conventional questions in tokamak physics, which will be addressed again in the steady state scenario. Some of these are related to energy, particle and impurity confinement, the effect of impurities and ELMs in steady state on energy confinement, stability limits and their dependence on current drive methods, resistive tearing activities in the presence of RF fields, disruptions and vertical displacement events (VDEs), 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 steady state operations with elongated plasma. Various aspects of divertor operation, such as steady state heat and particle removal, erosion and particle Nuclear Fusion, Vol. 40, No. 6
recycling, radiative divertors and pumped divertors, will be studied. Advanced tokamak regimes are of prime interest in fusion research. These regimes are characterized by high βN 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 within the limitations of the machine. 2.2.
Machine parameters and features
The choice of the parameters is dictated by the technological and physics goals. As this is our first experience with SC coils, we have decided to use an NbTi superconductor at 4.5 K and hence have restricted the toroidal field to 3 T at the plasma centre. Low aspect ratio machines are difficult to design using SC coils due to space restrictions. Furthermore, higher aspect ratios have advantages such as high bootstrap current and better confinement. We have, therefore, opted for a large aspect ratio (∼5) in SST-1. At other tokamaks substantial improvements in confinement (VH mode) and βN with higher triangularity (δ ∼ 0.4–0.8) have been observed. Elongation improves the current carrying capacity of the plasma. With elongation in the range κ ∼ 1.6–2.0, improvement in βN has been observed. We have, therefore, chosen ranges of κ and δ similar to these ranges. The double null configuration allows for the distribution of power between a larger number of divertor plates, thus reducing the heat load per plate. We have, therefore, selected the double null configuration, with the provision to go to single null operations in future. The machine has a major radius of 1.1 m, a minor radius of 0.20 m, a toroidal field of 3.0 T at the plasma centre and a plasma current of 220 kA. Elongated plasma with elongation in the range
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2000, IAEA, Vienna
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Y.C. Saxena and SST-1 Team
Table 1. Typical operating points of SST-1 Phase I (L mode)
Phase II (H mode)
Basic plasma parameters
(Step 1) Circular ohmic
(Step 2) Circular LHCD
(Step 3) Elongated LHCD (0.5 MW)
(Step 4) Elongated LHCD (1 MW )
Low power
Low field
High power
BT (T) Ip (kA) q ∗ (Cylindrical q) H factor, τE /τITER89 -P hni (m−3 ) n ¯ 0 (m−3 ) Paux (MW) Zeff Elongation, κx Triangularity, δx Elongation, κ95 Triangularity, δ95 τE (ms) hTe i (keV) hTi i (keV) Te0 (keV) Ti0 (keV) β (%) βN (%, m T/MA) βp fbs BS collisionality correction
1.5 110 2.5 — 3 × 1019 4.5 × 1019 — 1.7 1 0 1 0 12 0.24 0.09 0.39 0.16 0.18 0.48 0.33 0.08 0.57
1.5 110 2.5 1 5 × 1018 7.5 × 1018 0.17 5 1 0 1 0 10 0.82 0.32 1.37 0.55 0.1 0.28 0.19 0.04 0.11
1.5 150 4.1 1 8 × 1018 1.2 × 1019 0.5 5 1.8 0.66 1.6 0.54 11 1 0.4 1.7 0.67 0.2 0.4 0.46 0.1 0.12
3 220 5.7 1 1 × 1019 1.5 × 1019 1 1.7 1.8 0.66 1.6 0.54 12 1.8 0.74 3.1 1.2 0.12 0.32 0.49 0.12 0.07
3 220 5.7 2 2 × 1019 3 × 1019 1 1.7 1.8 0.66 1.6 0.54 26 2 0.8 3.3 1.3 0.25 0.68 1.1 0.25 0.09
1.6 100 6.6 2 2 × 1019 3 × 1019 1 1.7 1.8 0.66 1.6 0.54 12 0.9 0.36 1.5 0.6 0.39 1.3 2.3 0.55 0.2
3 220 5.7 2 5 × 1019 7.5 × 1019 5 1.7 1.8 0.66 1.6 0.54 13 1.9 0.78 3.2 1.3 0.61 1.7 2.6 0.62 0.14
1.7–1.9 and triangularity in the range 0.4–0.7 can be produced. Hydrogen gas will be used and the plasma discharge duration will be 1000 s. Auxiliary current drive will be based mainly on 1.0 MW of LHCD at 3.7 GHz. Auxiliary heating systems include 1 MW of ICRH at 22–91 MHz, 0.2 MW of ECRH at 84 GHz and an NBI system with peak power of 0.8 MW (at 80 keV) with a variable beam energy in the range 10–80 keV. Superconducting coils for both toroidal field (TF) and poloidal field (PF) are to be deployed in the SST-1 tokamak. An ultrahigh vacuum (UHV) compatible vacuum vessel, placed in the bore of the TF coils, will house the plasma facing components (PFCs). A high vacuum cryostat will enclose all the SC coils and the vacuum vessel. Liquid nitrogen (LN2 ) cooled thermal shields between the vacuum vessel and the SC coils as well as between the cryostat and the SC coils will reduce the radiation heat load on the SC coils. A normal conductor ohmic transformer system will be provided to initiate the 1070
plasma and sustain the current for the initial period. A pair of vertical field coils will be provided for circular plasma equilibrium at the startup stage of the plasma. A set of saddle coils placed inside the vacuum vessel will provide fast vertical control of the plasma, while PF coils are to be used for shape control. Other subsystems include RF systems for pre-ionization, auxiliary current drive and heating, an NBI system for supplementary heating, cryogenic systems at liquid helium (LHe) and LN2 temperatures, and a chilled water system for heat removal from the various subsystems. A large number of diagnostics for plasma and machine monitoring will be deployed along with a distributed data acquisition and control system. 2.3.
Operational scenario
SST-1 is to be operated in two phases. Phase I will involve various steps, starting with ohmic circular plasma to full power operations with the divertor Nuclear Fusion, Vol. 40, No. 6 (2000)
Article: Present status of the SST-1 project configuration. Phase II will involve advanced tokamak operations. Table 1 shows typical operating parameters in various stages of phase I and phase II. The machine operations will commence with a circular, pulsed plasma driven by ohmic field with a pulse length of ∼1 s. Lower hybrid current drive will be attempted to sustain current in the limiter configuration. Up to 200 kW power will be used at this stage. Step 3 will involve divertor operation at 1.5 T magnetic field. The plasma will be initiated as circular plasma, current drive will then be taken over by LHCD and divertor coils will be brought in on a slow timescale (∼3–4 s) to produce elongated divertor plasma. In the next step of SST-1 operations, the toroidal field will be increased to 3.0 T and long pulse, elongated plasma will be produced. Full design parameters will be attempted and a total of 1 MW auxiliary power will be introduced. Advanced tokamak modes will be explored in phase II of the SST-1 operations. In the normal operating range (phase I), the values of β, βN and βp are low. In order to improve β values with the same auxiliary power of 1 MW, we may operate at low magnetic field. Further improvement of βN and βp can be achieved by operating at lower currents. Operations with VH mode, non-monotonic q profiles and significant bootstrap current will also be attempted after suitable modification of the divertor for enhanced power handling capacity and enhancement of the RF power systems.
3.
Magnet system
The magnet system comprises the TF coil system, the PF coil system, the ohmic transformer, the vertical field coils and the vertical position control coils. A cross-section of SST-1, indicating various magnets, is shown in Fig. 1. 3.1.
Toroidal field coil system
The TF coil system design requirements include the production of a 3.0 T magnetic field on the plasma axis with
