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In India, scientific research in tokamak plasmas has been continuing for more than a decade now. In tokamaks, the plasma is formed by an electrical breakdown in an ultra high vacuum toroidal vessel and a current is inductively driven in the plasma. As the plasma temperature rises the efficiency to heat the plasma drops. To further raise the temperature of the plasma to fusion grade, one has to use auxiliary heating schemes. During experimentation at high temperatures, it is also required to diagnose the plasma with various sophisticated diagnostic tools. Inherent drawback for future uses is the pulsed nature of tokamaks. One of the areas of research, therefore, has been steady state operation of tokamaks.
A steady state superconducting tokamak, SST-1, is in advanced stage of fabrication at the Institute for Plasma Research, Gandhinagar. The objectives of SST-1 include :
To study physics of plasma Processes in tokamak under steady-state conditions & contribute to the tokamak physics database for very long pulse operations.
Learning new technologies relevant to steady state tokamak operation.
Superconducting magnets and associated power supplies and protection system.
Large scale cryogenic system (Liquid helium and liquid nitrogen).
High Power Radio Frequency Systems.
Energetic Neutral Particle Beams.
High heat flux handling.
The machine has a major radius of 1.1 m, minor radius of 0.20 m, a toroidal field of 3.0 Tesla at plasma centre and a plasma current of 220 kA.
Controlled thermonuclear fusion is one of the attractive futuristic sources of energy. All over the world, the research in this field of energy has been continuing for the last fifty years.
Research efforts in this area are broadly divided into inertial confinement and magnetically confined plasmas. Among the magnetically confined systems, Tokamaks have been the most successful machines to achieve the technological goals .
In India, the scientific research in tokamak plasmas has been continuing for more than a decade now.
The tokamak Aditya developed at the Institute for Plasma Research, Gandhinagar, Gujarat, is one of the milestones of this endeavour. A steady state super-conducting tokamak, SST-1 is in advnced stage of fabrication at the Institute. Present here is the status of this venture.
Superconducting coils for both toroidal field and poloidal field are to be deployed in the SST-1 tokamak. NbTi superconductor at 4.5K is used for the superconducting magnets and maximum field at the conductor is restricted to 5.1 Tesla. An ultra high vacuum compatible vacuum vessel, placed in the bore of the toroidal field coils, houses the plasma facing components. A high vacuum cryostat encloses all the superconducting coils and the vacuum vessel. Liquid nitrogen cooled thermal shield between the vacuum vessel and superconducting coils as well as between cryostat and the superconducting coils reduce the radiation heat load on the superconducting coils.
The sketch showing relative positions of various components.
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 poloidal field coils are to be used for shape control. Other subsystems include radiofrequency systems for pre-ionization, current drive and heating, neutral beam injection system for supplementary heating, cryogenic systems at liquid helium and liquid nitrogen 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.
The above three dimensional sketch shows the relative positions of various components.
All superconducting coils have been successfully fabricated using a cable-in-conduit conductor (CICC) based on niobium-titanium (NbTi) and copper. The CICC has been fabricated by a Japanese firm under specification and supervision of IPR. In order to test the performance of this CICC under SST-1 operating scenarios, a Model Coil was designed, fabricated and tested at Kurchatov Institute(KI), Russia using the SST-1 CICC. The results obtained from these model coil tests have validated the CICC design parameters as well as its appropriateness as the base conductor for the SST-1 superconducting magnet systems.
The toroidal field coils are encased in a stainless steel casings to take care of forces acting on the coils. The coils and the casings have been manufactured by the Bharat Heavy Electicals Ltd., Bhopal with specifications and supervision from IPR. Such large size superconducting coils have been manufactured for the first time in the country. An insulation system, compatible with low temperature (4.5K) operation of these coils, and the winding technologies have been indigenously developed for these superconducting coils.
The superconducting magnet system, consisting of toroidal field and poloidal field coils, in SST-1 has to be maintained at 4.5 K in presence of steady state heat loads. In addition, the pulsed heat loads during the plasma operation have to be taken care of by the cooling system so as to maintain the magnets in superconducting state.
The magnets will be cooled using forced flow of supercritical helium through the void space in the CICC. Further the magnets have to be energized from power supplies at room temperature. A closed cycle 1 kW class He refrigerator/liquefier, has been deployed for this purpose The system is at present under commissioning tests at IPR. He gas management system, including high pressure and medium pressure storage vessels and recovery system, required for the He refrigerator/liquefier, has been commissioned. This is the biggest liquid helium system in the country at present.
In order to minimize the heat loads on magnets and support system at 4.5 K, liquid nitrogen shields are provided between the cold mass at 4.5 K and warmer surfaces. A liquid nitrogen management system, including liquid nitrogen storage and distribution system, has been commissioned for this purpose. An integrated flow distribution system for distribution of cryogens to magnets and radiation shield has been installed and is in final stages of testing.
CONTINUED-----
A steady state superconducting tokamak, SST-1, is in advanced stage of fabrication at the Institute for Plasma Research, Gandhinagar. The objectives of SST-1 include :
To study physics of plasma Processes in tokamak under steady-state conditions & contribute to the tokamak physics database for very long pulse operations.
Learning new technologies relevant to steady state tokamak operation.
Superconducting magnets and associated power supplies and protection system.
Large scale cryogenic system (Liquid helium and liquid nitrogen).
High Power Radio Frequency Systems.
Energetic Neutral Particle Beams.
High heat flux handling.
The machine has a major radius of 1.1 m, minor radius of 0.20 m, a toroidal field of 3.0 Tesla at plasma centre and a plasma current of 220 kA.
Controlled thermonuclear fusion is one of the attractive futuristic sources of energy. All over the world, the research in this field of energy has been continuing for the last fifty years.
Research efforts in this area are broadly divided into inertial confinement and magnetically confined plasmas. Among the magnetically confined systems, Tokamaks have been the most successful machines to achieve the technological goals .
In India, the scientific research in tokamak plasmas has been continuing for more than a decade now.
The tokamak Aditya developed at the Institute for Plasma Research, Gandhinagar, Gujarat, is one of the milestones of this endeavour. A steady state super-conducting tokamak, SST-1 is in advnced stage of fabrication at the Institute. Present here is the status of this venture.
Superconducting coils for both toroidal field and poloidal field are to be deployed in the SST-1 tokamak. NbTi superconductor at 4.5K is used for the superconducting magnets and maximum field at the conductor is restricted to 5.1 Tesla. An ultra high vacuum compatible vacuum vessel, placed in the bore of the toroidal field coils, houses the plasma facing components. A high vacuum cryostat encloses all the superconducting coils and the vacuum vessel. Liquid nitrogen cooled thermal shield between the vacuum vessel and superconducting coils as well as between cryostat and the superconducting coils reduce the radiation heat load on the superconducting coils.
The sketch showing relative positions of various components.
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 poloidal field coils are to be used for shape control. Other subsystems include radiofrequency systems for pre-ionization, current drive and heating, neutral beam injection system for supplementary heating, cryogenic systems at liquid helium and liquid nitrogen 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.
The above three dimensional sketch shows the relative positions of various components.
All superconducting coils have been successfully fabricated using a cable-in-conduit conductor (CICC) based on niobium-titanium (NbTi) and copper. The CICC has been fabricated by a Japanese firm under specification and supervision of IPR. In order to test the performance of this CICC under SST-1 operating scenarios, a Model Coil was designed, fabricated and tested at Kurchatov Institute(KI), Russia using the SST-1 CICC. The results obtained from these model coil tests have validated the CICC design parameters as well as its appropriateness as the base conductor for the SST-1 superconducting magnet systems.
The toroidal field coils are encased in a stainless steel casings to take care of forces acting on the coils. The coils and the casings have been manufactured by the Bharat Heavy Electicals Ltd., Bhopal with specifications and supervision from IPR. Such large size superconducting coils have been manufactured for the first time in the country. An insulation system, compatible with low temperature (4.5K) operation of these coils, and the winding technologies have been indigenously developed for these superconducting coils.
The superconducting magnet system, consisting of toroidal field and poloidal field coils, in SST-1 has to be maintained at 4.5 K in presence of steady state heat loads. In addition, the pulsed heat loads during the plasma operation have to be taken care of by the cooling system so as to maintain the magnets in superconducting state.
The magnets will be cooled using forced flow of supercritical helium through the void space in the CICC. Further the magnets have to be energized from power supplies at room temperature. A closed cycle 1 kW class He refrigerator/liquefier, has been deployed for this purpose The system is at present under commissioning tests at IPR. He gas management system, including high pressure and medium pressure storage vessels and recovery system, required for the He refrigerator/liquefier, has been commissioned. This is the biggest liquid helium system in the country at present.
In order to minimize the heat loads on magnets and support system at 4.5 K, liquid nitrogen shields are provided between the cold mass at 4.5 K and warmer surfaces. A liquid nitrogen management system, including liquid nitrogen storage and distribution system, has been commissioned for this purpose. An integrated flow distribution system for distribution of cryogens to magnets and radiation shield has been installed and is in final stages of testing.
CONTINUED-----
