WorldWideScience

Sample records for tokamak sst-1 diagnostics

  1. Application of automatic gain control for radiometer diagnostic in SST-1 tokamak.

    Science.gov (United States)

    Makwana, Foram R; Siju, Varsha; Edappala, Praveenlal; Pathak, S K

    2017-12-01

    This paper describes the characterisation of a negative feedback type of automatic gain control (AGC) circuit that will be an integral part of the heterodyne radiometer system operating at a frequency range of 75-86 GHz at SST-1 tokamak. The developed AGC circuit is a combination of variable gain amplifier and log amplifier which provides both gain and attenuation typically up to 15 dB and 45 dB, respectively, at a fixed set point voltage and it has been explored for the first time in tokamak radiometry application. The other important characteristics are that it exhibits a very fast response time of 390 ns to understand the fast dynamics of electron cyclotron emission and can operate at very wide input RF power dynamic range of around 60 dB that ensures signal level within the dynamic range of the detection system.

  2. Optimization design for SST-1 Tokamak insulators

    International Nuclear Information System (INIS)

    Zhang Yuanbin; Pan Wanjiang

    2012-01-01

    With the help of ANSYS FEA technique, high voltage and cryogenic proper- ties of the SST-1 Tokamak insulators were obtained, and the structure of the insulators was designed and modified by taking into account the simulation results. The simulation results indicate that the optimization structure has better high voltage insulating property and cryogenic mechanics property, and also can fulfill the qualification criteria of the SST-1 Tokamak insulators. (authors)

  3. Gas Fuelling System for SST-1Tokamak

    Science.gov (United States)

    Dhanani, Kalpesh; Raval, D. C.; Khan, Ziauddin; Semwal, Pratibha; George, Siju; Paravastu, Yuvakiran; Thankey, Prashant; Khan, M. S.; Pradhan, Subrata

    2017-04-01

    SST-1 Tokamak, the first Indian Steady-state Superconducting experimental device is at present under operation in the Institute for Plasma Research. For plasma break down & initiation, piezoelectric valve based gas feed system is implemented as a primary requirement due to its precise control, easy handling, low construction and maintenance cost and its flexibility in the selection of the working gas. Hydrogen gas feeding with piezoelectric valve is used in the SST-1 plasma experiments. The piezoelectric valves used in SST-1 are remotely driven by a PXI based platform and are calibrated before each SST-1 plasma operation with precise control. This paper will present the technical development and the results of the gas fuelling system of SST-1.

  4. Plasma position control in SST1 tokamak

    Indian Academy of Sciences (India)

    also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...

  5. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2005-01-01

    SST-1, a steady state superconducting tokamak, is undergoing commissioning tests at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in a 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. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. The auxiliary current drive is based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating 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. The ICRF system would also be used for initial breakdown and wall conditioning experiments. Detailed commissioning tests on the cryogenic system and experiments on the hydraulic characters and cool down features of single TF coils have been completed prior to the cool down of the entire superconducting system. Results of the single TF magnet cool down, and testing of the magnet system are presented. First experiments related to the breakdown and the current ramp up will subsequently be carried out. (author)

  6. Vessel eddy current characteristics in SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jana, Subrata; Pradhan, Subrata, E-mail: pradhan@ipr.res.in; Dhongde, Jasraj; Masand, Harish

    2016-11-15

    Highlights: • Eddy current distribution in the SST-1 vacuum vessel. • Circuit model analysis of eddy current. • A comparison of the field lines with and without the plasma column in identical conditions. • The influence of eddy current in magnetic NULL dynamics. - Abstract: Eddy current distribution in the vacuum vessel of the Steady state superconducting (SST-1) tokamak has been determined from the experimental data obtained using an array of internal voltage loops (flux loop) installed inside the vacuum vessel. A simple circuit model has been employed. The model takes into account the geometric and constructional features of SST-1 vacuum vessel. SST-1 vacuum vessel is a modified ‘D’ shaped vessel having major axis of 1.285 m and minor axis of 0.81 m and has been manufactured from non-magnetic stainless steel. The Plasma facing components installed inside the vacuum vessel are graphite blocks mounted on Copper Chromium Zirconium (CuCrZr) heat sink plates on inconel supports. During discharge of the central solenoid, eddy currents get generated in the vacuum vessel and passive supports on it. These eddy currents influence the early magnetic NULL dynamics and plasma break-down and start-up characteristics. The computed results obtained from the model have been benchmarked against experimental data obtained in large number of SST-1 plasma shots. The results are in good agreement. Once bench marked, the calculated eddy current based on flux loop signal and circuit equation model has been extended to the reconstruction of the overall B- field contours of SST-1 tokamak in the vessel region. A comparison of the field lines with and without the plasma column in identical conditions of the central solenoid and equilibrium field profiles has also been done with an aim to quantify the diagnostics responses in vacuum shots.

  7. Vacuum system of SST-1 Tokamak

    International Nuclear Information System (INIS)

    Khan, Ziauddin; Pathan, Firozkhan; George, Siju; Semwal, Pratibha; Dhanani, Kalpesh; Paravastu, Yuvakiran; Thankey, Prashant; Ramesh, Gattu; Himabindu, Manthena; Pradhan, Subrata

    2013-01-01

    Highlights: ► Air leaks developed during ongoing SST-1 cooldown campaign were detected online using RGA. ► The presence of N 2 and O 2 gases with the ratio of their partial pressures with ∼3.81:1 confirmed the air leaks. ► Baking of SST-1 was done efficiently by flowing hot N 2 gas in C-channels welded on inner surfaces without any problem. ► In-house fabricated demountable bull nose couplers were demonstrated for high temperature and pressure applications. ► Cryopumping effect was observed when liquid helium cooled superconducting magnets reached below 63 K. -- Abstract: Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN 2 cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10 −4 mbar and 1.0 × 10 −5 mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10 −6 mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10 −5 mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ∼3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10 −5 mbar is achieved inside the cryostat. Baking of the vacuum vessel up to 110 °C with ±10

  8. Vacuum system of SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Ziauddin, E-mail: ziauddin@ipr.res.in [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382 428 (India); Pathan, Firozkhan; George, Siju; Semwal, Pratibha; Dhanani, Kalpesh; Paravastu, Yuvakiran; Thankey, Prashant; Ramesh, Gattu; Himabindu, Manthena; Pradhan, Subrata [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382 428 (India)

    2013-10-15

    Highlights: ► Air leaks developed during ongoing SST-1 cooldown campaign were detected online using RGA. ► The presence of N{sub 2} and O{sub 2} gases with the ratio of their partial pressures with ∼3.81:1 confirmed the air leaks. ► Baking of SST-1 was done efficiently by flowing hot N{sub 2} gas in C-channels welded on inner surfaces without any problem. ► In-house fabricated demountable bull nose couplers were demonstrated for high temperature and pressure applications. ► Cryopumping effect was observed when liquid helium cooled superconducting magnets reached below 63 K. -- Abstract: Vacuum chambers of Steady State Superconducting (SST-1) Tokamak comprises of the vacuum vessel and the cryostat. The plasma will be confined inside the vacuum vessel while the cryostat houses the superconducting magnet systems (TF and PF coils), LN{sub 2} cooled thermal shields and hydraulics for these circuits. The vacuum vessel is an ultra-high (UHV) vacuum chamber while the cryostat is a high-vacuum (HV) chamber. In order to achieve UHV inside the vacuum vessel, it would be baked at 150 °C for longer duration. For this purpose, U-shaped baking channels are welded inside the vacuum vessel. The baking will be carried out by flowing hot nitrogen gas through these channels at 250 °C at 4.5 bar gauge pressure. During plasma operation, the pressure inside the vacuum vessel will be raised between 1.0 × 10{sup −4} mbar and 1.0 × 10{sup −5} mbar using piezoelectric valves and control system. An ultimate pressure of 4.78 × 10{sup −6} mbar is achieved inside the vacuum vessel after 100 h of pumping. The limitation is due to the development of few leaks of the order of 10{sup −5} mbar l/s at the critical locations of the vacuum vessel during baking which was confirmed with the presence of nitrogen gas and oxygen gas with the ratio of ∼3.81:1 indicating air leak. Similarly an ultimate vacuum of 2.24 × 10{sup −5} mbar is achieved inside the cryostat. Baking of the

  9. Nitrogen Gas Heating and Supply System for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Khan, Ziauddin; Pathan, Firozkhan; Paravastu, Yuvakiran; George, Siju; Ramesh, Gattu; Bindu, Hima; Raval, Dilip C.; Thankey, Prashant; Dhanani, Kalpesh; Pradhan, Subrata

    2013-01-01

    Steady State Tokamak (SST-1) vacuum vessel baking as well as baking of the first wall components of SST-1 are essential to plasma physics experiments. Under a refurbishment spectrum of SST-1, the nitrogen gas heating and supply system has been fully refurbished. The SST-1 vacuum vessel consists of ultra-high vacuum (UHV) compatible eight modules and eight sectors. Rectangular baking channels are embedded on each of them. Similarly, the SST-1 plasma facing components (PFC) are comprised of modular graphite diverters and movable graphite based limiters. The nitrogen gas heating and supply system would bake the plasma facing components at 350°C and the SST-1 vacuum vessel at 150°C over an extended duration so as to remove water vapour and other absorbed gases. An efficient PLC based baking facility has been developed and implemented for monitoring and control purposes. This paper presents functional and operational aspects of a SST-1 nitrogen gas heating and supply system. Some of the experimental results obtained during the baking of SST-1 vacuum modules and sectors are also presented here. (fusion engineering)

  10. Superconducting magnets of SST-1 tokamak

    International Nuclear Information System (INIS)

    Subrata Pradhan; Saxena, Y.C.; Sarkar, B.; Bansal, G.; Sharma, A.N.; Thomas, K.J.; Bedakihale, V.; Doshi, B.; Dhard, C.P.; Prasad, U.; Rathod, P.; Bahl, R.; Varadarajulu, A.; Mankani, A.

    2005-01-01

    Magnet System of SST-1 comprises of sixteen superconducting D-shaped Toroidal Field (TF) coils, nine superconducting Poloidal Field (PF) coils and a pair of resistive PF coils inside the vacuum vessel. TF magnets generate the basic 3.0 T field at the major radius of 1.1 m. Low resistance lap inter-pancake joints within and inter-coil joints between the coils have been made. Magnets are cooled with supercritical helium at 4 bar and 4.5 K, which is fed at the high field region in the middle of each of the double pancake over a hydraulic path length of 47 m. Voltage taps across joints and termination location are used for quench detection. The quench detection front-end electronics ensures fail proof quench detection based on subtraction logic. Quench detection system sends the quench trigger to the power supply system directly on a dedicated fiber optic link. Flow meters at the inlet of the TF and PF magnets, temperature sensors at the critical joint locations and at the outlet of the flow paths for enthalpy estimation, hall probes for field direction and magnitude measurements are the other sensors. A 20 V, 10 kA power supply will excite the TF magnets whereas the PF power supplies have voltages from few volts to in excess of 100 V to cater the fast current ramp-up of the PF magnets during start-up scenarios. All power supplies have been equipped with dump resisters of appropriate ratings in parallel with a series combination of DC circuit interrupters and pyro-breakers. (author)

  11. Overview of data acquisition system for SST-1 diagnostics

    International Nuclear Information System (INIS)

    Sharma, Manika; Mansuri, Imran; Raval, Tushar; Sharma, A.L; Pradhan, S.

    2016-01-01

    Highlights: • An account of architecture and data acquisition activities of SST-1 data acquisition system (DAS) for SST-1 diagnostics and subsystems. • PXI based Data acquisition system and CAMAC based Data acquisition system for slow and fast plasma diagnostics. • SST-1 DAS interface and its communication with SST-1 central control system. Integration of SST-1 DAS with timing system. • SST-1 DAS data archival and data analysis. - Abstract: The recent first phase operations of SST-1 in short pulse mode have provided an excellent opportunity for the essential initial tests and benchmark of the SST-1 Data Acquisition System. This paper describes the SST-1 Data Acquisition systems (DAS), which with its heterogeneous composition and distributed architecture, aims to cover a wide range of slow to fast channels interfaced with a large set of diagnostics. The DAS also provides the essential user interface for data acquisition to cater both on and off-line data usage. The central archiving and retrieval service is based on a dual step architecture involving a combination of Network Attached Server (NAS) and a Storage Area Network (SAN). SST-1 Data Acquisition Systems have been reliably operated in the SST-1 experimental campaigns. At present different distributed DAS caters the need of around 130 channels from different SST-1 diagnostics and its subsystems. PXI based DAS and CAMAC based DAS have been chosen to cater the need, with sampling rates varying from 10Ksamples/sec to 1Msamples/sec. For these large sets of channels acquiring from individual diagnostics and subsystems has been a combined setup, subjected to a gradual phase of optimization and tests resulting into a series of improvisations over the recent operations. In order to facilitate a reliable data acquisition, the model further integrates the objects of the systems with the Central Control System of SST-1 using the TCP/IP communication. The associated DAS software essentially addresses the

  12. Overview of data acquisition system for SST-1 diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Manika, E-mail: bithi@ipr.res.in; Mansuri, Imran; Raval, Tushar; Sharma, A.L; Pradhan, S.

    2016-11-15

    Highlights: • An account of architecture and data acquisition activities of SST-1 data acquisition system (DAS) for SST-1 diagnostics and subsystems. • PXI based Data acquisition system and CAMAC based Data acquisition system for slow and fast plasma diagnostics. • SST-1 DAS interface and its communication with SST-1 central control system. Integration of SST-1 DAS with timing system. • SST-1 DAS data archival and data analysis. - Abstract: The recent first phase operations of SST-1 in short pulse mode have provided an excellent opportunity for the essential initial tests and benchmark of the SST-1 Data Acquisition System. This paper describes the SST-1 Data Acquisition systems (DAS), which with its heterogeneous composition and distributed architecture, aims to cover a wide range of slow to fast channels interfaced with a large set of diagnostics. The DAS also provides the essential user interface for data acquisition to cater both on and off-line data usage. The central archiving and retrieval service is based on a dual step architecture involving a combination of Network Attached Server (NAS) and a Storage Area Network (SAN). SST-1 Data Acquisition Systems have been reliably operated in the SST-1 experimental campaigns. At present different distributed DAS caters the need of around 130 channels from different SST-1 diagnostics and its subsystems. PXI based DAS and CAMAC based DAS have been chosen to cater the need, with sampling rates varying from 10Ksamples/sec to 1Msamples/sec. For these large sets of channels acquiring from individual diagnostics and subsystems has been a combined setup, subjected to a gradual phase of optimization and tests resulting into a series of improvisations over the recent operations. In order to facilitate a reliable data acquisition, the model further integrates the objects of the systems with the Central Control System of SST-1 using the TCP/IP communication. The associated DAS software essentially addresses the

  13. Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2000-01-01

    SST-1 is a steady state superconducting tokamak for studying the physics of the plasma processes in tokamak under steady state conditions and to learn technologies related to the steady state operation of the tokamak. SST-1 will have superconducting magnets made from NbTi based conductors operating at 4.5 K temperature. The design of the superconducting magnets and the cryogenic system of SST-1 tokamak are described. (author)

  14. Test results on systems developed for SST-1 tokamak

    International Nuclear Information System (INIS)

    Bora, D.

    2003-01-01

    Steady state Superconducting Tokamak (SST-1) is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation (κ) and triangularity Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. A NbTi based cable-in-conduit conductor (CICC) has been fabricated by M/S Hitachi Cables Ltd., Japan under specification and supervision of IPR. The suitability of this CICC for the SST-1 magnets has been validated through test carried out on a model coil (MC) wound from this CICC. Toroidal and poloidal SC magnets have been fabricated and factory acceptance tests have been performed. SC magnets require liquid helium (LHe) cooled current leads, electrical isolators at LHe temperature, superconducting bus bars and LHe transfer lines. Full scale prototypes of these have been developed and tested successfully. SC magnets will be cooled to 4.5K by forced flow of supercritical Helium through the CICC. A 1 kW grade liquefier/refrigerator has been installed and is in final stages of commissioning at IPR. SST-1 deploys a fully welded ultra high vacuum vessel, made up of 16 vessel sectors having ports and 16 rings with D-shaped cross-section. To establish the fabrication methodology for this, a full scale proto-type of the vessel with two vessel sectors and three rings has been fabricated and tested successfully. Based on this the fabrication of the vessel sectors and rings is in final stage of fabrication. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. SST-1 will have three different high power radio frequency (RF) systems to additionally heat and non-inductively drive plasma current to sustain the plasma in steady state for a duration of up to 1000 sec. Ion cyclotron resonance frequency (ICRF) and electron cyclotron resonance frequency (ECRF) systems will primarily be

  15. Development and Integration of a Data Acquisition System for SST-1 Phase-1 Plasma Diagnostics

    International Nuclear Information System (INIS)

    Srivastava, Amit K; Sharma, Manika; Mansuri, Imran; Sharma, Atish; Raval, Tushar; Pradhan, Subrata

    2012-01-01

    Long pulse (of the order of 1000 s or more) SST-1 tokamak experiments demand a data acquisition system that is capable of acquiring data from various diagnostics channels without losing useful data (and hence physics information) while avoiding unnecessary generation of a large volume data. SST-1 Phase-1 tokamak operation has been envisaged with data acquisition of several essential diagnostics channels. These channels demand data acquisition at a sampling rate ranging from 1 kilo samples per second (KSPS) to 1 mega samples per second (MSPS). Considering the technical characteristics and requirements of the diagnostics, a data acquisition system based on PXI and CAMAC has been developed for SST-1 plasma diagnostics. Both these data acquisition systems are scalable. Present data acquisition needs involving slow plasma diagnostics are catered by the PXI based data acquisition system. On the other hand, CAMAC data acquisition hardware meets all requirements of the SST-1 Phase-1 fast plasma diagnostics channels. A graphical user interface for both data acquisition systems (PXI and CAMAC) has been developed using LabVIEW application development software. The collected data on the local hard disk are directly streaming to the central server through a dedicated network for post-shot data analysis. This paper describes the development and integration of the data acquisition system for SST-1 Phase-1 plasma diagnostics. The integrated testing of the developed data acquisition system has been performed using SST-1 central control and diagnostics signal conditioning units. In the absence of plasma shots, the integrated testing of the data acquisition system for the initial diagnostics of SST-1 Phase-1 operation has been performed with simulated physical signals. The primary engineering objective of this integrated testing is to validate the performance of the developed data acquisition system under simulated conditions close to that of actual tokamak operation. The data

  16. Design of plasma facing components for the SST-1 tokamak

    International Nuclear Information System (INIS)

    Jacob, S.; Chenna Reddy, D.; Choudhury, P.; Khirwadkar, S.; Pragash, R.; Santra, P.; Saxena, Y.C.; Sinha, P.

    2000-01-01

    Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m 2 . (author)

  17. Design of new superconducting central solenoid of SST-1 tokamak

    International Nuclear Information System (INIS)

    Prasad, Upendra; Pradhan, Subrata; Ghate, Mahesh

    2015-01-01

    The key role of the central solenoid (CS) magnet of a Tokamak is for gas breakdown, ramp up and maintaining of plasma current for longer duration. The magnetic flux change in CS along with other PF coils generates magnetic null and induces electric field in toroidal direction. The induced toroidal electric field accelerates the residual electrons which collide with the neutrals and an avalanche takes place which led to the net plasma in the vacuum vessel of a Tokamak. In order to maximize the CS volt-sec capability, the higher magnetic field with a greater magnetic flux linkage is necessary. In order to facilitate all these requirements of SST-1 a new superconducting CS has been designed for SST-1. The design of new central solenoid has two bases; first one is physics and second is smart engineering in limited bore diameter of ∼655 mm. The physics basis of the design includes volt-sec storage capacity of ∼0.8 volt-sec, magnetic field null around 0.2 m over major radius of 1.1 m and toroidal electric field of ∼0.3 volt/m.The engineering design of new CS consists of Nb 3 Sn cable in conduit conductor (CICC) of operating current of 14 kA @ 4.5 K at 6 T, consolidated winding pack, smart quench detection system, protection system, housing cryostat and conductor terminations and joint design. The winding pack consists of 576 numbers of turns distributed in four layers with 0.75 mm FRP tape soaked with cyanide Easter based epoxy resin turn insulation and 3 mm of ground insulation. The inter-layer low resistance (∼1 nΩ) at 14 kA @ 4.5 K terminal praying hand joints has been designed for making winding pack continuous. The total height of winding pack is 2500 mm. The stored energy of this winding pack is ∼3 MJ at 14 kA of operating current. The expected heat load at cryogenic temperature is ∼10 W per layer, which requires helium mass flow rate of 1.4 g/s at 1.4 bars @ 4.5 K. The typical diameter and height of housing cryostat are 650 mm and 2563 mm with 80 K

  18. Overview of time synchronization system of steady state superconducting tokamak SST-1

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A., E-mail: aveg@ipr.res.in; Masand, H.; Dhongde, J.; Patel, K.; Mahajan, K.; Gulati, H.; Bhandarkar, M.; Chudasama, H.; Pradhan, S.

    2016-11-15

    The Steady State Superconducting Tokamak (SST-1) consists of many distributed and heterogeneous plant/experiment systems viz. Water-Cooling, Power Supplies, Cryogenics, Vacuum, Magnets, Auxiliary-Heating sources, Diagnostics, Front End Electronics (FEE) & Data Acquisition systems, having their own data acquisition & control systems and control & monitor by Central Control System (CCS) during the machine operation. With distributed computing and interdependent systems, it is essential that all the data/event acquired must be with disciplined & precise time-base, so as to make the co-relation of the data/event from various plant and experiment systems easy. Hence it is important to have accurate and precise Time Synchronization in place. The two systems fulfill the requirement of the time synchronization in SST-1. The VME based Timing System (TS) provides synchronization amongst various experiment systems during the plasma discharges and works as discharge control system (DCS) while the GPS based Time Synchronization System (TSS) caters the requirement of synchronization during the continuous operation of various plant systems by feeding a central clock to all the plant systems. This paper presents the Time Synchronization System of SST-1, the results of the integrated testing and engineering validation with various SST-1 subsystems.

  19. CORBA-based solution for remote participation in SST-1 tokamak control and operation

    Energy Technology Data Exchange (ETDEWEB)

    Mahajan, Kirti [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India)]. E-mail: kirti@ipr.res.in; Ravikiran, M. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Gulati, Hitesh [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Dave, H.J. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Kumar, Neeraj [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Patel, Kirit [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Kumar, Aveg [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Raju, D. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Bhandarkar, M. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Chudasama, H. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Kulkarni, S.V. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Saxena, Y.C. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India)

    2006-07-15

    The steady state superconducting tokamak (SST-1) central control system is a distributed heterogeneous process communication system built on socket programming. It consists of machine, experiment and discharge control plus timing and a database. The software controls and monitors SST-1 subsystems: water-cooling, power supplies, cryogenics and vacuum over a local area network (LAN). The SST-1 control room is the place where all the activities like session announcement, machine control, experiment control, discharge control and monitoring are performed. We have realized that, instead of having a single monitoring place, we should have multiple monitoring points and it should be made possible to control the experiment from any PC over the LAN. In order to meet such requirements for remote participation in tokamak operation, we are upgrading the existing software. The upgraded software is based on Common Object Request Broker Architecture (CORBA) technology. The software is utilizing CORBA-services such as event service, naming services, interface repository and security services. The inherent features of CORBA make the software, platform and language independent. The software supports a variety of communication paradigms including publish-subscribe, peer-to-peer, and request-reply. Based on this software, one can participate in SST-1 tokamak operation from the LAN, or a wide area network (WAN) connection anywhere on the Internet. Each user can customize plasma parameters and diagnostics data that he wants to monitor, at any time without any change in the software and a copy of these parameters will be available to him. This paper focuses on the publish-subscribe communication paradigm and its application for a machine monitoring system.

  20. CORBA-based solution for remote participation in SST-1 tokamak control and operation

    International Nuclear Information System (INIS)

    Mahajan, Kirti; Ravikiran, M.; Gulati, Hitesh; Dave, H.J.; Kumar, Neeraj; Patel, Kirit; Kumar, Aveg; Raju, D.; Bhandarkar, M.; Chudasama, H.; Kulkarni, S.V.; Saxena, Y.C.

    2006-01-01

    The steady state superconducting tokamak (SST-1) central control system is a distributed heterogeneous process communication system built on socket programming. It consists of machine, experiment and discharge control plus timing and a database. The software controls and monitors SST-1 subsystems: water-cooling, power supplies, cryogenics and vacuum over a local area network (LAN). The SST-1 control room is the place where all the activities like session announcement, machine control, experiment control, discharge control and monitoring are performed. We have realized that, instead of having a single monitoring place, we should have multiple monitoring points and it should be made possible to control the experiment from any PC over the LAN. In order to meet such requirements for remote participation in tokamak operation, we are upgrading the existing software. The upgraded software is based on Common Object Request Broker Architecture (CORBA) technology. The software is utilizing CORBA-services such as event service, naming services, interface repository and security services. The inherent features of CORBA make the software, platform and language independent. The software supports a variety of communication paradigms including publish-subscribe, peer-to-peer, and request-reply. Based on this software, one can participate in SST-1 tokamak operation from the LAN, or a wide area network (WAN) connection anywhere on the Internet. Each user can customize plasma parameters and diagnostics data that he wants to monitor, at any time without any change in the software and a copy of these parameters will be available to him. This paper focuses on the publish-subscribe communication paradigm and its application for a machine monitoring system

  1. Overview of data acquisition and central control system of steady state superconducting Tokamak (SST-1)

    International Nuclear Information System (INIS)

    Pradhan, S.; Mahajan, K.; Gulati, H.K.; Sharma, M.; Kumar, A.; Patel, K.; Masand, H.; Mansuri, I.; Dhongde, J.; Bhandarkar, M.; Chudasama, H.

    2016-01-01

    Highlights: • The paper gives overview on SST-1 data acquisition and central control system and future upgrade plans. • The lossless PXI based data acquisition of SST-1 is capable of acquiring around 130 channels with sampling frequency ranging from 10 KHz to 1 MHz sampling frequency. • Design, architecture and technologies used for central control system (CCS) of SST-1. • Functions performed by CCS. - Abstract: Steady State Superconducting Tokamak (SST-1) has been commissioned successfully and has been carrying out limiter assisted ohmic plasma experiments since the beginning of 2014 achieving a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ∼500 ms. In near future, SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1 s. The data acquisition and central control system (CCS) for SST-1 are distributed, modular, hierarchical and scalable in nature The CCS has been indigenously designed, developed, implemented, tested and validated for the operation of SST-1. The CCS has been built using well proven technologies like Redhat Linux, vxWorks RTOS for deterministic control, FPGA based hardware implementation, Ethernet, fiber optics backbone for network, DSP for real-time computation & Reflective memory for high-speed data transfer etc. CCS in SST-1 controls & monitors various heterogeneous SST-1 subsystems dispersed in the same campus. The CCS consists of machine control system, basic plasma control system, GPS time synchronization system, storage area network (SAN) for centralize data storage, SST-1 networking system, real-time networks, SST-1 control room infrastructure and many other supportive systems. Machine Control System (MCS) is a multithreaded event driven system running on Linux based servers, where each thread of the software communicates to a unique subsystem for monitoring and control from SST-1 central control room through network programming. The CCS hardware

  2. Overview of data acquisition and central control system of steady state superconducting Tokamak (SST-1)

    Energy Technology Data Exchange (ETDEWEB)

    Pradhan, S., E-mail: pradhan@ipr.res.in; Mahajan, K.; Gulati, H.K.; Sharma, M.; Kumar, A.; Patel, K.; Masand, H.; Mansuri, I.; Dhongde, J.; Bhandarkar, M.; Chudasama, H.

    2016-11-15

    Highlights: • The paper gives overview on SST-1 data acquisition and central control system and future upgrade plans. • The lossless PXI based data acquisition of SST-1 is capable of acquiring around 130 channels with sampling frequency ranging from 10 KHz to 1 MHz sampling frequency. • Design, architecture and technologies used for central control system (CCS) of SST-1. • Functions performed by CCS. - Abstract: Steady State Superconducting Tokamak (SST-1) has been commissioned successfully and has been carrying out limiter assisted ohmic plasma experiments since the beginning of 2014 achieving a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ∼500 ms. In near future, SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1 s. The data acquisition and central control system (CCS) for SST-1 are distributed, modular, hierarchical and scalable in nature The CCS has been indigenously designed, developed, implemented, tested and validated for the operation of SST-1. The CCS has been built using well proven technologies like Redhat Linux, vxWorks RTOS for deterministic control, FPGA based hardware implementation, Ethernet, fiber optics backbone for network, DSP for real-time computation & Reflective memory for high-speed data transfer etc. CCS in SST-1 controls & monitors various heterogeneous SST-1 subsystems dispersed in the same campus. The CCS consists of machine control system, basic plasma control system, GPS time synchronization system, storage area network (SAN) for centralize data storage, SST-1 networking system, real-time networks, SST-1 control room infrastructure and many other supportive systems. Machine Control System (MCS) is a multithreaded event driven system running on Linux based servers, where each thread of the software communicates to a unique subsystem for monitoring and control from SST-1 central control room through network programming. The CCS hardware

  3. Conditioning of SST-1 Tokamak Vacuum Vessel by Baking and Glow Discharge Cleaning

    International Nuclear Information System (INIS)

    Khan, Ziauddin; George, Siju; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; Pathan, Firozkhan S.; Paravastu, Yuvakiran; Raval, Dilip C.; Babu, Gattu Ramesh; Khan, Mohammed Shoaib; Pradhan, Subrata

    2016-01-01

    Highlights: • SST-1 Tokamak was successfully commissioned. • Vacuum vessel was pumped down to 4.5 × 10"–"8 mbar after baking and continuous GDC. • GDC reduced the water vapour by additional 57% while oxygen was reduced by 50%. • Under this condition, an initial plasma breakdown with current of 40 kA for 75 ms was achieved. - Abstract: Steady-state Superconducting Tokamak (SST-1) vacuum vessel (VV) adopts moderate baking at 110 ± 10 °C and the limiters baking at 250 ± 10 °C for ∼ 200 h followed by glow discharge cleaning in hydrogen (GDC-H) with 0.15 A/m"2 current density towards its conditioning prior to plasma discharge experiment. The baking in SST-1 reduces the water (H_2O) vapor by 95% and oxygen (O_2) by 60% whereas the GDC reduces the water vapor by an additional 57% and oxygen by another 50% as measured with residual gas analyzer. The minimum breakdown voltage for H-GDC in SST-1 tokamak was experimentally observed to 300 V at 8 mbar cm. As a result of these adherences, SST-1 VV achieves an ultimate of 4.5 × 10"−"8 mbar with two turbo-molecular pumps with effective pumping speed of 3250 l/s. In the last campaign, SST-1 has achieved successful plasma breakdown, impurity burn through and a plasma current of ∼ 40 kA for 75 ms.

  4. Soft-X ray electronics for temperature measurement in SST-1 tokamak

    International Nuclear Information System (INIS)

    Kumari, Praveena; Raval, Jayesh V.; Chauhan, Harsad; Hansalia, C.J.; Joisa, Y.S.; Rajpal, Rachana

    2015-01-01

    Soft-X ray diagnostic is used for the measurement of core temperature of plasma in tokamak. Signal conditioning electronics is designed, developed and tested for Soft-X ray measurement in SST-1. Silicon Surface Barrier Detectors (SBD) are used for detection of Soft -X ray. The detector is very sensitive and have a large leakage current (1-10) nA/cm"2. The preamplifier is designed to measure (10-100) nA of current signal. Virtual bias is supplied to detector through preamplifier. The front end electronics are mounted directly on the feed through in air side. Detectors are interfaced with feed through by 2-wire shielded cable. In the way of getting good results, problems are identified and troubleshooted. Soft-X ray signals are observed consistently in SST-1 campaign XIII. Different scheme were tested during the plasma experimental shots to get better measurement. This poster will describe the design details, interfacing with detector, problem faced, remedy and results. (author)

  5. Development of lab scale fast gas injection system for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Pathan, F.S.; Banaudha, Moni; Khristi, Yohan; Khan, M.S.; Khan, Ziauddin; Raval, D.C.; Khirwadkar, Samir

    2017-01-01

    The plasma density control plays an important role in Tokamak operation. The factors that influence plasma density in a Tokamak device are working gas injection, pumping, ionization rate and the recycle coefficient representing the wall conditions. Among these factors, gas injection is relatively convenient to be controlled. Hence, the most frequently adopted method to control the plasma density is to control the fast gas injection. This paper describes the design and experimental work carried out towards the development of Fast Gas Injection System for SST-1 Tokamak. Laboratory based test setup was successfully established for Fast Gas Injection System that can feed predefined quantity of gas in a controlled manner into vacuum chamber. Further, this FGIS system will be implemented in SST-1 Tokamak environment with online density feedback signal

  6. Conditioning of SST-1 Tokamak Vacuum Vessel by Baking and Glow Discharge Cleaning

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Ziauddin, E-mail: ziauddin@ipr.res.in; George, Siju; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; Pathan, Firozkhan S.; Paravastu, Yuvakiran; Raval, Dilip C.; Babu, Gattu Ramesh; Khan, Mohammed Shoaib; Pradhan, Subrata

    2016-02-15

    Highlights: • SST-1 Tokamak was successfully commissioned. • Vacuum vessel was pumped down to 4.5 × 10{sup –8} mbar after baking and continuous GDC. • GDC reduced the water vapour by additional 57% while oxygen was reduced by 50%. • Under this condition, an initial plasma breakdown with current of 40 kA for 75 ms was achieved. - Abstract: Steady-state Superconducting Tokamak (SST-1) vacuum vessel (VV) adopts moderate baking at 110 ± 10 °C and the limiters baking at 250 ± 10 °C for ∼ 200 h followed by glow discharge cleaning in hydrogen (GDC-H) with 0.15 A/m{sup 2} current density towards its conditioning prior to plasma discharge experiment. The baking in SST-1 reduces the water (H{sub 2}O) vapor by 95% and oxygen (O{sub 2}) by 60% whereas the GDC reduces the water vapor by an additional 57% and oxygen by another 50% as measured with residual gas analyzer. The minimum breakdown voltage for H-GDC in SST-1 tokamak was experimentally observed to 300 V at 8 mbar cm. As a result of these adherences, SST-1 VV achieves an ultimate of 4.5 × 10{sup −8} mbar with two turbo-molecular pumps with effective pumping speed of 3250 l/s. In the last campaign, SST-1 has achieved successful plasma breakdown, impurity burn through and a plasma current of ∼ 40 kA for 75 ms.

  7. Non-inductive current drive and RF heating in SST-1 tokamak

    International Nuclear Information System (INIS)

    2000-01-01

    Steady state superconducting tokamak (SST-1) machine is being developed for 1000 sec operation at different operating parameters. Radio Frequency (RF) and neutral beam injection (NBI) methods are planned in SST-1 for noninductive current drive and heating. In this paper, we describe the non-inductive current drive and RF heating methods that are being developed for this purpose. SST-1 is a large aspect ratio tokamak configured to run double-null divertor plasmas with significant elongation (κ = 1.7-1.9) and triangularity (δ = 0.4-0.7). SST-1 has a major radius of 1.1 in and minor radius of 0.2 m. Circular and shaped plasma experiments would be conducted at 1.5 and 3 T toroidal magnetic field in three different phases with I p = 110 kA and 220 kA. Two main factors have been considered during the development of auxiliary systems, namely, high heat flux (1 MW/m 2 ) incident on the plasma facing antennae components and fast feedback for constant power input due to small energy confinement time (∼ 10 ms). (author)

  8. PXIe based data acquisition and control system for ECRH systems on SST-1 and Aditya tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Patel, Jatinkumar J., E-mail: jatin@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar (India); Shukla, B.K.; Rajanbabu, N.; Patel, H.; Dhorajiya, P.; Purohit, D. [Institute for Plasma Research, Bhat, Gandhinagar (India); Mankadiya, K. [Optimized Solutions Pvt. Ltd (India)

    2016-11-15

    Highlights: • Data Aquisition and control system (DAQ). • PXIe hardware–(PXI–PCI bus extension for Instrumention Express). • RHVPS–Regulated High Voltage Power supply. • SST1–Steady state superconducting tokamak. - Abstract: In Steady State Superconducting (SST-1) tokamak, various RF heating sub-systems are used for plasma heating experiments. In SST-1, Two Electron Cyclotron Resonance Heating (ECRH) systems have been installed for pre-ionization, heating and current drive experiments. The 42 GHz gyrotron based ECRH system is installed and in operation with SST-1 plasma experiments. The 82.6 GHz gyrotron delivers 200 kW CW power (1000 s) while the 42 GHz gyrotron delivers 500 kW power for 500 ms duration. Each gyrotron system consists of various auxiliary power supplies, the crowbar unit and the water cooling system. The PXIe (PCI bus extension for Instrumentation Express)bus based DAC (Data Acquisition and Control) system has been designed, developed and under implementation for safe and reliable operation of the gyrotron. The Control and Monitoring Software applications have been developed using NI LabView 2014 software with real time support on windows platform.

  9. Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhuri, Paritosh E-mail: paritosh@ipr.res.in; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C

    2001-09-01

    The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m{sup 2}. In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper.

  10. Design and thermal-hydraulic analysis of PFC baking for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Reddy, D. Chenna; Khirwadkar, S.; Prakash, N. Ravi; Santra, P.; Saxena, Y.C.

    2001-01-01

    The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m 2 . In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper

  11. PXIe based data acquisition and control system for ECRH systems on SST-1 and Aditya tokamak

    International Nuclear Information System (INIS)

    Patel, Jatinkumar J.; Shukla, B.K.; Rajanbabu, N.; Patel, H.; Dhorajiya, P.; Purohit, D.; Mankadiya, K.

    2016-01-01

    Highlights: • Data Aquisition and control system (DAQ). • PXIe hardware–(PXI–PCI bus extension for Instrumention Express). • RHVPS–Regulated High Voltage Power supply. • SST1–Steady state superconducting tokamak. - Abstract: In Steady State Superconducting (SST-1) tokamak, various RF heating sub-systems are used for plasma heating experiments. In SST-1, Two Electron Cyclotron Resonance Heating (ECRH) systems have been installed for pre-ionization, heating and current drive experiments. The 42 GHz gyrotron based ECRH system is installed and in operation with SST-1 plasma experiments. The 82.6 GHz gyrotron delivers 200 kW CW power (1000 s) while the 42 GHz gyrotron delivers 500 kW power for 500 ms duration. Each gyrotron system consists of various auxiliary power supplies, the crowbar unit and the water cooling system. The PXIe (PCI bus extension for Instrumentation Express)bus based DAC (Data Acquisition and Control) system has been designed, developed and under implementation for safe and reliable operation of the gyrotron. The Control and Monitoring Software applications have been developed using NI LabView 2014 software with real time support on windows platform.

  12. Baking and helium glow discharge cleaning of SST-1 tokamak with graphite plasma facing components

    International Nuclear Information System (INIS)

    Semwal, Pratibha; Khan, Ziauddin; Raval, Dilip

    2015-01-01

    Graphite plasma facing components (PFCs) were installed inside SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 X 10 -5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium (He) glow discharge cleaning (GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nanometers from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.48. In this paper, the results of effect of baking and He-GDC experiments of SST-1 will be presented in detail. (author)

  13. Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

    International Nuclear Information System (INIS)

    Semwal, P; Khan, Z; Raval, D C; Dhanani, K R; George, S; Paravastu, Y; Prakash, A; Thankey, P; Ramesh, G; Khan, M S; Saikia, P; Pradhan, S

    2017-01-01

    Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10 -5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail. (paper)

  14. Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

    Science.gov (United States)

    Semwal, P.; Khan, Z.; Raval, D. C.; Dhanani, K. R.; George, S.; Paravastu, Y.; Prakash, A.; Thankey, P.; Ramesh, G.; Khan, M. S.; Saikia, P.; Pradhan, S.

    2017-04-01

    Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10-5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail.

  15. Conceptual design of plasma position control of SST-1 tokamak using vertical field coil

    International Nuclear Information System (INIS)

    Gulati, Hitesh Kumar; Patel, Kiritkumar B.; Dhongde, Jasraj

    2015-01-01

    SST-1 (Steady State Superconducting Tokamak) is a plasma confinement device in Institute for Plasma Research (IPR) India. SST-1 has been commissioned successfully and has been carrying out plasma experiments since the beginning of 2014 achieved a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ∼ 500 ms. SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1s. Based on the solution of Grad-Shafranov equation the shift of plasma column center from geometrical centre of vacuum chamber is measured using various magnetic probes and flux loops installed in the machine. The closed feedback loop uses plasma current (Ip), Delta R as feedback signal and manipulate the vertical field current (Ivf). The discharge starts with feed forward loop using initially provided reference then the active feedback starts after discharge by few msec once plasma column is completely formed. The feedback loop time is of the order of 10 msec. The primary objective is to acquire plasma position control related signals, compute plasma position and generate position correction signal for VF coil power supply, communicate correction to VF coil power supply and modify VF power supply output in a deterministic time span. In this we present the methodology used for plasma horizontal displacement control using vertical field and discuss the preliminary results. (author)

  16. Thermal-hydraulic and thermo-mechanical design of plasma facing components for SST-1 tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Santra, P.; Chenna Reddy, D.; Parashar, S.K.S.

    2014-01-01

    The Plasma Facing Components (PFCs) are one of the major sub-systems of ssT-1 tokamak. PFC of ssT-1 consisting of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC cooling is the steady state heat removal of up to 1 MW/m 2 . The PFC has been designed to withstand the peak heat fluxes and also without significant erosion such that frequent replacement of the armor is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to carry out the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal analysis of the PFC is carried out with the purpose of evaluating the thermal mechanical behavior of PFCs. The detailed thermal-hydraulic and thermo-mechanical designs of PFCs of ssT-1 are discussed in this paper. (authors)

  17. An overview of SST-1 diagnostics and results from recent campaigns

    International Nuclear Information System (INIS)

    Kumar, Ajai; Adhiya, Asha N.; Joshi, Hemchandra C.

    2015-01-01

    SST-1 is a large aspect ratio tokomak with superconducting magnets designed to operate in steady-state mode for around 1000 seconds. All essential diagnostics for the machine operation and advance diagnostics are commissioned in SST-1 during the different phases of its operation. This report describes the various diagnostics in SST-1 and the results of recent SST-1 campaign with Plasma Facing components. The chord averaged electron density of SST-1 plasma is recorded in the range of 2-5 x 10 12 /cc and the electron temperature is estimated around 100 eV. Various spectral line emissions from plasma and temporal evolutions of some of them have been recorded by spectroscopy diagnostics to understand the impurity behaviour in the SST-1 plasma. The radiation power loss and the power deposited on limiter has been estimated using bolometry and IR thermography respectively. Plasma evolution recorded using visible imaging diagnostics. The energy distribution of non-thermal electron has been characterised using LaBr spectrometer and NaI detector. This article will also be discussing about the possible additions and modification planned for the near future. (author)

  18. Quality control of FWC during assembly and commissioning in SST-1 Tokamak

    Science.gov (United States)

    Patel, Hitesh; Santra, Prosenjit; Parekh, Tejas; Biswas, Prabal; Jayswal, Snehal; Chauhan, Pradeep; Paravastu, Yuvakiran; George, Siju; Semwal, Pratibha; Thankey, Prashant; Ramesh, Gattu; Prakash, Arun; Dhanani, Kalpesh; Raval, D. C.; Khan, Ziauddin; Pradhan, Subrata

    2017-04-01

    First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma, comprises of limiters, divertors, baffles, passive stabilizers designed to operate long duration (∼1000 s) discharges of elongated plasma. All FWC consist of copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at inter-connected ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a rigorous quality control and checks at every stage of the assembly process. This paper will present the quality control aspects and checks of FWC from commencement of assembly procedure, namely material test reports, leak testing of high temperature baked components, assembled dimensional tolerances, leak testing of all welded joints, graphite tile tightening torques, electrical continuity and electrical isolation of passive stabilizers from vacuum vessel, baking and cooling hydraulic connections inside vacuum vessel.

  19. Engineering and thermal-hydraulic design of water cooled PFC for SST-1 tokamak

    International Nuclear Information System (INIS)

    Paritosh Chaudhuri; Santra, P.; Rabi Prakash, N.; Khirwadkar, S.; Arun Prakash, A.; Ramash, G.; Dubey, S.; Chenna Reddy, D.; Saxena, Y.C.

    2005-01-01

    Full text of publication follows: Steady state Superconducting Tokamak (SST-1) is a medium size tokamak with superconducting magnetic field coils. It is a large aspect ratio tokamak with a major radius of 1.1 m and minor radius of 0.20 m. SST-1 is designed for plasma discharge duration of ∼1000 seconds to obtain fully steady state plasma with total input power up to 1.0 MW. First Wall or Plasma Facing Components (PFC) is one or the major sub-systems of SST-1 tokamak consisting of divertors, passive stabilizers, baffles, and poloidal limiters are designed to be compatible for steady state operation. All the PFC has the same basic configuration: graphite tiles are mechanically attached to a back plate made of high strength copper alloy, and SS tubes are embedded in the groove made in the back plate. Same tube will be used for cooling during plasma operation and baking during wall conditioning. The main consideration in the design of the PFC is the steady state heat removal of up to 1 MW/m 2 . In addition to remove high heat fluxes, the PFC are also designed to be compatible for high temperature baking at 350 deg. C. Water was chosen as the coolant because of its appropriate thermal properties, and while baking, hot nitrogen gas would flow through these tubes to bake the PFC at high temperature. Extensive studies, involving different flow parameters and various cooling layouts, has been done to select the final cooling parameters and layout, compatible for cooling and baking. During steady state operation, divertor and passive stabilizer heat loads are expected to be 0.6 and 0.25 MW/m 2 . The PFC also has been design to withstand the peak heat fluxes without significant erosion such that frequent replacement is not necessary. Since the tile must be mechanically attached to the back plate (heat sink), the fitting technique must provide the highest mechanical stress so that thermal transfer efficiency is maximized. Proper brazing of cooling tube on the copper back

  20. Modeling of Eddy current distribution and equilibrium reconstruction in the SST-1 Tokamak

    International Nuclear Information System (INIS)

    Banerjee, Santanu; Sharma, Deepti; Radhakrishnana, Srinivasan; Daniel, Raju; Shankara Joisa, Y.; Atrey, Parveen Kumar; Pathak, Surya Kumar; Singh, Amit Kumar

    2015-01-01

    Toroidal continuity of the vacuum vessel and the cryostat leads to the generation of large eddy currents in these passive structures during the Ohmic phase of the steady state superconducting tokamak SST-1. This reduces the magnitude of the loop voltage seen by the plasma as also delays its buildup. During the ramping down of the Ohmic transformer current (OT), the resultant eddy currents flowing in the passive conductors play a crucial role in governing the plasma equilibrium. Amount of this eddy current and its distribution has to be accurately determined such that this can be fed to the equilibrium reconstruction code as an input. For the accurate inclusion of the effect of eddy currents in the reconstruction, the toroidally continuous conducting structures like the vacuum vessel and the cryostat with large poloidal cross-section and any other poloidal field (PF) coil sitting idle on the machine are broken up into a large number of co-axial toroidal current carrying filaments. The inductance matrix for this large set of toroidal current carrying conductors is calculated using the standard Green's function and the induced currents are evaluated for the OT waveform of each plasma discharge. Consistency of this filament model is cross-checked with the 11 in-vessel and 12 out-vessel toroidal flux loop signals in SST-1. Resistances of the filaments are adjusted to reproduce the experimental measurements of these flux loops in pure OT shots and shots with OT and vertical field (BV). Such shots are taken routinely in SST-1 without the fill gas to cross-check the consistency of the filament model. A Grad-Shafranov (GS) equation solver, named as IPREQ, has been developed in IPR to reconstruct the plasma equilibrium through searching for the best-fit current density profile. Ohmic transformer current (OT), vertical field coil current (BV), currents in the passive filaments along with the plasma pressure (p) and current (I p ) profiles are used as inputs to the IPREQ

  1. Design and performance of main vacuum pumping system of SST-1 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Ziauddin, E-mail: ziauddin@ipr.res.in; Pathan, Firozkhan; George, Siju; Dhanani, Kalpesh; Paravastu, Yuvakiran; Semwal, Pratibha; Pradhan, Subrata

    2014-01-15

    Highlights: •SST-1 Tokamak was successfully commissioned. •Vacuum vessel and cryostat were pumped down to 6.3 × 10{sup −7} mbar and 1.3 × 10{sup −5} mbar. •Leaks developed during baking were detected in-situ by RGA and confirmed later on. •Cryo-pumping effect was observed when LN2 thermal shields reached below 273 K. •Non-standard aluminum wire-seals have shown leak tightness < 1.0 × 10{sup −9} mbar l/s. -- Abstract: Steady-state Superconducting Tokamak (SST-1) was installed and it is commissioning for overall vacuum integrity, magnet systems functionality in terms of successful cool down to 4.5 K and charging up to 10 kA current was started from August 2012. Plasma operation of 100 kA current for more than 100 ms was also envisaged. It is comprised of vacuum vessel (VV) and cryostat (CST). Vacuum vessel, an ultra-high (UHV) vacuum chamber with net volume of 23 m{sup 3} was maintained at the base pressure of 6.3 × 10{sup −7} mbar for plasma confinement. Cryostat, a high-vacuum (HV) chamber with empty volume 39 m{sup 3} housing superconducting magnet system, bubble thermal shields and hydraulics for these circuits, maintained at 1.3 × 10{sup −5} mbar in order to provide suitable environment for these components. In order to achieve these ultimate vacuums, two numbers of turbo-molecular pumps (TMP) are installed in vacuum vessel while three numbers of turbo-molecular pumps are installed in cryostat. Initial pumping of both the chambers was carried out by using suitable Roots pumps. PXI based real time controlled system is used for remote operation of the complete pumping operation. In order to achieve UHV inside the vacuum vessel, it was baked at 150 °C for longer duration. Aluminum wire-seals were used for all non-circular demountable ports and a leak tightness < 1.0 × 10{sup −9} mbar l/s were achieved.

  2. Engineering and thermal-hydraulics design of PFC cooling for SST-1 Tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Reddy, D. Chenna; Santra, P.; Khiwadkar, S.; Prakash, N. Rabi; Ramash, G.; Dubey, Santosh; Prakash, Arun; Saxena, Y. C.

    2003-01-01

    The main consideration in the design of the PFC cooling for SST-1 tokamak is the steady state heat removal of upto 1MW/m2. The PFC also has been design to withstand the peak heat fluxes without significant erosion such that frequent replacement is not necessary. Proper brazing of cooling tube on the copper back plate is necessary for the efficient heat transfer from the tube to the back plate. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to conduct the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The temperature distribution results for different PFC obtained by FE results were assessed by comparison with 2-D Finite Difference code. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. The contact at the brazed joint of the tube to the backplate is critical for the above application. The manufactured modules need to be evaluated for the quality of brazed joint. Using an infra-red-camera, spatial and temporal evaluation of the temperature profile is studied under various flow parameters. These results of this study will be presented in details in this paper

  3. Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

    Science.gov (United States)

    Pathan, F. S.; Khan, Z.; Semwal, P.; Raval, D. C.; Joshi, K. S.; Thankey, P. L.; Dhanani, K. R.

    2008-05-01

    Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN2 panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN2 panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN2 panels during sniffer test and pressure drop test using helium gas.

  4. Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pathan, F S; Khan, Z; Semwal, P; Raval, D C; Joshi, K S; Thankey, P L; Dhanani, K R [Institute for Plasma Research, Bhat, Gandhinagar - 382 428, Gujarat (India)], E-mail: firose@ipr.res.in

    2008-05-01

    Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN{sub 2} panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN{sub 2} panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN{sub 2} panels during sniffer test and pressure drop test using helium gas.

  5. Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

    International Nuclear Information System (INIS)

    Pathan, F S; Khan, Z; Semwal, P; Raval, D C; Joshi, K S; Thankey, P L; Dhanani, K R

    2008-01-01

    Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN 2 panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN 2 panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN 2 panels during sniffer test and pressure drop test using helium gas

  6. Cryogenic system of steady state superconducting Tokamak SST-1: Operational experience and controls

    International Nuclear Information System (INIS)

    Sarkar, B.; Tank, Jignesh; Panchal, Pradip; Sahu, A.K.; Bhattacharya, Ritendra; Phadke, Gaurang; Gupta, N.C.; Gupta, Girish; Shah, Nitin; Shukla, Pawan; Singh, Manoj; Sonara, Dasarath; Sharma, Rajiv; Saradha, S.; Patel, J.C.; Saxena, Y.C.

    2006-01-01

    The cryogenic system of SST-1 consists of the helium cryogenic system and the nitrogen cryogenic system. The main components of the helium cryogenic system are (a) 1.3 kW helium refrigerator/liquefier (HRL) and (b) warm gas management system (WGM), where as, the nitrogen cryogenic system called as liquid nitrogen (LN 2 ) management system consists of storage tanks and a distribution system. The helium flow distribution and control to different sub-systems is achieved by the integrated flow distribution and control (IFDC) system. The HRL has been commissioned and operated for performing a single toroidal field coil test as well as for the first commissioning of SST-1 superconducting-magnets up to 68 K. Analysis of the results shows that the compressor and turbine parameters of the HRL, namely, the speed and pressure are very stable during operation of the HRL, confirming to the reliability in control of thermo-dynamic parameters of the system. The thermal shield of the SST-1 cryostat consists of ten different types of panels, which have been cooled down to the minimum temperature of 80 K and maintained during the first commissioning of SST-1. The operation and controls of the LN2 management system have been found to be as per the design consideration

  7. FPGA based phase detection technique for electron density measurement in SST-1 tokamak

    International Nuclear Information System (INIS)

    Pramila; Mandaliya, Hitesh; Rajpal, Rachana; Kaur, Rajwinder

    2016-01-01

    A multi-channel signal-conditioning and phase-detection concept is implemented in the prototype design using the high-precision OPAMP, high-speed comparators, high Q filters, high-density FPGA (Field Programmable Gate array), 10 MHz parallel-multiplying DACs (Digital to Analog Converter), etc. The complete digital-logic for the phase-detection is implemented inside the logic cells of FPGA using VHDL code, with high speed 100 MHz clock generated from Digital Clock Manager (DCM), which is used to measure the time elapsed between zero crossings of the two signals coming from reference and probe paths of the diagnostics. The logic is implemented to measure either leading or lagging phase and also to accumulate the total phase difference throughout the shot duration with the maximum value of accumulated phase of 5760 (16 cycles × 360°) degree and a resolution of 3.6 °. A precision high speed and high bandwidth (80 MHz) operational amplifiers are used as the front end-electronics component for conditioning the high-frequency (1 MHz) and low amplitude signal (μV). The hardware detail, implementation concept in FPGA and testing results will be presented in the paper.

  8. FPGA based phase detection technique for electron density measurement in SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pramila, E-mail: pramila@ipr.res.in; Mandaliya, Hitesh; Rajpal, Rachana; Kaur, Rajwinder

    2016-11-15

    A multi-channel signal-conditioning and phase-detection concept is implemented in the prototype design using the high-precision OPAMP, high-speed comparators, high Q filters, high-density FPGA (Field Programmable Gate array), 10 MHz parallel-multiplying DACs (Digital to Analog Converter), etc. The complete digital-logic for the phase-detection is implemented inside the logic cells of FPGA using VHDL code, with high speed 100 MHz clock generated from Digital Clock Manager (DCM), which is used to measure the time elapsed between zero crossings of the two signals coming from reference and probe paths of the diagnostics. The logic is implemented to measure either leading or lagging phase and also to accumulate the total phase difference throughout the shot duration with the maximum value of accumulated phase of 5760 (16 cycles × 360°) degree and a resolution of 3.6 °. A precision high speed and high bandwidth (80 MHz) operational amplifiers are used as the front end-electronics component for conditioning the high-frequency (1 MHz) and low amplitude signal (μV). The hardware detail, implementation concept in FPGA and testing results will be presented in the paper.

  9. The first experiments in SST-1

    Science.gov (United States)

    Pradhan, S.; Khan, Z.; Tanna, V. L.; Sharma, A. N.; Doshi, K. J.; Prasad, U.; Masand, H.; Kumar, Aveg; Patel, K. B.; Bhandarkar, M. K.; Dhongde, J. R.; Shukla, B. K.; Mansuri, I. A.; Varadarajulu, A.; Khristi, Y. S.; Biswas, P.; Gupta, C. N.; Sharma, D. K.; Raval, D. C.; Srinivasan, R.; Pandya, S. P.; Atrey, P. K.; Sharma, P. K.; Patel, P. J.; Patel, H. S.; Santra, P.; Parekh, T. J.; Dhanani, K. R.; Paravastu, Y.; Pathan, F. S.; Chauhan, P. K.; Khan, M. S.; Tank, J. K.; Panchal, P. N.; Panchal, R. N.; Patel, R. J.; George, S.; Semwal, P.; Gupta, P.; Mahesuriya, G. I.; Sonara, D. P.; Jayswal, S. P.; Sharma, M.; Patel, J. C.; Varmora, P. P.; Patel, D. J.; Srikanth, G. L. N.; Christian, D. R.; Garg, A.; Bairagi, N.; Babu, G. R.; Panchal, A. G.; Vora, M. M.; Singh, A. K.; Sharma, R.; Raju, D.; Kulkarni, S. V.; Kumar, M.; Manchanda, R.; Joisa, S.; Tahiliani, K.; Pathak, S. K.; Patel, K. M.; Nimavat, H. D.; Shah, P. R.; Chudasma, H. H.; Raval, T. Y.; Sharma, A. L.; Ojha, A.; Parghi, B. R.; Banaudha, M.; Makwana, A. R.; Chowdhuri, M. B.; Ramaiya, N.; kumar, A.; Raval, J. V.; Gupta, S.; Purohit, S.; Kaur, R.; Adhiya, A. N.; Jha, R.; Kumar, S.; Nagora, U. C.; Siju, V.; Thomas, J.; Chaudhari, V. R.; Patel, K. G.; Ambulkar, K. K.; Dalakoti, S.; Virani, C. G.; Parmar, P. R.; Thakur, A. L.; Das, A.; Bora, D.; the SST-1 Team

    2015-10-01

    A steady state superconducting tokamak (SST-1) has been commissioned after the successful experimental and engineering validations of its critical sub-systems. During the ‘engineering validation phase’ of SST-1; the cryostat was demonstrated to be leak-tight in all operational scenarios, 80 K thermal shields were demonstrated to be uniformly cooled without regions of ‘thermal runaway and hot spots’, the superconducting toroidal field magnets were demonstrated to be cooled to their nominal operational conditions and charged up to 1.5 T of the field at the major radius. The engineering validations further demonstrated the assembled SST-1 machine shell to be a graded, stress-strain optimized and distributed thermo-mechanical device, apart from the integrated vacuum vessel being validated to be UHV compatible etc. Subsequently, ‘field error components’ in SST-1 were measured to be acceptable towards plasma discharges. A successful breakdown in SST-1 was obtained in SST-1 in June 2013 assisted with electron cyclotron pre-ionization in the second harmonic mode, thus marking the ‘first plasma’ in SST-1 and the arrival of SST-1 into the league of contemporary steady state devices. Subsequent to the first plasma, successful repeatable plasma start-ups with E ˜ 0.4 V m-1, and plasma current in excess of 70 kA for 400 ms assisted with electron cyclotron heating pre-ionization at a field of 1.5 T have so far been achieved in SST-1. Lengthening the plasma pulse duration with lower hybrid current drive, confinement and transport in SST-1 plasmas and magnetohydrodynamic activities typical to large aspect ratio SST-1 discharges are presently being investigated in SST-1. In parallel, SST-1 has uniquely demonstrated reliable cryo-stable high field operation of superconducting TF magnets in the two-phase cooling mode, operation of vapour-cooled current leads with cold gas instead of liquid helium and an order less dc joint resistance in superconducting magnet winding

  10. Design of the 3.7 GHz, 500 kW CW circulator for the LHCD system of the SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Dixit, Harish V., E-mail: hvdixit48@yahoo.com [Veermata Jijabai Technological Institute, Mumbai, Maharashtra 400019 (India); Jadhav, Aviraj R. [Veermata Jijabai Technological Institute, Mumbai, Maharashtra 400019 (India); Jain, Yogesh M. [Institute for Plasma Research, Gandhinagar, Gujarat 382428 (India); Homi Bhabha National Institute, Training School Complex, Anushakti Nagar, Mumbai 400094 (India); Cheeran, Alice N. [Veermata Jijabai Technological Institute, Mumbai, Maharashtra 400019 (India); Gupta, Vikas [Vidyavardhini' s College of Engineering and Technology, Vasai, Maharashtra 401202 (India); Sharma, P.K. [Institute for Plasma Research, Gandhinagar, Gujarat 382428 (India); Homi Bhabha National Institute, Training School Complex, Anushakti Nagar, Mumbai 400094 (India)

    2017-06-15

    Highlights: • Design of a 500 kW CW circulator for LHCD system at 3.7 GHz. • Mechanism for thermal management of ferrite tile. • Scheme for uniform magnetisation of the ferrite tiles. • Design of high CW power CW quadrature and 180 ° hybrid coupler. - Abstract: Circulators are used in high power microwave systems to protect the vacuum source against reflection. The Lower Hybrid Current Drive (LHCD) system of SST-1 tokamak commissioned at IPR, Gandhinagar in India comprises of four high power circulators to protect klystrons (supplying 500 kW CW each at 3.7 GHz) which power the system. This paper presents the design of a Differential Phase Shift Circulator (DPSC) capable of handling 500 kW CW power at 3.7 GHz so that four circulators can be used to protect the four available klystrons. As the DPSC is composed by three main components, viz., magic tee, ferrite phase shifter and 3 dB hybrid coupler, the designing of each of the proposed components is described. The design of these components is carried out factoring various multiphysics aspects of RF, heating due to high CW power and magnetic field requirement of the ferrite phase shifter. The primary objective of this paper is to present the complete RF, magnetic and thermal design of a high CW power circulator. All the simulations have been carried out in COMSOL Multiphysics. The designed circulator exhibits an insertion loss of 0.13 dB with a worst case VSWR of 1.08:1. The total length of the circulator is 3 m.

  11. Plasma diagnostics on large tokamaks

    International Nuclear Information System (INIS)

    Orlinskij, D.V.; Magyar, G.

    1988-01-01

    The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs

  12. Initial results in SST-1 after up-gradation

    Science.gov (United States)

    Pradhan, S.; Khan, Z.; Tanna, V. L.; Prasad, U.; Paravastu, Y.; Raval, D. C.; Masand, H.; Kumar, Aveg; Dhongde, J. R.; Jana, S.; Kakati, B.; Patel, K. B.; Bhandarkar, M. K.; Shukla, B. K.; Ghosh, D.; Patel, H. S.; Parekh, T. J.; Mansuri, I. A.; Dhanani, K. R.; Varadharajulu, A.; Khristi, Y. S.; Biswas, P.; Gupta, C. N.; George, S.; Semwal, P.; Sharma, D. K.; Gulati, H. K.; Mahajan, K.; Praghi, B. R.; Banaudha, M.; Makwana, A. R.; Chudasma, H. H.; Kumar, M.; Manchanda, R.; Joisa, Y. S.; Asudani, K.; Pandya, S. N.; Pathak, S. K.; Banerjee, S.; Patel, P. J.; Santra, P.; Pathan, F. S.; Chauhan, P. K.; Khan, M. S.; Thankey, P. L.; Prakash, A.; Panchal, P. N.; Panchal, R. N.; Patel, R. J.; Mahsuria, G. I.; Sonara, D. P.; Patel, K. M.; Jayaswal, S. P.; Sharma, M.; Patel, J. C.; Varmora, P.; Srikanth, G. L. N.; Christian, D. R.; Garg, A.; Bairagi, N.; Babu, G. R.; Panchal, A. G.; Vora, M. M.; Singh, A. K.; Sharma, R.; Nimavat, H. D.; Shah, P. R.; Purwar, G.; Raval, T. Y.; Sharma, A. L.; Ojha, A.; Kumar, S.; Ramaiya, N. K.; Siju, V.; Gopalakrishna, M. V.; Kumar, A.; Sharma, P. K.; Atrey, P. K.; Kulkarni, SV; Ambulkar, K. K.; Parmar, P. R.; Thakur, A. L.; Raval, J. V.; Purohit, S.; Mishra, P. K.; Adhiya, A. N.; Nagora, U. C.; Thomas, J.; Chaudhari, V. K.; Patel, K. G.; Dalakoti, S.; Virani, C. G.; Gupta, S.; Kumar, Ajay; Chaudhari, B.; Kaur, R.; Srinivasan, R.; Raju, D.; Kanabar, D. H.; Jha, R.; Das, A.; Bora, D.

    2017-04-01

    SST-1 Tokamak has recently completed the 1st phase of up-gradation with successful installation and integration of all its First Wall components. The First Wall of SST-1 comprises of ∼ 3800 high heat flux compatible graphite tiles being assembled and installed on 132 CuCrZr heat sink back plates engraved with ∼ 4 km of leak tight baking and cooling channels in five major sub groups equipped with ∼ 400 sensors and weighing ∼ 6000 kg in total in thirteen isolated galvanic and six isolated hydraulic circuits. The phase-1 up-gradation spectrum also includes addition of Supersonic Molecular Beam Injection (SMBI) both on the in-board and out-board side, installation of fast reciprocating probes, adding some edge plasma probe diagnostics in the SOL region, installation and integration of segmented and up-down symmetric radial coils aiding/controlling plasma rotations, introduction of plasma position feedback and density controls etc. Post phase-I up-gradation spanning from Nov 2014 till June 2016, initial plasma experiments in up-graded SST-1 have begun since Aug 2016 after a brief engineering validation period in SST-1. The first experiments in SST-1 have revealed interesting aspects on the ‘eddy currents in the First Wall support structures’ influencing the ‘magnetic Null evolution dynamics’ and the subsequent plasma start-up characteristics after the ECH pre-ionization, the influence of the first walls on the ‘field errors’ and the resulting locked modes observed, the magnetic index influencing the evolution of the equilibrium of the plasma column, low density supra-thermal electron induced discharges and normal ohmic discharges etc. Presently; repeatable ohmic discharges regimes in SST-1 having plasma currents in excess of 65 KA (qa ∼ 3.8, BT = 1.5 T) with a current ramp rates ∼ 1.2 MA/s over a duration of ∼ 300 ms with line averaged densities ∼ 0.8 × 1019 and temperatures ∼ 200 eV with copious MHD signatures have been experimentally

  13. Gas fueling system for SST-1

    International Nuclear Information System (INIS)

    Dhanani, Kalpeshkumar R.; Khan, Ziauddin; Raval, Dilip; Semwal, Pratibha; George, Siju; Paravastu, Yuvakiran; Thankey, Prashant; Khan, Mohammad Shoaib; Pradhan, Subrata

    2015-01-01

    SST-1 Tokamak, the first Indian Steady-state Superconducting experimental device is at present under operation in Institute for Plasma Research. For plasma break down and initiation, the piezoelectric valve based gas feed system is implemented as primary requirement due to its precise control, easy handling, low costs for both construction and maintenance and its flexibility in working gas selection. The main functions of SST-1 gas feed system are to feed the required amount of ultrahigh purity hydrogen gas for specified period into the vessel during plasma operation and ultrahigh helium gas for glow discharge cleaning. In addition to these facilities, the gas feed system is used to feed a mixture gas of hydrogen and helium as well as other gases like nitrogen and Argon during divertor cooling etc. The piezoelectric valves used in SST-1 are remotely driven by a PXI based platform and are calibrated before the plasma operation during each SST-1 plasma operation with precise control. This paper will present the technical development and the results of gas fueling in SST-1. (author)

  14. Quench detection, protection and simulation studies on SST-1 magnets

    International Nuclear Information System (INIS)

    Sharma, Aashoo N.; Khristi, Yohan; Pradhan, Subrata; Doshi, Kalpesh; Prasad, Upendra; Banaudha, Moni; Varmora, Pankaj; Praghi, Bhadresh R.

    2015-01-01

    Steady-state Superconducting Tokamak-1 (SST-1) is India's first tokamak with superconducting toroidal field (TF) and Poloidal Field (PF) magnets. These magnets are made with NbTi based Cable-In-Conduit-Conductors. The quench characteristic of SST-1 CICC has been extensively studied both analytically and using simulation codes. Dedicated experiments like model coil test program, TF coil test program and laboratory experiments were conducted to fully characterize the performance of the CICC and the magnets made using this CICC. Results of quench experiments performed during these tests have been used to design the SST-1 quench detection and protection system. Simulation results of TF coil quenches and slow propagation quench of TF busbars have been used to further optimize these systems during the SST-1 tokamak operation. Redundant hydraulic based quench detection is also proposed for the TF coil quench detection. This paper will give the overview of these development and simulation activities. (author)

  15. MHD mode evolutions prior to minor and major disruptions in SST-1 plasma

    Energy Technology Data Exchange (ETDEWEB)

    Dhongde, Jasraj; Pradhan, Subrata, E-mail: pradhan@ipr.res.in; Bhandarkar, Manisha

    2017-01-15

    Highlights: • Observation of different regimes of MHD phenomena in SST-1 plasma. • MHD mode (m/n = 1/1, m/n = 2/1) evolutions prior to minor and major disruptions in SST-1 plasma. • MHD mode characteristics such as mode frequency, mode number, island width etc. in different regimes. - Abstract: Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (R{sub 0} = 1.1 m, a = 0.2 m, B{sub T} = 1.5T, Ip ∼ 110 kA) in operation at the Institute for Plasma Research, India. SST-1 uniquely experiments large aspect ratio (∼5.5) plasma in different operation regimes. In these experiments, repeatable characteristic MHD phenomena have been consistently observed. As the large aspect ratio plasma pulse progresses, these MHD phenomena display minor-major disruptions ably indicated in Mirnov oscillations, Mirnov oscillations with saw teeth and locked modes etc. Even though somewhat similar observations have been found in some other machines, these observations are found for the first time in large aspect ratio plasma of SST-1. This paper elaborates the magnetic field perturbations and mode evolutions due to MHD activities from Mirnov coils (poloidal and toroidal), Soft X-ray diagnostics, ECE diagnostics etc. This work further, for the first time reports quantitatively different regimes of MHD phenomena observed in SST-1 plasma, their details of mode evolutions characteristics as well as the subsequently observed minor, major disruptions supported with the physical explanations. This study will help developing disruption mitigation and avoidance scenarios for having better confinement plasma experiments.

  16. Preliminary analysis of accident in SST-1 current feeder system

    International Nuclear Information System (INIS)

    Roy, Swati; Kanabar, Deven; Garg, Atul; Singh, Amit; Tanna, Vipul; Prasad, Upendra; Srinivasan, R.

    2017-01-01

    Steady-state Tokamak-1 (SST-1) has 16 superconducting Toroidal field (TF) and 9 superconducting poloidal field (PF) coils rated for 10kA DC. All the TF are connected in series and are operated in DC condition whereas PF coils are individually operated in pulse mode during SST-1 campaigns. SST-1 current feeder system (CFS) houses 9 pairs of PF current leads and 1 pair of TF current leads. During past SST-1 campaign, there were arcing incidents within SST-1 CFS chamber which caused significant damage to PF superconducting current leads as well as its Helium cooling lines of the current leads. This paper brings out the preliminary analysis of the mentioned arcing incident, possible reasons and its investigation thereby laying out the sequence of events. From this analysis and observations, various measures to avoid such arcing incidents have also been proposed. (author)

  17. Diagnostics for the Rijnhuizen Tokamak Project

    NARCIS (Netherlands)

    Donne, A. J. H.

    1994-01-01

    The research program of the Rijnhuizen Tokamak Project is concentrated on the study of plasma transport processes. The RTP tokamak is therefore equipped with an extensive set of multichannel diagnostics, including a 19-channel FIR interferometer, a 20-channel heterodyne ECE system, an 80-channel

  18. Ion diagnostics for the tokamak boundary

    International Nuclear Information System (INIS)

    Matthews, G.F.

    1991-01-01

    In this paper, recent developments in ion diagnostic probes for tokamak boundary plasmas are discussed. Three areas are covered: retarding field analysers, sniffer probes and plasma ion mass spectrometers. The contribution of these diagnostics to our understanding of plasma surface interactions is summarised. (author)

  19. Quench detection electronics testing protocol for SST-1 magnets

    International Nuclear Information System (INIS)

    Banaudha, Moni; Varmora, Pankaj; Parghi, Bhadresh; Prasad, Upendra

    2017-01-01

    Quench Detection (QD) system consisting 204 signal channels has been successfully installed and working well during plasma experiment of SST-1 Tokamak. QD system requires testing, validation and maintenance in every SST-1 campaign for better reliability and maintainability of the system. Standalone test of each channel of the system is essential for hard-ware validation. The standard Testing Protocol follow in every campaign which validate each section of QD electronics as well as voltage tap signal cables which are routed inside the cryostat and then extended outside of the SST-1 machine up-to the magnet control room. Fiber link for Quench signal transmission to the SST-1 magnet power supply is also test and validate before every plasma campaign. Precise instrument used as a dummy source of quench signal and for manual quench generation to test the each channel and Master Quench Logic. Each signal Integrated with the magnet DAQ system, signal observed at 1Hz and 50Hz configuration to validate the logging data, compare with actual and previous test data. This paper describes the testing protocol follow in every campaign to validate functionality of QD electronics, limitation of testing, test results and overall integration of the quench detection system for SST-1 magnet. (author)

  20. Tokamak physics experiment: Diagnostic windows study

    International Nuclear Information System (INIS)

    Merrigan, M.; Wurden, G.A.

    1995-11-01

    We detail the study of diagnostic windows and window thermal stress remediation in the long-pulse, high-power Tokamak Physics Experiment (TPX) operation. The operating environment of the TPX diagnostic windows is reviewed, thermal loads on the windows estimated, and cooling requirements for the windows considered. Applicable window-cooling technology from other fields is reviewed and its application to the TPX windows considered. Methods for TPX window thermal conditioning are recommended, with some discussion of potential implementation problems provided. Recommendations for further research and development work to ensure performance of windows in the TPX system are presented

  1. Enhancing detection sensitivity of SST-1 Thomson scattering experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhari, Vishnu; Patel, Kiran; Thomas, Jinto; Kumar, Ajai, E-mail: ajai@ipr.res.in

    2016-10-15

    Thomson Scattering System (TSS) is the main diagnostic to extract electron temperature and density of steady state superconducting (SST-1) tokamak plasma. Silicon avalanche photo diode is used with low noise and fast signal conditioning electronics (SCE) to detect incoming Thomson scattered laser photons. A stringent requirement for the measurement is to detect high speed and low level light signal (detection of 100 numbers of Thomson scattered photons for 50 ns pulse width at input of active area of detector) in the presence of wide band electro-magnetic interference (EMI) noise. The electronics and instruments for different sub-systems kept in laboratory contribute to the radiated and conductive noise in a complex manner to the experiment, which can degrade the resultant signal to noise ratio (SNR <1). In general a repeated trial method with flexible grounding scheme are used to improve system signal to noise ratio, which is time consuming and less efficient. In the present work a simple, robust, cost-effective instrumentation system is used for the measurement and monitoring with improved ground scheme and shielding method to minimize noise, isolating the internal sub-system generated noise and external interference which leads to an improved SNR.

  2. Overall behaviour of PFC integrated SST-1 vacuum system

    Science.gov (United States)

    Khan, Ziauddin; Raval, Dilip C.; Paravasu, Yuvakiran; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; George, Siju; Shoaib, Mohammad; Prakash, Arun; Babu, Gattu R.; Thankey, Prashant; Pathan, Firozkhan S.; Pradhan, Subrata

    2017-04-01

    As a part of phase-I up-gradation of Steady-state Superconducting Tokamak (SST-1), Graphite Plasma Facing Components (PFCs) have been integrated inside SST-1 vacuum vessel as a first wall (FW) during Nov 14 and May 2015. The SST-1 FW has a total surface area of the installed PFCs exposed to plasma is ∼ 40 m2 which is nearly 50% of the total surface area of stainless steel vacuum chamber (∼75 m2). The volume of the vessel within the PFCs is ∼ 16 m3. After the integration of PFCs, the entire vessel as well as the PFC cooling/baking circuits has been qualified with an integrated helium leak tightness of baked at 250 °C for nearly 20 hours employing hot nitrogen gas to remove the absorbed water vapours. Thereafter, Helium glow discharges cleaning were carried out towards the removal of surface impurities. The pump down characteristics of SST-1 vacuum chamber and the changes in the residual gaseous impurities after the installation of the PFCs will be discussed in this paper.

  3. Quench characterization and thermo hydraulic analysis of SST-1 TF magnet busbar

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, A.N., E-mail: ansharma@ipr.res.in [Institute for Plasma Research, Gandhinagar (India); Pradhan, S. [Institute for Plasma Research, Gandhinagar (India); Duchateau, J.L. [CEA Cadarache, 13108 St Paul lez Durance Cedex (France); Khristi, Y.; Prasad, U.; Doshi, K.; Varmora, P.; Tanna, V.L.; Patel, D.; Panchal, A. [Institute for Plasma Research, Gandhinagar (India)

    2015-01-15

    Highlights: • Details of SST-1 TF busbar quench detection. • Simulation of slow propagating normal zone. • Thermo hydraulic analyses of TF busbar in current feeder system. - Abstract: Toroidal field (TF) magnet system of steady-state superconducting tokamak-1 (SST-1) has 16 superconducting coils. TF coils are cooled with forced flow supercritical helium at 0.4 MPa, at 4.5 K and operate at nominal current of 10,000 A. Prior to TF magnet system assembly in SST-1 tokamak, each TF coil was tested individually in a test cryostat. During these tests, TF coil was connected to a pair of conventional helium vapor cooled current leads. The connecting busbar was made from the same base cable-in-conduit-conductor (CICC) of SST-1 superconducting magnet system. Quenches experimentally observed in the busbar sections of the single coil test setups have been analyzed in this paper. A steady state thermo hydraulic analysis of TF magnet busbar in actual SST-1 tokamak assembly has been done. The experimental observations of quench and results of relevant thermo hydraulic analyses have been used to predict the safe operation regime of TF magnet system busbar during actual SST-1 tokamak operational scenarios.

  4. Magnetic diagnostics for the proto-eta Tokamak

    International Nuclear Information System (INIS)

    Ferreira, J.L.; Aso, Y.; Ueda, M.; Ferreira, J.G.

    1991-04-01

    This work gives a general view of the magnetic diagnostics rat will be used in the Proto-Eta Tokamak. These diagnostics will be useful tools to measure currents, electric and magnetic fields involved in the plasma magnetic confinement. (author)

  5. Baking of SST-1 vacuum vessel modules and sectors

    International Nuclear Information System (INIS)

    Pathan, Firozkhan S; Khan, Ziauddin; Yuvakiran, Paravastu; George, Siju; Ramesh, Gattu; Manthena, Himabindu; Shah, Virendrakumar; Raval, Dilip C; Thankey, Prashant L; Dhanani, Kalpesh R; Pradhan, Subrata

    2012-01-01

    SST-1 Tokamak is a steady state super-conducting tokamak for plasma discharge of 1000 sec duration. The plasma discharge of such long time duration can be obtained by reducing the impurities level, which will be possible only when SST-1 vacuum chamber is pumped to ultra high vacuum. In order to achieve UHV inside the chamber, the baking of complete vacuum chamber has to be carried out during pumping. For this purpose the C-channels are welded inside the vacuum vessel. During baking of vacuum vessel, these welded channels should be helium leak tight. Further, these U-channels will be in accessible under operational condition of SST-1. So, it will not possible to repair if any leak is developed during experiment. To avoid such circumstances, a dedicated high vacuum chamber is used for baking of the individual vacuum modules and sectors before assembly so that any fault during welding of the channels will be obtained and repaired. This paper represents the baking of vacuum vessel modules and sectors and their temperature distribution along the entire surface before assembly.

  6. Plasma diagnostics for the compact ignition tokamak

    International Nuclear Information System (INIS)

    Medley, S.S.; Young, K.M.

    1988-06-01

    The primary mission of the Compact Ignition Tokamak (CIT) is to study the physics of alpha-particle heating in an ignited D-T plasma. A burn time of about 10 /tau//sub E/ is projected in a divertor configuration with baseline machine design parameters of R=2.10 m, 1=0.65 m, b=1.30 m, I/sub p/=11 MA, B/sub T/=10 T and 10-20 MW of auxiliary rf heating. Plasma temperatures and density are expected to reach T/sub e/(O) /approximately/20 keV, T/sub i/(O) /approximately/30 keV, and n/sub e/(O) /approximately/ 1 /times/ 10 21 m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs

  7. Progress in diagnostics of the COMPASS tokamak

    Science.gov (United States)

    Weinzettl, V.; Adamek, J.; Berta, M.; Bilkova, P.; Bogar, O.; Bohm, P.; Cavalier, J.; Dejarnac, R.; Dimitrova, M.; Ficker, O.; Fridrich, D.; Grover, O.; Hacek, P.; Havlicek, J.; Havranek, A.; Horacek, J.; Hron, M.; Imrisek, M.; Komm, M.; Kovarik, K.; Krbec, J.; Markovic, T.; Matveeva, E.; Mitosinkova, K.; Mlynar, J.; Naydenkova, D.; Panek, R.; Paprok, R.; Peterka, M.; Podolnik, A.; Seidl, J.; Sos, M.; Stockel, J.; Tomes, M.; Varavin, M.; Varju, J.; Vlainic, M.; Vondracek, P.; Zajac, J.; Zacek, F.; Stano, M.; Anda, G.; Dunai, D.; Krizsanoczi, T.; Refy, D.; Zoletnik, S.; Silva, A.; Gomes, R.; Pereira, T.; Popov, Tsv.; Sarychev, D.; Ermak, G. P.; Zebrowski, J.; Jakubowski, M.; Rabinski, M.; Malinowski, K.; Nanobashvili, S.; Spolaore, M.; Vianello, N.; Gauthier, E.; Gunn, J. P.; Devitre, A.

    2017-12-01

    The COMPASS tokamak at IPP Prague is a small-size device with an ITER-relevant plasma geometry and operating in both the Ohmic as well as neutral beam assisted H-modes since 2012. A basic set of diagnostics installed at the beginning of the COMPASS operation has been gradually broadened in type of diagnostics, extended in number of detectors and collected channels and improved by an increased data acquisition speed. In recent years, a significant progress in diagnostic development has been motivated by the improved COMPASS plasma performance and broadening of its scientific programme (L-H transition and pedestal scaling studies, magnetic perturbations, runaway electron control and mitigation, plasma-surface interaction and corresponding heat fluxes, Alfvenic and edge localized mode observations, disruptions, etc.). In this contribution, we describe major upgrades of a broad spectrum of the COMPASS diagnostics and discuss their potential for physical studies. In particular, scrape-off layer plasma diagnostics will be represented by a new concept for microsecond electron temperature and heat flux measurements - we introduce a new set of divertor Langmuir and ball-pen probe arrays, newly constructed probe heads for reciprocating manipulators as well as several types of standalone probes. Among optical tools, an upgraded high-resolution edge Thomson scattering diagnostic for pedestal studies and a set of new visible light and infrared (plasma-surface interaction investigations) cameras will be described. Particle and beam diagnostics will be covered by a neutral particle analyzer, diagnostics on a lithium beam, Cherenkov detectors (for a direct detection of runaway electrons) and neutron detectors. We also present new modifications of the microwave reflectometer for fast edge density profile measurements.

  8. Electronics and instrumentation for the SST-1 superconducting magnet system

    International Nuclear Information System (INIS)

    Khristi, Yohan; Pradhan, Subrata; Varmora, Pankaj; Banaudha, Moni; Praghi, Bhadresh R.; Prasad, Upendra

    2015-01-01

    Steady State Superconducting Tokamak-1 (SST-1) at Institute for Plasma Research (IPR), India is now in operation phase. The SST-1 magnet system consists of sixteen superconducting (SC), D-shaped Toroidal Field (TF) coils and nine superconducting Poloidal Field (PF) coils together with a pair of resistive PF coils, inside the vacuum vessel of SST-1. The magnets were cooled down to 4.5 K using either supercritical or two-phase helium, after which they were charged up to 10 kA of transport current. Precise quench detection system, cryogenic temperature, magnetic field, strain, displacement, flow and pressure measurements in the Superconducting (SC) magnet were mandatory. The Quench detection electronics required to protect the SC magnets from the magnet Quench therefore system must be reliable and prompt to detect the quench from the harsh tokamak environment and high magnetic field interference. A ∼200 channels of the quench detection system for the TF magnet are working satisfactorily with its design criteria. Over ∼150 channels Temperature measurement system was implemented for the several locations in the magnet and hydraulic circuits with required accuracy of 0.1K at bellow 30K cryogenic temperature. Whereas the field, strain and displacement measurements were carried out at few predefined locations on the magnet. More than 55 channels of Flow and pressure measurements are carried out to know the cooling condition and the mass flow of the liquid helium (LHe) coolant for the SC Magnet system. This report identifies the different in-house modular signal conditioning electronics and instrumentation systems, calibration at different levels and the outcomes for the SST-1 TF magnet system. (author)

  9. Filterscope diagnostic system on EAST tokamak

    International Nuclear Information System (INIS)

    Xu, Z.; Wu, Z.W.; Gao, W.; Zhang, L.; Huang, J.; Chen, Y.J.; Wu, C.R.; Zhang, P.F.

    2015-01-01

    Filterscope diagnostic system, which is designed for monitoring the line emission in fusion plasma has been widely used on fusion devices such as DIII-D, NSTX, CDX-U, KSTAR etc. On EAST (Experimental Advanced Superconducting Tokamak), a filterscope diagnostic system has been mounted to observe the line emission and visible bremsstrahlung emission in plasma from discharge campaign of 2014. It plays a crucial role in studying Edge Localized Modes (ELM) and H-mode, thanks to its high temporal resolution (0.005ms) and good spatial resolution (∼2cm). Furthermore, multi-channel signals at up to 200kHz sampling rates can be digitized simultaneously. The wavelength covers He II (468.5nm), Li I (670.8nm), Li II (548.3nm), C III (465.0nm), O II (441.5nm), Mo I (386.4nm), W I (400.9nm) and visible bremsstrahlung radiation at 538nm besides Dα (656.1nm) and Dγ (433.9nm) with the corresponding wavelength filters. The new developed filterscope system was operating during the EAST 2014 fall experimental campaign and several types ELMs has been observed. (author)

  10. Diagnostics systems for the TBR-E tokamak

    International Nuclear Information System (INIS)

    Ueda, M.; Ferreira, J.L.; Aso, Y.; Ferreira, J.G.

    1992-08-01

    A general view of the several diagnostics systems proposed for the TBR-E tokamak is given. This project is a joint undertaking of INPE, USP and UNICAMP plasma laboratories. The requirements for the measurements of the plasma produced parameters are described. Special attention is given for diagnostics used to investigate new physical issues on a low aspect ratio tokamak such as TBR-E. (author)

  11. Archiving and retrieval of experimental data using SAN based centralized storage system for SST-1

    Energy Technology Data Exchange (ETDEWEB)

    Bhandarkar, Manisha, E-mail: manisha@ipr.res.in; Masand, Harish; Kumar, Aveg; Patel, Kirit; Dhongde, Jasraj; Gulati, Hitesh; Mahajan, Kirti; Chudasama, Hitesh; Pradhan, Subrata

    2016-11-15

    Highlights: • SAN (Storage Area Network) based centralized data storage system of SST-1 has envisaged to address the need of centrally availability of SST-1 storage system to archive/retrieve experimental data for the authenticated users for 24 × 7. • The SAN based data storage system has been designed/configured with 3-tiered architecture and GFS cluster file system with multipath support. • The adopted SAN based data storage for SST-1 is a modular, robust, and allows future expandability. • Important considerations has been taken like, Handling of varied Data writing speed from different subsystems to central storage, Simultaneous read access of the bulk experimental and as well as essential diagnostic data, The life expectancy of data, How often data will be retrieved and how fast it will be needed, How much historical data should be maintained at storage. - Abstract: SAN (Storage Area Network, a high-speed, block level storage device) based centralized data storage system of SST-1 (Steady State superconducting Tokamak) has envisaged to address the need of availability of SST-1 operation & experimental data centrally for archival as well as retrieval [2]. Considering the initial data volume requirement, ∼10 TB (Terabytes) capacity of SAN based data storage system has configured/installed with optical fiber backbone with compatibility considerations of existing Ethernet network of SST-1. The SAN based data storage system has been designed/configured with 3-tiered architecture and GFS (Global File System) cluster file system with multipath support. Tier-1 is of ∼3 TB (frequent access and low data storage capacity) comprises of Fiber channel (FC) based hard disks for optimum throughput. Tier-2 is of ∼6 TB (less frequent access and high data storage capacity) comprises of SATA based hard disks. Tier-3 will be planned later to store offline historical data. In the SAN configuration two tightly coupled storage servers (with cluster configuration) are

  12. Archiving and retrieval of experimental data using SAN based centralized storage system for SST-1

    International Nuclear Information System (INIS)

    Bhandarkar, Manisha; Masand, Harish; Kumar, Aveg; Patel, Kirit; Dhongde, Jasraj; Gulati, Hitesh; Mahajan, Kirti; Chudasama, Hitesh; Pradhan, Subrata

    2016-01-01

    Highlights: • SAN (Storage Area Network) based centralized data storage system of SST-1 has envisaged to address the need of centrally availability of SST-1 storage system to archive/retrieve experimental data for the authenticated users for 24 × 7. • The SAN based data storage system has been designed/configured with 3-tiered architecture and GFS cluster file system with multipath support. • The adopted SAN based data storage for SST-1 is a modular, robust, and allows future expandability. • Important considerations has been taken like, Handling of varied Data writing speed from different subsystems to central storage, Simultaneous read access of the bulk experimental and as well as essential diagnostic data, The life expectancy of data, How often data will be retrieved and how fast it will be needed, How much historical data should be maintained at storage. - Abstract: SAN (Storage Area Network, a high-speed, block level storage device) based centralized data storage system of SST-1 (Steady State superconducting Tokamak) has envisaged to address the need of availability of SST-1 operation & experimental data centrally for archival as well as retrieval [2]. Considering the initial data volume requirement, ∼10 TB (Terabytes) capacity of SAN based data storage system has configured/installed with optical fiber backbone with compatibility considerations of existing Ethernet network of SST-1. The SAN based data storage system has been designed/configured with 3-tiered architecture and GFS (Global File System) cluster file system with multipath support. Tier-1 is of ∼3 TB (frequent access and low data storage capacity) comprises of Fiber channel (FC) based hard disks for optimum throughput. Tier-2 is of ∼6 TB (less frequent access and high data storage capacity) comprises of SATA based hard disks. Tier-3 will be planned later to store offline historical data. In the SAN configuration two tightly coupled storage servers (with cluster configuration) are

  13. Plasma diagnostics using synchrotron radiation in tokamaks

    International Nuclear Information System (INIS)

    Fidone, I.; Giruzzi, G.; Granata, G.

    1995-09-01

    This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs

  14. Edge plasma diagnostics on Tore Supra tokamak

    International Nuclear Information System (INIS)

    Fujita, Junji

    1991-01-01

    From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)

  15. Design of mass flow rate measurement system for SST-1 superconducting magnet system

    Energy Technology Data Exchange (ETDEWEB)

    Varmora, P., E-mail: pvamora@ipr.res.in; Sharma, A.N.; Khristi, Y.; Prasad, U.; Patel, D.; Doshi, K.; Pradhan, S.

    2016-11-15

    Highlights: • Design of Venturi meter for SST-1 magnet system. • Details of Helium mass flow measurement system used in SST-1. • Instruments and measurement techniques for flow measurement. • VME based data acquisition system details and flow calculation and results from SST-1 campaigns. - Abstract: Superconducting Magnet System (SCMS) of Steady State Superconducting Tokamak – 1 (SST-1) is forced-flow cooled by a closed cycle 1.3 kW (at 4.5 K) class Helium Refrigerator cum Liquefier (HRL) system. An accurate measurement of helium mass flow rate in different coils is required to ensure the uniform cooling of the cold mass in the entire range of operating temperature (300 K to 4.5 K) and pressure (0.9–0.4 MPa). To meet this requirement, indigenously designed and fabricated venturi meters are installed on 27 different coils of SST-1 SCMS. A VME based Data Acquisition System (DAS) has been developed and used to acquire the flow measurement data from different flowmeters. The details of the design of venturi meter, its different measurement and signal conditioning components, the data acquisition system and the mass flow rate calculation method are described in this paper. The mass flow rate measurement data from cryogenic acceptance and SST-1 magnet commissioning experiments are also presented and discussed in this paper.

  16. Design of mass flow rate measurement system for SST-1 superconducting magnet system

    International Nuclear Information System (INIS)

    Varmora, P.; Sharma, A.N.; Khristi, Y.; Prasad, U.; Patel, D.; Doshi, K.; Pradhan, S.

    2016-01-01

    Highlights: • Design of Venturi meter for SST-1 magnet system. • Details of Helium mass flow measurement system used in SST-1. • Instruments and measurement techniques for flow measurement. • VME based data acquisition system details and flow calculation and results from SST-1 campaigns. - Abstract: Superconducting Magnet System (SCMS) of Steady State Superconducting Tokamak – 1 (SST-1) is forced-flow cooled by a closed cycle 1.3 kW (at 4.5 K) class Helium Refrigerator cum Liquefier (HRL) system. An accurate measurement of helium mass flow rate in different coils is required to ensure the uniform cooling of the cold mass in the entire range of operating temperature (300 K to 4.5 K) and pressure (0.9–0.4 MPa). To meet this requirement, indigenously designed and fabricated venturi meters are installed on 27 different coils of SST-1 SCMS. A VME based Data Acquisition System (DAS) has been developed and used to acquire the flow measurement data from different flowmeters. The details of the design of venturi meter, its different measurement and signal conditioning components, the data acquisition system and the mass flow rate calculation method are described in this paper. The mass flow rate measurement data from cryogenic acceptance and SST-1 magnet commissioning experiments are also presented and discussed in this paper.

  17. Magnetic diagnostic plasma position in the TCA/BR tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Kuznetsov, Yu.K.; Nascimento, I.C.

    1996-01-01

    The cross-section of the plasma column is TCA/BR has a nearly circular plasma shape. This allows implementation of simplified methods of magnetic diagnostics. Although these methods were in may tokamaks and are well described, their accuracies are not clearly defined because the very simplified theoretical model of plasma equilibrium on which they are based differs from the real conditions in tokamaks like TCA/BR. In this paper we present the methods of plasma position diagnostics in TCA/BR from external magnetic measurements with an error analysis. (author). 4 refs., 3 figs

  18. Assembly and metrology of first wall components of SST-1

    International Nuclear Information System (INIS)

    Parekh, Tejas; Santra, Prosenjit; Biswas, Prabal

    2015-01-01

    First Wall components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring and port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 under going a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects and procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel. (author)

  19. Assembly & Metrology of First Wall Components of SST-1

    Science.gov (United States)

    Parekh, Tejas; Santra, Prosenjit; Biswas, Prabal; Patel, Hiteshkumar; Paravastu, Yuvakiran; Jaiswal, Snehal; Chauhan, Pradeep; Babu, Gattu Ramesh; A, Arun Prakash; Bhavsar, Dhaval; Raval, Dilip C.; Khan, Ziauddin; Pradhan, Subrata

    2017-04-01

    First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects & procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel.

  20. Plasma diagnostics for tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Stott, P E; Sanchez, J

    1994-07-01

    A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: Magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Scattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma. (Author) 451 refs.

  1. Plasma diagnostics for tokamaks and stellarators

    International Nuclear Information System (INIS)

    Stott, P.E.; Sanchez, J.

    1994-01-01

    A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Sattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma

  2. Edge transport and fluctuation induced turbulence characteristics in early SST-1 plasma

    Energy Technology Data Exchange (ETDEWEB)

    Kakati, B., E-mail: bharat.kakati@ipr.res.in; Pradhan, S., E-mail: pradhan@ipr.res.in; Dhongde, J.; Semwal, P.; Yohan, K.; Banaudha, M.

    2017-02-15

    Highlights: • Anomalous particle transport during the high MHD activity at SST-1. • Electrostatic turbulence is modulated by MHD activity at SST-1 tokamak. • Edge floating potential fluctuations shows poloidal long-range cross correlation. - Abstract: Plasma edge transport characteristics are known to be heavily influenced by the edge fluctuation induced turbulences. These characteristics play a critical role towards the confinement of plasma column in a Tokamak. The edge magnetic fluctuations and its subsequent effect on electrostatic fluctuations have been experimentally investigated for the first time at the edge of the SST-1 plasma column. This paper reports the correlations that exist and is experimentally been observed between the edge densities and floating potential fluctuations with the magnetic fluctuations. The edge density and floating potential fluctuations have been measured with the help of poloidally separated Langmuir probes, whereas the magnetic fluctuations have been measured with poloidally spaced Mirnov coils. Increase in magnetic fluctuations associated with enhanced MHD activities has been found to increase the floating potential and ion saturation current. These observations indicate electrostatic turbulence getting influenced with the MHD activities and reveal the edge anomalous particle transport during SST-1 tokamak discharge. Large-scale coherent structures have been observed in the floating potential fluctuations, indicating long-distance cross correlation in the poloidal directions. From bispectral analysis, a strong nonlinear coupling among the floating potential fluctuations is observed in the low-frequency range about 0–15 kHz.

  3. Design of new central solenoid for SST-1

    Science.gov (United States)

    Prasad, Upendra; Pradhan, Subrata; Ghate, Mahesh; Raj, Piyush; Tanna, V. L.; Khan, Ziauddin; Roy, Swati; Santra, Prosenjit; Biswas, Prabal; Sharma, A. N.; Khristi, Yohan; Kanaber, Deven; Varmora, Pankaj

    2017-04-01

    The key role of central solenoid (CS) magnet of a Tokamak is for gas breakdown, ramp up and maintaining of plasma current. The magnetic flux change in CS along with other PF coils generates magnetic null and induces electric field in toroidal direction. The induced toroidal electric field accelerates the residual electrons which collide with the neutrals and an avalanche takes place which led to the net plasma in the vacuum vessel of a Tokamak. In order to maximize the CS volt-sec capability, the higher magnetic field with a greater magnetic flux linkage is necessary. In order to facilitate all these requirements of SST-1 a new superconducting CS has been designed for SST-1. The design of new central solenoid has two bases; first one is physics and second is smart engineering in limited bore diameter of ∼ 655 mm. The physics basis of the design includes volt-sec storage capacity of ∼ 0.8 volt-sec, magnetic field null around 0.2 m over major radius of 1.1 m and toroidal electric field of ∼ 0.3 volt/m. The engineering design of new CS consists of Nb3Sn cable in conduit conductor (CICC) of operating current of 14 kA @ 4.5 K at 6 T, consolidated winding pack, smart quench detection system, protection system, housing cryostat and conductor terminations and joint design. The winding pack consists of 576 numbers of turns distributed in four layers with 0.75 mm FRP tape soaked with cyanide Easter based epoxy resin turn insulation and 3 mm of ground insulation. The interlayer low resistance (∼1 nΩ) terminal praying hand joints at 14 kA at 4.5 K has been designed for making winding pack continuous. The total height of winding pack is 2500 mm. The stored energy of this winding pack is ∼ 3 MJ at 14 kA of operating current. The expected heat load at cryogenic temperature is ∼ 10 W per layer, which requires helium mass flow rate of 1.4 g/s at 1.4 bars @ 4.5 K. The typical diameter and height of housing cryostat are 650 mm and 2563 mm with 80 K shield respectively

  4. Design of new central solenoid for SST-1

    International Nuclear Information System (INIS)

    Prasad, Upendra; Pradhan, Subrata; Ghate, Mahesh; Raj, Piyush; Tanna, V L; Khan, Ziauddin; Roy, Swati; Santra, Prosenjit; Biswas, Prabal; Sharma, A N; Khristi, Yohan; Kanaber, Deven; Varmora, Pankaj

    2017-01-01

    The key role of central solenoid (CS) magnet of a Tokamak is for gas breakdown, ramp up and maintaining of plasma current. The magnetic flux change in CS along with other PF coils generates magnetic null and induces electric field in toroidal direction. The induced toroidal electric field accelerates the residual electrons which collide with the neutrals and an avalanche takes place which led to the net plasma in the vacuum vessel of a Tokamak. In order to maximize the CS volt-sec capability, the higher magnetic field with a greater magnetic flux linkage is necessary. In order to facilitate all these requirements of SST-1 a new superconducting CS has been designed for SST-1. The design of new central solenoid has two bases; first one is physics and second is smart engineering in limited bore diameter of ∼ 655 mm. The physics basis of the design includes volt-sec storage capacity of ∼ 0.8 volt-sec, magnetic field null around 0.2 m over major radius of 1.1 m and toroidal electric field of ∼ 0.3 volt/m. The engineering design of new CS consists of Nb3Sn cable in conduit conductor (CICC) of operating current of 14 kA @ 4.5 K at 6 T, consolidated winding pack, smart quench detection system, protection system, housing cryostat and conductor terminations and joint design. The winding pack consists of 576 numbers of turns distributed in four layers with 0.75 mm FRP tape soaked with cyanide Easter based epoxy resin turn insulation and 3 mm of ground insulation. The interlayer low resistance (∼1 nΩ) terminal praying hand joints at 14 kA at 4.5 K has been designed for making winding pack continuous. The total height of winding pack is 2500 mm. The stored energy of this winding pack is ∼ 3 MJ at 14 kA of operating current. The expected heat load at cryogenic temperature is ∼ 10 W per layer, which requires helium mass flow rate of 1.4 g/s at 1.4 bars @ 4.5 K. The typical diameter and height of housing cryostat are 650 mm and 2563 mm with 80 K shield respectively

  5. One channel bolometric diagnostic for TBR-1 tokamak

    International Nuclear Information System (INIS)

    Silveira, R.M.; Vannucci, A.; Silva, R.P. da; Machida, M.

    1993-01-01

    One channel bolometric diagnostic has been designed and built for the TBR-1 tokamak to measure the machine radiation power loss. The detector, a thinistor, was calibrated and its characteristics experimentally determined. In this work the first results obtained with the bolometer are presented and discussed. (author)

  6. Development in Diagnostics Application to Control Advanced Tokamak Plasma

    International Nuclear Information System (INIS)

    Koide, Y.

    2008-01-01

    For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)

  7. Some theoretical problems of magnetic diagnostics in tokamaks and stellarators

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    1993-12-01

    The main problem of magnetic diagnostics is discussed here: which plasma characteristics can be determined from magnetic measurements in tokamaks and stellarators. The reasons are elucidated why diamagnetic measurements are reliable and easily interpreted. We discuss also the capabilities of diagnostics based on the measurements of poloidal fields outside the plasma. This article is based on a lecture delivered at the Third International School on Plasma Physics and Controlled Fusion, held 15-22 June 1993 at St. Petersburg - Kizhi, Russia. (author)

  8. Some theoretical problems of magnetic diagnostics in tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Pustovitov, V.D.

    1993-12-01

    The main problem of magnetic diagnostics is discussed here: which plasma characteristics can be determined from magnetic measurements in tokamaks and stellarators. The reasons are elucidated why diamagnetic measurements are reliable and easily interpreted. We discuss also the capabilities of diagnostics based on the measurements of poloidal fields outside the plasma. This article is based on a lecture delivered at the Third International School on Plasma Physics and Controlled Fusion, held 15-22 June 1993 at St. Petersburg - Kizhi, Russia. (author).

  9. Bragg rotor spectrometer for tokamak diagnostics

    International Nuclear Information System (INIS)

    Barnsley, R.; Evans, K.D.; Peacock, N.J.

    1986-01-01

    This paper discusses high thru'put broad band (1-24 angstrom) x-ray spectrometer having absolute calibration for wavelength and intensity and demonstrated on the DITE tokamak. This instrument has a self-contained vacuum system which allows full spatial scans of the DITE plasma. Data acquisition and drive mechanism for the rotor and filter selection are operated remotely from a SADA. Results are presented of fast spectral surveys and λ-lock time-evolution of impurity emission during neutral beam injection. Spatial scans of the absolute impurity concentrations are derived

  10. Upgradation in SCADA and PLC of existing LN_2 control system for SST-1

    International Nuclear Information System (INIS)

    Panchal, Pradip; Mahesuria, Gaurang; Panchal, Rohit; Patel, Rakesh; Sonara, Dashrath; Pitroda, Dipen; Nimavat, Hiren; Tanna, Vipul; Pradhan, Subrata

    2016-01-01

    Highlights: • The control system of LN_2 Management System of SST-1 is designed on PLC and SCADA. • The implementation and results of up-gradation in PLC and SCADA are reported. • The up-gradation in PLC and SCADA has improved the reliability & availability of SST-1 LN_2 system. - Abstract: Helium Refrigerator/Liquefier system of Steady State Superconducting Tokamak (SST-1) incorporates Liquid Nitrogen (LN_2) pre-cooling system. LN_2 is used for 80 K thermal shields of SST-1, current feeder system and integrated flow distribution and control system. The LN_2 management system is distributed system and requires automatic control. Initially LN_2 control system had Citect based Supervisory Control and Data Acquisition (SCADA) and Koyo make Programmable Logic Controller (PLC). With the passage of time and due to unavailability of their hardware, it is being obsoleted. So, the requirements of new PLC and SCADA systems have been envisaged to make uninterruptable operation of SST-1 cryogenic system. Therefore, Wonderware SCADA and Schneider Electric make PLC is programmed to replace Citect SCADA and Koyo PLC. New control features have been added in upgraded control system for better management of LN_2 system. This upgradation of SCADA and PLC is completed, tested successfully and in operation. Operational performance highlights of the new upgraded system are presented in this paper.

  11. Upgradation in SCADA and PLC of existing LN{sub 2} control system for SST-1

    Energy Technology Data Exchange (ETDEWEB)

    Panchal, Pradip, E-mail: pradip@ipr.res.in; Mahesuria, Gaurang; Panchal, Rohit; Patel, Rakesh; Sonara, Dashrath; Pitroda, Dipen; Nimavat, Hiren; Tanna, Vipul; Pradhan, Subrata

    2016-11-15

    Highlights: • The control system of LN{sub 2} Management System of SST-1 is designed on PLC and SCADA. • The implementation and results of up-gradation in PLC and SCADA are reported. • The up-gradation in PLC and SCADA has improved the reliability & availability of SST-1 LN{sub 2} system. - Abstract: Helium Refrigerator/Liquefier system of Steady State Superconducting Tokamak (SST-1) incorporates Liquid Nitrogen (LN{sub 2}) pre-cooling system. LN{sub 2} is used for 80 K thermal shields of SST-1, current feeder system and integrated flow distribution and control system. The LN{sub 2} management system is distributed system and requires automatic control. Initially LN{sub 2} control system had Citect based Supervisory Control and Data Acquisition (SCADA) and Koyo make Programmable Logic Controller (PLC). With the passage of time and due to unavailability of their hardware, it is being obsoleted. So, the requirements of new PLC and SCADA systems have been envisaged to make uninterruptable operation of SST-1 cryogenic system. Therefore, Wonderware SCADA and Schneider Electric make PLC is programmed to replace Citect SCADA and Koyo PLC. New control features have been added in upgraded control system for better management of LN{sub 2} system. This upgradation of SCADA and PLC is completed, tested successfully and in operation. Operational performance highlights of the new upgraded system are presented in this paper.

  12. Conceptual & Engineering Design of Plug-in Cryostat Cylinder for Super-Conducting Central Solenoid of SST-1

    Science.gov (United States)

    Biswas, Prabal; Santra, Prosenjit; Vasava, Kirit; Jayswal, Snehal; Parekh, Tejas; Chauhan, Pradeep; Patel, Hitesh; Pradhan, Subrata

    2017-04-01

    SST-1, country’s first indigenously built steady state super-conducting tokamak is planned to be equipped with an Nb3Sn based superconducting central solenoid, which will replace the existing copper conductor TR1 coil for the purpose of Ohmic breakdown. This central solenoid (CS) of four layers with each layer having 144 turns with an OD of 573 mm, ID of 423 mm length of 2483 mm will be housed inside a high vacuum, CRYO compatible plug-in cryostat thin shell having formed from SS 304L plate duly rolled and welded to form cylinder along with necessary accessories like LN2 bubble panel, current lead chamber, coil and cylinder support structure etc. This paper will present the design drivers, material selection, advantages and constraints of the plug-in cryostat concept, sub-systems of plug-in cryostat, its conceptual and engineering design, CAD models, finite element analysis using ANSYS, safety issues and diagnostics, on-going works about fabrication, quality assurance/control and assembly/integration aspects with in the existing SST-1 machine bore.

  13. MTX [Microwave Tokamak Experiment] plasma diagnostic system

    International Nuclear Information System (INIS)

    Rice, B.W.; Hooper, E.B.; Brooksby, C.A.

    1987-01-01

    In this paper, a general overview of the MTX plasma diagnostics system is given. This includes a description of the MTX machine configuration and the overall facility layout. The data acquisition system and techniques for diagnostic signal transmission are also discussed. In addition, the diagnostic instruments planned for both an initial ohmic-heating set and a second FEL-heating set are described. The expected range of plasma parameters along with the planned plasma measurements will be reviewed. 7 refs., 5 figs

  14. Design and Architecture of SST-1 basic plasma control system

    Energy Technology Data Exchange (ETDEWEB)

    Patel, Kirit, E-mail: kpatel@ipr.res.in; Raju, D.; Dhongde, J.; Mahajan, K.; Chudasama, H.; Gulati, H.; Chauhan, A.; Masand, H.; Bhandarkar, M.; Pradhan, S.

    2016-11-15

    Highlights: • Reflective Memory network. • FPAG based Timing system for trigger distribution. • IRIG-B network for GPS time synchronization. • PMC based Digital Signal Processors and VME. • Simultaneous sampling ADC. - Abstract: Primary objective of SST-1 Plasma control system is to achieve Plasma position, shape and current profile control. Architecture of control system for SST-1 is distributed in nature. Fastest control loop time requirement of 100 μs is achieved using VME based simultaneous sampling ADCs, PMC based quad core DSP, Reflective Memory [RFM] based real-time network, VME based real-time trigger distribution network and Ethernet network. All the control loops for shape control, position control and current profile control share common signals from Magnetic diagnostic so it is planned to accommodate all the algorithms on the same PMC based quad core DSP module TS C-43. RFM based real-time data network replicate data from one node to next node in a ring network topology at sustained throughput rate of 13.4 MBps. Real-time Timing System network provides guaranteed trigger distribution in 3.8 μs from one node to all node of the network. Monitoring and configuration of different systems participating in the operation of SST-1 is done by Ethernet network. Magnetic sensors data is acquired using Pentek 6802 simultaneously sampling ADC card at the rate of 10KSPS. All the real-time raw data along with the control data will be archived using RFM network and SCSI HDD for the experiment duration of 1000 s. RFM network is also planned for real-time plotting of key parameter of Plasma during long experiment. After experiment this data is transferred to central storage server for archival purpose. This paper discusses the architecture and hardware implementation of the control system by describing all the involved hardware and software along with future plans for up-gradations.

  15. Progress in diagnostics of the COMPASS tokamak.

    Czech Academy of Sciences Publication Activity Database

    Weinzettl, Vladimír; Adámek, Jiří; Berta, Miklós; Bílková, Petra; Bogár, Ondrej; Böhm, Petr; Cavalier, Jordan; Dejarnac, Renaud; Dimitrova, Miglena; Ficker, Ondřej; Fridrich, David; Grover, Ondřej; Háček, Pavel; Havlíček, Josef; Havránek, Aleš; Horáček, Jan; Hron, Martin; Imríšek, Martin; Komm, Michael; Kovařík, Karel; Krbec, Jaroslav; Markovič, Tomáš; Matveeva, Ekaterina; Mitošinková, Klára; Mlynář, Jan; Naydenkova, Diana; Pánek, Radomír; Papřok, Richard; Peterka, Matěj; Podolník, Aleš; Seidl, Jakub; Šos, Miroslav; Stöckel, Jan; Tomeš, Matěj; Varavin, Mykyta; Varju, Jozef; Vlainic, Milos; Vondráček, Petr; Zajac, Jaromír; Žáček, František; Stano, M.; Anda, G.; Dunai, D.; Krizsanóczi, T.; Refy, D.; Zoletnik, S.; Silva, A.; Gomes, R.; Pereira, T.; Popov, Tsv.; Sarychev, D.; Ermak, G.P.; Zebrowski, J.; Jakubowski, M.; Rabinski, M.; Malinowski, K.; Nanobashvili, S.; Spolaore, M.; Vianello, N.; Gauthier, E.; Gunn, J. P.; Devitre, A.

    2017-01-01

    Roč. 12, December (2017), č. článku C12015. ISSN 1748-0221. [European Conference on Plasma Diagnostics (ECPD2017)/2./. Bordeaux, 18.04.2017-21.04.2017] R&D Projects: GA MŠk(CZ) LM2015045; GA ČR(CZ) GA15-10723S; GA ČR(CZ) GA16-25074S; GA ČR(CZ) GA14-35260S Institutional support: RVO:61389021 Keywords : Detector design and construction technologies and materials * Nuclear instruments and methods for hot plasma diagnostics * Plasma diagnostics - interferometry,spectroscopy and imaging * Plasma diagnostics - probes Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.220, year: 2016 http://iopscience.iop.org/article/10.1088/1748-0221/12/12/C12015/pdf

  16. Diagnostics Neutral Beam Injector at the TCV Tokamak

    International Nuclear Information System (INIS)

    Mlynar, J.; Shukaev, A.N.; Bosshard, P.; Duval, B.P.; Ivanov, A.A.; Kollegov, M.; Kolmogorov, V.V.; Llobet, X.; Pitts, R.A.; Weisen, H.

    2001-10-01

    Within this report we summarize the technical and experimental effort made on diagnostics neutral beam injector (DNBI) which was installed at tokamak TCV last year. Basic components of DNBI are reviewed, its remote control is presented in more detail. Profile and attenuation studies are referred to. First experimental results obtained with DNBI, which led to a decision to upgrade the machine, are discussed in the last section. (author)

  17. Edge Plasma Physics and Relevant Diagnostics on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Devynck, P.; Gunn, J.; Martines, E.; Bonhomme, G.; Van Oost, G.; Hron, Martin; Ďuran, Ivan; Pánek, Radomír; Stejskal, Pavel; Adámek, Jiří

    2004-01-01

    Roč. 3, - (2004), s. 1-6 ISSN 1433-5581. [First Cairo Conference on Plasma Physics & Applications. Cairo, 11.10.2003-15.10.2003] R&D Projects: GA ČR GA202/03/0786; GA ČR GP202/03/P062 Keywords : tokamak * edge plasma * probe diagnostics * biasing * turbulence * polarization Subject RIV: BL - Plasma and Gas Discharge Physics

  18. Lithium beam diagnostic system on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Anda, G.; Bencze, A.; Berta, Miklós; Dunai, D.; Háček, Pavel; Krbec, Jaroslav; Réfy, D.; Krizsanóczi, T.; Bató, S.; Ilkei, T.; Kiss, I.G.; Veres, G.; Zoletnik, S.

    2016-01-01

    Roč. 108, October (2016), s. 1-6 ISSN 0920-3796 R&D Projects: GA MŠk(CZ) LM2011021 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : BES * Plasma diagnostics * COMPASS tokamak Plasma density profile Plasma current fluctuations * Plasma density profile * Plasma current fluctuations Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379616303131

  19. Advanced probes for edge plasma diagnostics on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Adámek, Jiří; Balan, P.; Hronová-Bilyková, Olena; Brotánková, Jana; Dejarnac, Renaud; Devynck, P.; Ďuran, Ivan; Gunn, J. P.; Hron, Martin; Horáček, Jan; Ionita, C.; Kocan, M.; Martines, E.; Pánek, Radomír; Peleman, P.; Schrittwieser, R.; Van Oost, G.; Žáček, František

    2006-01-01

    Roč. 63, č. 0 (2006), 012001-012002 E-ISSN 1742-6596. [SECOND INTERNATIONAL WORKSHOP AND SUMMER SCHOOL ON PLASMA PHYSICS. Kiten, 03.07.2006-09.07.2006] R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma * tokamak * electric probes * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  20. Operation of SST-1 TF power supply during SST-1 campaigns

    International Nuclear Information System (INIS)

    Sharma, Dinesh Kumar; Vora, Murtuza M.; Ojha, Amit; Singh, Akhilesh Kumar; Bhavsar, Chirag

    2015-01-01

    Highlights: • SST-1 TF power supply is 12 pulse SCR converter circuit. • TF power supply protection, measurement and control scheme are explained. • Quench, emergency and normal shot process is explained and results of SST-1 campaigns are shown. • Dynamic control of TF current. • The paper shows the results of last ten SST-1 campaigns. - Abstract: SST-1 TF power supply provides the direct current for the required magnetic field of TF coil. TF power supply includes transformer, 12-pulse converter, bus bar, water-cooled cable, protection and measuring equipments, and isolator, VME DAC system and GUI software. TF power supply is operated through GUI software built in TCL/Tk. VME DAC system monitors the parameters, provides On/Off commands, voltage and current references and initiates predefined reference to emergency shutdown. The emergency shutdown is hardwired to TF power supply from central control. During quench power supply converter opens DCCB and dump resistor is connected in the circuit and VME DAC system acquires bus bar voltage, dump voltage and dump current. Operation of TF power supply also requires monitoring of SCR and transformer temperature and water flow rate of water-cooled cable during high current long pulse shot. Before start up of TF power supply a quench simulation is performed to check the readiness of protection. This paper describes pre startup operation, normal shot operation, emergency and quench process, dynamic control and complete shutdown operation of TF power supply.

  1. Operation of SST-1 TF power supply during SST-1 campaigns

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Dinesh Kumar, E-mail: dinesh@ipr.res.in; Vora, Murtuza M.; Ojha, Amit; Singh, Akhilesh Kumar; Bhavsar, Chirag

    2015-10-15

    Highlights: • SST-1 TF power supply is 12 pulse SCR converter circuit. • TF power supply protection, measurement and control scheme are explained. • Quench, emergency and normal shot process is explained and results of SST-1 campaigns are shown. • Dynamic control of TF current. • The paper shows the results of last ten SST-1 campaigns. - Abstract: SST-1 TF power supply provides the direct current for the required magnetic field of TF coil. TF power supply includes transformer, 12-pulse converter, bus bar, water-cooled cable, protection and measuring equipments, and isolator, VME DAC system and GUI software. TF power supply is operated through GUI software built in TCL/Tk. VME DAC system monitors the parameters, provides On/Off commands, voltage and current references and initiates predefined reference to emergency shutdown. The emergency shutdown is hardwired to TF power supply from central control. During quench power supply converter opens DCCB and dump resistor is connected in the circuit and VME DAC system acquires bus bar voltage, dump voltage and dump current. Operation of TF power supply also requires monitoring of SCR and transformer temperature and water flow rate of water-cooled cable during high current long pulse shot. Before start up of TF power supply a quench simulation is performed to check the readiness of protection. This paper describes pre startup operation, normal shot operation, emergency and quench process, dynamic control and complete shutdown operation of TF power supply.

  2. Thermal structural analysis of SST-1 vacuum vessel and cryostat assembly using ANSYS

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Bedakihale, Vijay; Ranganath, Tata

    2009-01-01

    Steady state super-conducting tokamak-1 (SST-1) is a medium sized tokamak, which has been designed to produce a 'D' shaped double null divertor plasma and operate in quasi steady state (1000 s). SST-1 vacuum system comprises of plasma chamber (vacuum vessel, interconnecting rings, baking and cooling channels), and cryostat all made of SS 304L material designed to meet ultra high vacuum requirements for plasma generation and confinement. Prior to plasma shot and operation the vessel assembly is baked to 250/150 deg. C from room temperature and discharge cleaned to remove impurities/trapped gases from wall surfaces. Due to baking the non-uniform temperature pattern on the vessel assembly coupled with atmospheric pressure loading and self-weight give rise to high thermal-structural stresses, which needs to be analyzed in detail. In addition the vessel assembly being a thin shell vessel structure needs to be checked for critical buckling load caused by atmospheric and baking thermal loads. Considering symmetry of SST-1, 1/16th of the geometry is modeled for finite element (FE) analysis using ANSYS for different loading scenarios, e.g. self-weight, pressure loading considering normal operating conditions, and off-normal loads coupled with baking of vacuum vessel from room temperature 250 deg. C to 150 deg. C, buckling and modal analysis for future dynamic analysis. The paper will discuss details about SST-1 vacuum system/cryostat, solid and FE model of SST-1, different loading scenarios, material details and the stress codes used. We will also present the thermal structural results of FE analysis using ANSYS for various load cases being investigated and our observations under different loading conditions.

  3. Criticality in the fabrication of ion extraction system for SST-1 neutral beam injector

    International Nuclear Information System (INIS)

    Jana, M.R.; Mattoo, S.K.

    2008-01-01

    For the heating of plasma in steady-state superconducting tokamak (SST-1) (Y.C. Saxena, SST-1 Team, Present status of the SST-1 project, Nucl. Fusion 40 (2000) 1069-1082; D. Bora, SST-1 Team, Test results on systems developed for the SST-1 tokamak, Nucl. Fusion 43 (2003) 1748-1758), a neutral beam injector is provided to raise the ion temperature to ∼1 keV. This injector has a capability of injecting hydrogen beam with the power of 0.5 MW at 30 keV. For the upgrade of SST-1, power of 1.7 MW at 55 KeV is required. Further, beam power is to be provided for a pulse length of 1000S. We have designed a neutral beam injector (S.K. Mattoo, A.K. Chakraborty, U.K. Baruah, P.K. Jayakumar, M. Bandyopadhyay, N. Bisai, Ch. Chakrapani, M.R. Jana, R. Onali, V. Prahlad, P.J. Patel, G.B. Patel, B. Prajapati, N.V.M. Rao, S. Rambabu, C. Rotti, S.K. Sharma, S. Shah, V. Sharma, M.J. Singh, Engineering design of the steady-state neutral beam injector for SST-1, Fusion Eng. Des. 56 (2001) 685-691; A.K. Chakraborty, N. Bisai, M.R. Jana, P.K. Jayakumar, U.K. Baruah, P.J. Patel, K. Rajasekar, S.K. Mattoo, Neutral beam injector for steady-state superconducting tokamak, Fusion Technol. (1996) 657-660; P.K. Jayakumar, M.R. Jana, N. Bisai, M. Bajpai, N.P. Singh, U.K. Baruah, A.K. Chakraborty, M. Bandyopadhyay, C. Chrakrapani, D. Patel, G.B. Patel, P. Patel, V. Prahlad, N.V.M. Rao, C. Rotti, V. Sreedhar, S.K. Mattoo, Engineering issues of a 1000S neutral beam ion source, Fusion Technol. 1 (1998) 419-422) satisfying the requirements for both SST-1 and its upgrade. Since intense power is to be transported to SST-1 situated at a distance of several meters from the ion source, the optical quality of the beam becomes a primary concern. This in turn, is determined by the uniformity of the ion source plasma and the extractor geometry. To obtain the desired optical quality of the beam, stringent tolerances are to be met during the fabrication of ion extractor system. SST-1 neutral beam injector is

  4. Instrumentation for NBI SST-1 cooling water system

    International Nuclear Information System (INIS)

    Qureshi, Karishma; Patel, Paresh; Jana, M.R.

    2015-01-01

    Neutral Beam Injector (NBI) System is one of the heating systems for Steady state Superconducting Tokamak (SST-1). It is capable of generating a neutral hydrogen beam of power 0.5 MW at 30 kV. NBI system consists of following sub-systems: Ion source, Neutralizer, Deflection Magnet and Magnet Liner (ML), Ion Dump (ID), V-Target (VT), Pre Duct Scraper (PDS), Beam Transmission Duct (BTD) and Shine Through (ST). For better heat removal management purpose all the above sub-systems shall be equipped with Heat Transfer Elements (THE). During beam operation these sub-systems gets heated due to the received heat load which requires to be removed by efficient supplying water. The cooling water system along with the other systems (External Vacuum System, Gas Feed System, Cryogenics System, etc.) will be controlled by NBI Programmable Logic Control (PLC). In this paper instrumentation and its related design for cooling water system is discussed. The work involves flow control valves, transmitters (pressure, temperature and water flow), pH and conductivity meter signals and its interface with the NBI PLC. All the analog input, analog output, digital input and digital output signals from the cooling water system will be isolated and then fed to the NBI PLC. Graphical Users Interface (GUI) needed in the Wonderware SCADA for the cooling water system shall also be discussed. (author)

  5. Simulation study of induced EMFs and the suppression during SST-1 start-up

    Energy Technology Data Exchange (ETDEWEB)

    Jain, V., E-mail: vishal@ipr.res.in; Sharma, D.; Vardhrajulu, A.; Gupta, C.N.; Srinivasan, R.; Daniel, R.

    2015-11-15

    Highlights: • Induced EMFs study in PF coils during SST-1 start up using MATlab simulink. • Integration of wave shaping network to generate practical OT current profile. • This study would protect coil insulation with identifying RC circulating network. • Study of MOV technique for circulation of current through RC. - Abstract: Steady State Superconducting Tokamak (SST-1) comprises of various copper and superconducting coils for generating magnetic field for initiation, providing equilibrium and shaping of plasma in tokamak. In this paper, an attempt is made to study the induced EMF in superconducting poloidal field coils (PF coils) due to fast ramp down of current in ohmic transformer copper coils (OT coils) for SST-1 plasma initiation. The fast ramp down of current, from few kA to zero amperes in just 50–100 ms in OT coils, is required to achieve plasma breakdown and ramp up of plasma current in tokamak. However, it induces nearly 5 kV EMF in one of the SST-1 PF coils that can damage the coil insulation and also bias negatively the electronic switching of power supply. It is necessary to maintain induced EMF below 1 kV in all PF coils for safe operation of SST-1. The induced EMF up to 1 kV can be clamped without any need of protection and circulating current. If the induced EMF is in excess of 1 kV, then it has to allow the circulation of current through RC network for coil protection from overvoltage. These circulating currents in PF coils will affect the shaping of plasma. In this paper, the induced EMF in PF coils are simulated using MATlab simulink for a typical SST-1 current profile of OT coils. Further, this simulation study is used to design the protection system for PF coils. In this paper, the worst-case induced EMF scenario is considered by excluding the effect of passive elements like vacuum vessel and cryostat on mutual coupling parameters. However, the implementation of the EMF suppression scheme need more elaborated study with considering

  6. Lithium beam diagnostic system on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anda, G.; Bencze, A. [Wigner – RCP, HAS, Budapest (Hungary); Berta, M., E-mail: bertam@sze.hu [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Hacek, P. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Réfy, D.; Krizsanóczi, T.; Bató, S.; Ilkei, T.; Kiss, I.G.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)

    2016-10-15

    Highlights: • Li-beam diagnostic system on the COMPASS tokamak is an improved and compact system to allow testing of Atomic Beam Probe. • The possibility to measure background corrected density profiles on the few microseconds time scale. • First Li-beam diagnostic system with recirculating neutralizer. • The system includes the redesigned ion source with longer lifetime. - Abstract: An improved lithium beam based beam emission spectroscopy system – installed on COMPASS tokamak – is described. The beam energy enhanced up to 120 keV for Atomic Beam Probe measurement. The size of the ion source is doubled, using a newly developed thermionic heater instead of the conventionally used heating (tungsten or molybdenum) filament. The neutralizer is also improved. It produces the same sodium vapor in a cell but minimize the loss condensing the vapor on a cold surface which is led back (in fluid state) into the sodium oven. This way we call it recirculating neutralizer. The observation system consists of a CCD camera and an avalanche photodiode array.

  7. Remote network control plasma diagnostic system for Tokamak T-10

    International Nuclear Information System (INIS)

    Troynov, V I; Zimin, A M; Krupin, V A; Notkin, G E; Nurgaliev, M R

    2016-01-01

    The parameters of molecular plasma in closed magnetic trap is studied in this paper. Using the system of molecular diagnostics, which was designed by the authors on the «Tokamak T-10» facility, the radiation of hydrogen isotopes at the plasma edge is investigated. The scheme of optical radiation registration within visible spectrum is described. For visualization, identification and processing of registered molecular spectra a new software is developed using MatLab environment. The software also includes electronic atlas of electronic-vibrational-rotational transitions for molecules of protium and deuterium. To register radiation from limiter cross-section a network control system is designed using the means of the Internet/Intranet. Remote control system diagram and methods are given. The examples of web-interfaces for working out equipment control scenarios and viewing of results are provided. After test run in Intranet, the remote diagnostic system will be accessible through Internet. (paper)

  8. Study of radiation heat transfer between PFC and vacuum vessel during SST-1 baking

    Energy Technology Data Exchange (ETDEWEB)

    Chaudhuri, Paritosh E-mail: paritosh@ipr.res.in; Chenna Reddy, D.; Santra, P.; Khirwadkar, S.; Ravi Pragash, N.; Saxena, Y.C

    2003-01-01

    Steady-state superconducting tokamak (SST-1) is a medium size tokamak with superconducting magnetic field coils. Plasma facing components (PFC) of SST-1 are placed inside the vacuum vessel (VV) of the tokamak and are designed to be compatible for steady-state operation. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m{sup 2}. In addition to remove high heat fluxes, the PFC are also designed to be compatible for baking at high temperature. Since it is difficult to calculate the radiation heat loads between PFC and VV in a 3-D irregular geometry, a simplified model of concentric cylinders has been chosen for the purpose of estimation of the power requirements and the thermal responses of PFC and VV during their bakeout phases. Thermal responses of the PFC and VV have been analysed and the analytical results have been compared with 2-D finite element analysis using ANSYS. The radiation losses between PFC and VV also have been evaluated on the actual model containing all PFC inside the VV.

  9. Design of vessel baking system and thermal radiation shields for SST-1

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, E.R.; Nagabhushana, S.; Pathak, H.A.; Panigrahi, S.; Nath, T.R.; Babu, A.V.S; Gangradey, R.; Patel, R.J.; Saxena, Y.C. [Institute for Plasma Research, Gandhinagar (India)

    1998-07-01

    SST-1 is a Steady State Tokamak with a major radius of 1.1 m, minor radius of 0.2 m and toroidal field of 3.0 T. The toroidal and poloidal field coils of SST-1 are superconducting. One of the main objectives of SST-1 is to demonstrate steady state particle removal and active plasma density control which states the necessity of wall conditioning. The vacuum vessel will be baked up to 525 K by passing hot nitrogen gas through the U - channels welded on the inner surface of vacuum vessel. The required mass flow rate at 5 bar is 0.712 Kg/s to maintain 525 K wall temperature in steady state. Superconducting coils operating at 4.5 K will be protected against thermal radiation from hot surfaces using liquid nitrogen cooled panels operating at 87 K. Maximum 1200 litres/hour liquid nitrogen is required during vessel baking. The design of vacuum vessel baking system and thermal radiation shields and related flow analysis are presented here. (authors)

  10. Design of vessel baking system and thermal radiation shields for SST-1

    International Nuclear Information System (INIS)

    Kumar, E.R.; Nagabhushana, S.; Pathak, H.A.; Panigrahi, S.; Nath, T.R.; Babu, A.V.S; Gangradey, R.; Patel, R.J.; Saxena, Y.C.

    1998-01-01

    SST-1 is a Steady State Tokamak with a major radius of 1.1 m, minor radius of 0.2 m and toroidal field of 3.0 T. The toroidal and poloidal field coils of SST-1 are superconducting. One of the main objectives of SST-1 is to demonstrate steady state particle removal and active plasma density control which states the necessity of wall conditioning. The vacuum vessel will be baked up to 525 K by passing hot nitrogen gas through the U - channels welded on the inner surface of vacuum vessel. The required mass flow rate at 5 bar is 0.712 Kg/s to maintain 525 K wall temperature in steady state. Superconducting coils operating at 4.5 K will be protected against thermal radiation from hot surfaces using liquid nitrogen cooled panels operating at 87 K. Maximum 1200 litres/hour liquid nitrogen is required during vessel baking. The design of vacuum vessel baking system and thermal radiation shields and related flow analysis are presented here. (authors)

  11. Design and implementation of quench detection instrumentation for TF magnet system of SST-1

    International Nuclear Information System (INIS)

    Khristi, Y.; Sharma, A.N.; Doshi, K.; Banaudha, M.; Prasad, U.; Varmora, P.; Patel, D.; Pradhan, S.

    2014-01-01

    Steady State Superconducting Tokamak-1 (SST-1) at Institute for Plasma Research (IPR), India is now in engineering validation phase. The assembled Toroidal Field (TF) magnet system of SST-1 will be operated at 10 kA of nominal current at helium cooled condition of 4.5 K. A reliable and fail proof quench detection (QD) system is essential for the safety and the investment protection requirements of the magnets. This QD system needs to continuously monitor all the superconducting coils, which include 16 TF magnets, return-loop, bus bars and current leads. In case of any event initiating the normal resistive zone and reaching thermal run-away, the QD system needs to trigger the magnet protection circuits. Precision instrumentation and control system with 204 signal channels had been developed for detection of quench anywhere in the entire TF magnet system. In the present configuration of quench detection scheme, the voltage drop across each double pancake (DP) of each TF coil are compared with its two adjacent DPs for the detection of normal zone and cancelation of inductive couples. Two identical redundant systems with one out of two configurations are successfully commissioned and tested at IPR. This paper describes the design and implementation of the QD system, Installation experience, validation test and initial results from the recent SST-1 magnet system charging

  12. Process optimization of helium cryo plant operation for SST-1 superconducting magnet system

    Science.gov (United States)

    Panchal, P.; Panchal, R.; Patel, R.; Mahesuriya, G.; Sonara, D.; Srikanth G, L. N.; Garg, A.; Christian, D.; Bairagi, N.; Sharma, R.; Patel, K.; Shah, P.; Nimavat, H.; Purwar, G.; Patel, J.; Tanna, V.; Pradhan, S.

    2017-02-01

    Several plasma discharge campaigns have been carried out in steady state superconducting tokamak (SST-1). SST-1 has toroidal field (TF) and poloidal field (PF) superconducting magnet system (SCMS). The TF coils system is cooled to 4.5 - 4.8 K at 1.5 - 1.7 bar(a) under two phase flow condition using 1.3 kW helium cryo plant. Experience revealed that the PF coils demand higher pressure heads even at lower temperatures in comparison to TF coils because of its longer hydraulic path lengths. Thermal run away are observed within PF coils because of single common control valve for all PF coils in distribution system having non-uniform lengths. Thus it is routine practice to stop the cooling of PF path and continue only TF cooling at SCMS inlet temperature of ˜ 14 K. In order to achieve uniform cool down, different control logic is adopted to make cryo stable system. In adopted control logic, the SCMS are cooled down to 80 K at constant inlet pressure of 9 bar(a). After authorization of turbine A/B, the SCMS inlet pressure is gradually controlled by refrigeration J-T valve to achieve stable operation window for cryo system. This paper presents process optimization for cryo plant operation for SST-1 SCMS.

  13. Engineering design and thermal hydraulics of plasma facing components of SST-1

    International Nuclear Information System (INIS)

    Pragash, N. Ravi; Chaudhuri, P.; Santra, P.; Chenna Reddy, D.; Khirwadkar, S.; Saxena, Y.C.

    2001-01-01

    SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (∼1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m 2 . In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 degree sign C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a FORTRAN code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ANSYS software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper

  14. Adaptation of fast responding power supply for radial position control in SST-1

    International Nuclear Information System (INIS)

    Sharma, Dinesh Kumar; Patel, Kiritkumar B.; Singh, Akhilesh Kumar; Dhongde, Jasraj

    2013-01-01

    A high current, fast responding power supply was installed in 2005 for vertical stabilization of elongated plasmas in SST-1 tokamak. Presently, during initial experiments of SST-1 tokamak the need for radial control during current build-up was envisaged. For this purpose the existing power supply was suitable and the same was re-commissioned and control adaptations were carried as per experimental requirements. This paper highlights the capabilities of the power supply and details the modifications in the control interfaces and test programs for the radial control purpose. Details of the operation of the power supply along with control interfaces with performance measurements are provided. The re-commissioning provided an opportunity in the trouble shooting methods and sequential operation of the system. With the operational use on the actual coil the mutual effects are understood better and appropriate test programs are prepared. The power supply provided satisfactory performance for the intended use. In additional the system is suitable to simulate a plasma current loop to enable the testing and calibration of Rogowski coil used for plasma current measurement. (author)

  15. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

  16. Thomson scattering diagnostic for the Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Foote, J.H.; Barter, J.D.; Sewall, N.R.; Jolly, J.J.; Schlander, L.F.

    1990-01-01

    The Thomson-scattering diagnostic system (TSS) on the Microwave Tokamak Experiment (MTX) at LLNL routinely monitors electron temperature (T e ) and density. Typical measured values at the plasma center under clean conditions are 900 ± 70 eV and 1 to 2 x 10 14 (±30%) cm -3 . The TSS apparatus is compact, with all elements mounted on one sturdy, two-level optics table. Because of this, we maintain with minimum effort the alignment of both the ruby-laser input optics and the scattered-light collecting optics. Undesired background signals, e.g., plasma light as well as ruby-laser light scattered off obstacles and walls, are generally small compared with the Thomson-scattered signals we normally detect. In the MTX T e region, the TSS data are definitely fitted better when relativistic effects are included in the equations. Besides determining the temperature of the Maxwellian electron distribution, the system is designed to detect electron heating from GW-level free-electron laser (FEL) pulses by measuring large wavelength shifts of the scattered laser photons. TSS data suggest that we may indeed by able to detect these electrons, which can have energies up to 10 keV, according to computer simulation. 7 refs., 4 figs

  17. Thomson scattering diagnostic for the microwave tokamak experiment

    International Nuclear Information System (INIS)

    Foote, J.H.; Barter, J.D.; Sewall, N.R.; Jolly, J.J.; Schlander, L.F.

    1990-01-01

    The Thomson scattering diagnostic system (TSS) on the microwave tokamak experiment (MTX) at LLNL routinely monitors electron temperature (T e ) and density. Typical measured values at the plasma center under clean conditions are 900±70 eV and 1--2x10 14 (±30%) cm -3 . The TSS apparatus is compact, with all elements mounted on one sturdy, two-level optics table. Because of this, we maintain with minimum effort the alignment of both the ruby-laser input optics and the scattered-light collecting optics. Undesired background signals, e.g., plasma light as well as ruby-laser light scattered off obstacles and walls, are generally small compared with the Thomson-scattered signals we normally detect. In the MTX T e region, the TSS data are definitely fitted better when relativistic effects are included in the equations. Besides determining the temperature of the Maxwellian electron distribution, the system is designed to detect electron heating from GW-level free-electron laser (FEL) pulses by measuring large wavelength shifts of the scattered laser photons. TSS data suggest that we may indeed be able to detect these electrons, which can have energies up to 10 keV, according to computer simulation

  18. Tokamak physics studies using x-ray diagnostic methods

    International Nuclear Information System (INIS)

    Hill, K.W.; Bitter, M.; von Goeler, S.

    1987-03-01

    X-ray diagnostic measurements have been used in a number of experiments to improve our understanding of important tokamak physics issues. The impurity content in TFTR plasmas, its sources and control have been clarified through soft x-ray pulse-height analysis (PHA) measurements. The dependence of intrinsic impurity concentrations and Z/sub eff/ on electron density, plasma current, limiter material and conditioning, and neutral-beam power have shown that the limiter is an important source of metal impurities. Neoclassical-like impurity peaking following hydrogen pellet injection into Alcator C and a strong effect of impurities on sawtooth behavior were demonstrated by x-ray imaging (XIS) measurements. Rapid inward motion of impurities and continuation of m = 1 activity following an internal disruption were demonstrated with XIS measurements on PLT using injected aluminum to enhance the signals. Ion temperatures up to 12 keV and a toroidal plasma rotation velocity up to 6 x 10 5 m/s have been measured by an x-ray crystal spectrometer (XCS) with up to 13 MW of 85-keV neutral-beam injection in TFTR. Precise wavelengths and relative intensities of x-ray lines in several helium-like ions and neon-like ions of silver have been measured in TFTR and PLT by the XCS. The data help to identify the important excitation processes predicted in atomic physics. Wavelengths of n = 3 to 2 silver lines of interest for x-ray lasers were measured, and precise instrument calibration techniques were developed. Electron thermal conductivity and sawtooth dynamics have been studied through XIS measurements on TFTR of heat-pulse propagation and compound sawteeth. A non-Maxwellian electron distribution function has been measured, and evidence of the Parail-Pogutse instability identified by hard x-ray PHA measurements on PLT during lower-hybrid current-drive experiments

  19. Cryogenic operation strategy for the SST-1 device

    International Nuclear Information System (INIS)

    Tanna, V.L.; Pradhan, S.

    2013-01-01

    The SST-1 has been operated since 2012 as part of its engineering commissioning and almost 5 experimental campaigns have been successfully completed. Before final assembling, cool-down and current excitation tests for the Toroidal field coils and PF 3 (Upper) coil were demonstrated successfully as part of validation under coils test program. These superconducting coils consist of a cable-in-conduit conductor, (CICC) is cooled by the forced-flow Two-phase flow as well as supercritical helium conditions. During the recent campaigns, hydraulic characteristics of whole superconducting magnets along with the TF case cooling were studied as an integral system. Based on the experimental observations, efforts have been made to cryo stable conditions of the SST-1 superconducting magnets system in order to produce steady state TF magnetic field of 1.5 T at the plasma center. Optimization of Helium plant related processes have been worked out and implemented to realize the successful SST-1 device operation over a week. In order to have long experimental campaign, an intermediate temperature cooling down philosophy has been adopted. The complete superconducting coils flow distribution among their cooling channels and pressure head requirements were studied from the measurements. In this paper, we will highlight the recent cool-down results, flow distribution and temperature uniformity aspects while cooling down the SST-1 magnets system. (author)

  20. Extraordinary mode absorption at the electron cyclotron harmonic frequencies as a Tokamak plasma diagnostic

    International Nuclear Information System (INIS)

    Pachtman, A.

    1986-09-01

    Measurements of Extraordinary mode absorption at the electron cyclotron harmonic frequencies are of unique value in high temperature, high density Tokamak plasma diagnostic applications. An experimental study of Extraordinary mode absorption at the semi-opaque second and third harmonics has been performed on the ALCATOR C Tokamak. A narrow beam of submillimeter laser radiation was used to illuminate the plasma in a horizontal plane, providing a continuous measurement of the one-pass, quasi-perpendicular transmission

  1. Bridge joint fabrication and validation for SST-1 PF coil winding pack

    International Nuclear Information System (INIS)

    Prasad, Upendra; Sharma, A.N.; Patel, D.; Doshi, K.; Varmora, P.; Khristi, Y.; Pradhan, S.

    2014-01-01

    Highlights: • Prototype of bridge type joints fabricated and validated successfully. • Bridge type joints fabricated and validated on one of the SST-1 PF#3T coil successfully. • Joint resistance was measured with precision nano volt meter and PXI based data acquisition system. • Leak tightness of joint box was better than 3 × 10 −6 Pa m 3 s −1 . • The measured joint resistance of bridge type joint was ∼1.6 nano ohm. - Abstract: A novel concept of bridge joint for Poloidal field (PF) magnet of SST-1 with damaged winding pack has been realized. This joint has been fabricated on 5th and 6th layers of PF#3T coil winding pack (WP) after validation at 10 kA at liquid helium temperature of 4.2 K in current lead test chamber. The joint resistance of bridge joint was measured ∼1.6 nΩ at flat top DC current of 10 kA. This type of joint could be economically useful for revival of a shorted and damaged WP superconducting PF magnets of Tokamaks. In this paper, details of bridge joint design, fabrication and validations are discussed

  2. Quality control of FWC during assembly/commissioning on SST-1

    International Nuclear Information System (INIS)

    Patel, Hiteshkumar; Santra, Prosenjit; Jaiswal, Snehal

    2015-01-01

    First Wall components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring and port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 under going a rigorous quality control and checks at every stage of the assembly process. This paper will present the quality control and checks of FWC from commencement of assembly procedure, namely material test reports, leak testing of high temperature baked components, assembled dimensional tolerances, leak testing of all welded joints, graphite tile tightening torques, electrical continuity of passive stabilizers, and electrical isolation of passive stabilizers from vacuum vessel, baking and cooling hydraulic connections inside vacuum vessel. (author)

  3. Helium leak testing of superconducting magnets, thermal shields and cryogenic lines of SST -1

    International Nuclear Information System (INIS)

    Thankey, P.L.; Joshi, K.S.; Semwal, P.; Pathan, F.S.; Raval, D.C.; Khan, Z.; Patel, R.J.; Pathak, H.A.

    2005-01-01

    Tokamak SST - 1 is under commissioning at Institute for Plasma Research. It comprises of a toroidal doughnut shaped plasma chamber, surrounded by liquid helium cooled superconducting magnets, housed in a cryostat chamber. The cryostat has two cooling circuits, (1) liquid nitrogen cooling circuit operating at 80 K to minimize the radiation heat load on the magnets, and (2) liquid helium cooling circuit to cool magnets and cold mass support structure to 4.5 K. In this paper we describe (a) the leak testing of copper - SS joints, brazing joints, interconnecting joints of the superconducting magnets, and (b) the leak testing of the liquid nitrogen cooling circuit, comprising of the main supply header, the thermal shields, interconnecting pipes, main return header and electrical isolators. All these tests were carried out using both vacuum and sniffer methods. (author)

  4. Electrical probe diagnostic with fast data acquisition systems of the Novillo tokamak

    International Nuclear Information System (INIS)

    Lopez-Callejas, R.; Benitez-Read, J.S.; Longoria-Gandara, L.C.; Pacheco-Sotelo, J.O.; Valencia-Alvarado, R.; Tamayo, F.J.; Valdes, A.; Fernandez, M.C.; Serrano, F.

    2001-01-01

    This paper describes the electrical probe diagnostic system used to measure the relevant parameters of the plasma column generated in the Novillo tokamak, and an inexpensive and fast data acquisition system that consists of 16 independent channels, 8 bit resolution, sampling frequency of 500 kHz, and 8 kword memory per channel. A distributed, modular and transparent approach was used for designing the software-operated data acquisition, diagnostic, and capacitor banks triggering systems, satisfying the pulsed nature of the tokamak discharge. Through a personal computer, the experimental data are available to engineers and physicists in a centralized database

  5. Operational and troubleshooting experiences in the SST-1 cryogenic system

    Science.gov (United States)

    Mahesuria, G.; Panchal, P.; Panchal, R.; Patel, R.; Sonara, D.; Gupta, N. C.; Srikanth, G. L. N.; Christian, D.; Garg, A.; Bairagi, N.; Patel, K.; Shah, P.; Nimavat, H.; Sharma, R.; Patel, J. C.; Tank, J.; Tanna, V. L.; Pradhan, S.

    2014-01-01

    Recently, the cooldown and current charging campaign have been carried out towards the demonstration of the first successful plasma discharge in the steady state superconducting Tokomak (SST-1). The SST-1 machine consists of cable-in-conduit wound superconducting toroidal as well as poloidal coils, cooled using 1.3 kW at 4.5 K helium refrigerator -cum- liquefier (HRL) system. The cryo system provides the two-phase helium at 0.13 MPa at 4.5 K as well as forced-flow pressurized helium at 0.4 MPa and in addition to 7 g-s-1 liquefaction capacity required for the current leads and other cold mass at 4.5 K. The entire integrated cold masses having different thermo hydraulic resistances cooled with the SST-1 HRL in optimised process parameters. In order to maintain different levels of temperatures and to facilitate smooth and reliable cooldown, warm-up, normal operations as well as to handle abnormal events such as, quench or utilities failures etc., exergy efficient process are adopted for the helium refrigerator-cum-liquefier (HRL) with an installed equivalent capacity of 1.3 kW at 4.5 K. Using the HRL, the cold mass of about 40 tons is being routinely cooled down from ambient temperature to 4.5 K with an average cooldown rate of 0.75 - 1 K-h-1. Long-term cryogenic stable conditions were obtained within 15 days in the superconducting coils and their connecting feeders. Afterwards, all of the cold mass is warmed-up in a controlled manner to ambient temperature. In this paper, we report the recent operational results of the cryogenic system during the first plasma discharge in SST-1 as well as the troubleshooting experiences of the cryogenic plant related hardware.

  6. Experience of superconducting current feeders system of SST-1

    International Nuclear Information System (INIS)

    Gupta, N.C.; Garg, A.; Sonara, D.

    2014-01-01

    The superconducting current feeder system for SST-1 which has been installed and commissioned recently along with SST-1, felicitates to energize the SST-1. The CFS consists of ten pairs of 10,000 Ampere (A) rating helium vapor cooled conventional current leads, interconnecting Cu-SC joints, three numbers of cryo-compatible SC feeders ducts, current leads assembly chamber, hydraulic network and three numbers of joint boxes operated at different current rating to charge Toroidal Field and Poloidal Field coils separately. During the last three campaigns, it was possible to achieve a controlled cool down up to 4 K and showed its rated operational performance. Actively cooled liquid nitrogen shield showed temperature profile in the temperature range of 80-85K and the whole system was evacuated up to 6x10 -6 mbar. The measured LHe consumption rates from TF VCCL were 0.3 g/s and 0.35 g/s at zero current and 1 kA respectively. (author)

  7. Diagnostic applications of transient synchrotron radiation in tokamak plasmas

    International Nuclear Information System (INIS)

    Fisch, N.J.; Kritz, A.H.

    1990-02-01

    Transient radiation, resulting from a brief, deliberate perturbation of the velocity distribution of superthermal tokamak electrons, can be more informative than the steady background radiation that is present in the absence of the perturbation. It is possible to define a number of interesting inverse problems, which exploit the two-dimensional frequency-time data of the transient radiation signal. 17 refs

  8. Operation and control of high power Gyrotrons for ECRH systems in SST-1 and Aditya

    Energy Technology Data Exchange (ETDEWEB)

    Shukla, B.K., E-mail: shukla@ipr.res.in; Bora, D.; Jha, R.; Patel, Jatin; Patel, Harshida; Babu, Rajan; Dhorajiya, Pragnesh; Dalakoti, Shefali; Purohit, Dharmesh

    2016-11-15

    Highlights: • Operation and control of high power Gyrotrons. • Data acquisition and control (DAQ) for Gyrotron system. • Ignitron based crowbar protection. • VME and PXI based systems. - Abstract: The Electron Cyclotron Resonance Heating (ECRH) system is an important heating system for the reliable start-up of tokamak. The 42 GHz and 82.6 GHz ECRH systems are used in tokamaks SST-1 and Aditya to carry out ECRH related experiments. The Gyrotrons are high power microwave tubes used as a source for ECRH systems. The Gyrotron is a delicate microwave tube, which deliver megawatt level power at very high voltage ∼40–50 kV with the current requirement ∼10 A–50 A. The Gyrotrons are associated with the subsystems like: High voltage power supplies (Beam voltage and anode voltage), dedicated crowbar system, magnet, filament and ion pump power supplies, cooling, interlocks and a dedicated data acquisition & control (DAC) system. There are two levels of interlocks used for the protection of Gyrotron: fast interlocks (arcing, beam over current, dI/dt, anode voltage and anode over current etc.) operate within 10 μs and slow interlocks (cooling, filament, silence of Gyrotron, ion pump and magnet currents) operate within 100 ms. Two Gyrotrons (42 GHz/500 kW/500 ms and 82.6 GHz/200 kW/1000 s) have been commissioned on dummy load for full parameters. The 42 GHz ECRH system has been integrated with SST-1 & Aditya tokamak and various experiments have been carried out related to ECRH assisted breakdown and start-up of tokamak at fundamental and second harmonic. These Gyrotrons are operated with VME based data acquisition and control (DAC) system. The DAC system is capable to acquire 64 digital and 32 analog signals. The system is used to monitor & acquire the data and also used for slow interlocks for the protection of Gyrotron. The data acquired from the system are stored online on VME system and after the shot stored in a file in binary format. The MDSPlus, a set of

  9. First 50 pps Thomson scattering diagnostics in a tokamak

    International Nuclear Information System (INIS)

    Roehr, H.; Schramm, G.; Steuer, K.H.; Hirsch, K.; Salzmann, H.

    1981-12-01

    Electron temperature and density measurements by Thomson scattering were performed for the first time for the whole duration of a tokamak discharge. A 50 pps Nd:YAG laser at 1.06 μm was used in ASDEX in combination with Si avalanche photodiode detectors. Density calibration was done by rotational anti-Stokes Raman scattering from hydrogen. The system is used for measurements at electron densities of as low as 2 x 10 12 cm -3 . (orig.)

  10. Power supply system on HT-7 tokamak for diagnostic neutral beam based on PLC

    International Nuclear Information System (INIS)

    Zhang Jian; Liu Baohua; Ding Tonghai; Du Shaowu

    2006-01-01

    A power supply system for diagnostic neutral beam on the HT-7 Tokamak was developed. Its logic control system based on S7-300 PLC was described. The experimental results show that the system is easy to operate and its performance is reliable. (authors)

  11. First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

    DEFF Research Database (Denmark)

    Bin, W.; Bruschi, A.; D'Arcangelo, O.

    2015-01-01

    Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin i...

  12. Diagnostics and control for the steady state and pulsed tokamak DEMO

    Czech Academy of Sciences Publication Activity Database

    Orsitto, F.P.; Villari, R.; Moro, F.; Todd, T.N.; Lilley, S.; Jenkins, I.; Felton, R.; Biel, W.; Silva, A.; Scholz, M.; Rzadkiewicz, J.; Ďuran, Ivan; Tardocchi, M.; Gorini, G.; Morlock, C.; Federici, G.; Litnovsky, A.

    2016-01-01

    Roč. 56, č. 2 (2016), č. článku 026009. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : measurement systems, fusion reactor, fusion plasma diagnostics * fusion reactor * fusion plasma diagnostics * DEMO * Hall sensors * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/2/026009

  13. Tokamak ion temperature and poloidal field diagnostics using 3 MeV protons

    International Nuclear Information System (INIS)

    Heidbrink, W.W.; Strachan, J.D.

    1984-10-01

    The 3 MeV protons created by d(d,p)t fusion reactions in a moderately sized tokamak leave the plasma on trajectories determined by the position of their birth and by the poloidal magnetic field. Pitch-angle resolution of the escaping 3 MeV protons can separately resolve the spatial distribution of the d(d,p)t fusion reactions and the poloidal field distribution inside the tokamak. These diagnostic techniques have been demonstrated on PLT with an array of collimated surface barrier detectors

  14. A study on the Fusion Reactor - Development of charge exchange recombination spectroscopy for tokamak diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tong Nyong; Kim, Dong Eon; Kim, Dae Sung; Kim, Seong Ho [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    1996-09-01

    This project has been carried to train people and accumulate the knowledge and techniques related to the measurement of the profiles of ion temperature, toroidal rotation velocity, and fully-stripped ion density in a fusion tokamak plasma by the development of plasma diagnostics using charge exchange recombination (CER) spectroscopy. Daring the 1 st year, the basic study and review on the charge exchange process and the conceptual design and review of the diagnostics have been conducted. In addition, the various atomic data centers around the world have been surveyed and atomic data related to CER have been constructed. The results of this project can be used to the construction and tokamak machine installation of a CER plasma diagnostic to a new superconducting supported by National Fusion Program. 42 refs., 3 tabs., 16 figs. (author)

  15. A simple in-vessel/FW component viewing system for SST-1

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Biswas, Prabal; Vasava, Kirit R.; Jaiswal, Snehal; Parekh, Tejas; Chauhan, Pradeep; Patel, Hiteshkumar; Pradhan, Subrata

    2015-01-01

    A simple compact system is being proposed for in-situ visual inspection of around 3800 First Wall (FW) graphite (armour) tiles in the vacuum vessel of SST-1 tokamak. The 2 DOF, manual driven system (permanently stationed inside vacuum vessel behind outer passive stabilizer) at top and bottom mid-plane locations consist of a rack and pinion mechanism operating a arm with a CCD camera/LED mounted on it, moving over a cam profile to cover approximately 1/8 th of the toroidal span of the vacuum vessel both at interior top/bottom locations with in the FW modules. The camera and LED light should withstand the ultrahigh vacuum conditions, prolonged baking temperatures of around 200°C along with high electromagnetic forces inside the vessel. This system can be operated remotely in-between shots from outside the VV through a linear motion feed through providing linear moment to a rack and pinion mechanism connected to the arm. This mechanism provides a better viewing of the inside FW components and vessel wall surface of tokamak with simple engineering and operational effort. Any information can be acquired from system regarding damages to FWC due to interaction with plasma as well as damage of other support structures inside VV. In comparison to more complicated and complex inspection system used in other tokamaks, this mechanism can be used for frequent in vessel visual inspection, which limits the system to be small, simple, occupying less space and custom made. This system is cheap with a minimum time for realization of the concept. The paper will present the conceptual and engineering design aspect of the in-viewing system, CAD images, its advantages and limitations, camera and LED details, data acquisition and the present status of realization of the project. (author)

  16. EDICAM fast video diagnostic installation on the COMPASS tokamak

    International Nuclear Information System (INIS)

    Szappanos, A.; Berta, M.; Hron, M.; Panek, R.; Stoeckel, J.; Tulipan, S.; Veres, G.; Weinzettl, V.; Zoletnik, S.

    2010-01-01

    A new camera system 'event detection intelligent camera' (EDICAM) is being developed by the Hungarian Association and has been installed on the COMPASS tokamak in the Institute of Plasma Physics AS CR in Prague, during February 2009. The standalone system contains a data acquisition PC and a prototype sensor module of EDICAM. Appropriate optical system have been designed and adjusted for the local requirements, and a mechanical holder keeps the camera out of the magnetic field. The fast camera contains a monochrome CMOS sensor with advanced control features and spectral sensitivity in the visible range. A special web based control interface has been implemented using Java spring framework to provide the control features in a graphical user environment. Java native interface (JNI) is used to reach the driver functions and to collect the data stored by direct memory access (DMA). Using a built in real-time streaming server one can see the live video from the camera through any web browser in the intranet. The live video is distributed in a Motion Jpeg format using real-time streaming protocol (RTSP) and a Java applet have been written to show the movie on the client side. The control system contains basic image processing features and the 3D wireframe of the tokamak can be projected to the selected frames. A MatLab interface is also presented with advanced post processing and analysis features to make the raw data available for high level computing programs. In this contribution all the concepts of EDICAM control center and the functions of the distinct software modules are described.

  17. Design and development of AXUV-based soft X-ray diagnostic camera for Aditya Tokamak

    International Nuclear Information System (INIS)

    Raval, Jayesh V.; Purohit, Shishir; Joisa, Y. Shankara

    2015-01-01

    The hot tokamak plasma emits Soft X-rays (SXR) in accordance with the temperature and density which are important to be studied. A silicon photo diode array (AXUV16ELG, Opto-diode, USA) based prototype SXR diagnostics is designed and developed for ADITYA tokamak for the study of SXR radial intensity profile, internal disruption (Saw-tooth crash), MHD instabilities. The diagnostic is having an array of 16 detector of millimeter dimension in a linear configuration. Absolute Extreme Ultra Violate (AXUV) detector offers compact size, improved time response with considerably good quantum efficiency in the soft X-ray range (200 eV to 10 keV). The diagnostic is designed in competence with the ADITYA tokamak protocol. The diagnostic design geometry allows detector view the plasma through a slot hole (0.5 cm X 0.05 cm), 10 μm Beryllium foil filter window, cutting off energies below 750 eV. The diagnostic was installed on Aditya vacuum vessel at radial port no 7 enabling the diagnostics to view the core plasma. The spatial resolution designed for diagnostic configuration is 1.3 cm at plasma centre. The signal generated from SXR detector is acquired with a dedicated single board computer based data acquisition system at 50 kHz. The diagnostic took observation for the ohmically heated plasma. The data was then processed to construct spatial and temporal profile of SXR intensity for Aditya plasma. This information was complimentary to the Silicon surface barrier detector (SBD) based array for the same plasma discharge. The cross calibration between the two was considerably satisfactory under the assumptions considered. (author)

  18. Soft x-ray virtual diagnostics for tokamak simulations

    Science.gov (United States)

    Kim, J. S.; Zhao, L.; Bogatu, I. N.; In, Y.; Turnbull, A.; Osborne, T.; Maraschek, M.; Comer, K.

    2009-11-01

    The numerical toolset, FAR-TECH Virtual Diagnostic Utility, for generating virtual experimental data based on theoretical models and comparing it with experimental data, has been developed for soft x-ray diagnostics on DIII-D. The virtual (or synthetic) soft x-ray signals for a sample DIII-D discharge are compared with the experimental data. The plasma density and temperature radial profiles needed in the soft x-ray signal modeling are obtained from experimental data, i.e., from Thomson scattering and electron cyclotron emission. The virtual soft x-ray diagnostics for the equilibriums have a good agreement with the experimental data. The virtual diagnostics based on an ideal linear instability also agree reasonably well with the experimental data. The agreements are good enough to justify the methodology presented here for utilizing virtual diagnostics for routine comparison of experimental data. The agreements also motivate further detailed simulations with improved physical models such as the nonideal magnetohydrodynamics contributions (resistivity, viscosity, nonaxisymmetric error fields, etc.) and other nonlinear effects, which can be tested by virtual diagnostics with various stability modeling.

  19. Soft x-ray virtual diagnostics for tokamak simulations

    International Nuclear Information System (INIS)

    Kim, J. S.; Zhao, L.; Bogatu, I. N.; In, Y.; Turnbull, A.; Osborne, T.; Maraschek, M.; Comer, K.

    2009-01-01

    The numerical toolset, FAR-TECH Virtual Diagnostic Utility, for generating virtual experimental data based on theoretical models and comparing it with experimental data, has been developed for soft x-ray diagnostics on DIII-D. The virtual (or synthetic) soft x-ray signals for a sample DIII-D discharge are compared with the experimental data. The plasma density and temperature radial profiles needed in the soft x-ray signal modeling are obtained from experimental data, i.e., from Thomson scattering and electron cyclotron emission. The virtual soft x-ray diagnostics for the equilibriums have a good agreement with the experimental data. The virtual diagnostics based on an ideal linear instability also agree reasonably well with the experimental data. The agreements are good enough to justify the methodology presented here for utilizing virtual diagnostics for routine comparison of experimental data. The agreements also motivate further detailed simulations with improved physical models such as the nonideal magnetohydrodynamics contributions (resistivity, viscosity, nonaxisymmetric error fields, etc.) and other nonlinear effects, which can be tested by virtual diagnostics with various stability modeling.

  20. Possibility of image converter application to X ray diagnostics in tokamaks

    International Nuclear Information System (INIS)

    Bryzgunov, V.A.; Chuvatin, S.A.

    1980-01-01

    To improve the sensitivity of spectral roentgen diagnostics of tokamak plasma possibilities for using a light amplifier having a PMU-1 microchannel plate for photorecording of roentgen spectra have been estimated. Amplification homogeneity over an image field has been improved 5-10 times by means of a screen. In bench tests spectra with an exposure 100 times lesser as compared with the RT-1 roentgen film have been taken. Five-layer protective screen has been developed for experiments in scattered magnetic fields of the tokamak. It is concluded that from the point of view of a number of parameters the PMU-1 electron optical converter quite meets the requirements of roentgen diagnostics. It has a sufficient amplification, small dark noises, small distortion. However, it is very desirable to have an instrument of a better resolution as compared with an obtained one as well as having a more light and comfortable magnetic protection

  1. Laser cleaning of diagnostic mirrors from tokamak-like carbon contaminants

    Energy Technology Data Exchange (ETDEWEB)

    Maffini, A., E-mail: alessandro.maffini@polimi.it [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Uccello, A. [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Dellasega, D. [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Istituto di Fisica del Plasma, Consiglio Nazionale delle Ricerche, EURATOM-ENEA-CNR Association, Milan (Italy); Russo, V. [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Perissinotto, S. [Center for Nano Science and Technology @ Polimi, Istituto Italiano di Tecnologia, Milan (Italy); Passoni, M. [Dipartimento di Energia, Politecnico di Milano, Milan (Italy); Istituto di Fisica del Plasma, Consiglio Nazionale delle Ricerche, EURATOM-ENEA-CNR Association, Milan (Italy)

    2015-08-15

    This paper presents a laboratory-scale experimental investigation of laser cleaning of diagnostic First Mirrors (FMs). Redeposition of contaminants sputtered from tokamak first wall onto FMs surface could dramatically decrease their reflectivity in an unacceptable way for the functioning of the plasma diagnostic systems. Laser cleaning is a promising solution to tackle this issue. In this work, pulsed laser deposition was exploited to produce rhodium films functional as FMs and to deposit onto them carbon contaminants with tailored features, resembling those found in tokamaks. The same laser system was also used to perform laser cleaning experiments by means of a sample handling procedure that allows to clean some cm{sup 2} in few minutes. The cleaning effectiveness was evaluated in terms of specular reflectivity recovery and mirror surface integrity. The effect of different laser wavelengths (λ = 1064, 266 nm) on the cleaning process is also addressed.

  2. Design of new Thomson scattering diagnostic system on COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Bílková, Petra; Aftanas, Milan; Böhm, Petr; Weinzettl, Vladimír; Šesták, David; Melich, Radek; Stöckel, Jan; Scannell, R.; Walsh, M.

    2010-01-01

    Roč. 623, č. 2 (2010), s. 656-659 ISSN 0168-9002. [International Conference on Frontiers in Diagnostic Technologies/1st./. Frascati, 25.11.2009-27.11.2009] R&D Projects: GA ČR GA202/09/1467; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Thomson scattering * Laser diagnostic * Electron temperature * Electron density Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.142, year: 2010 www.elsevier.com/locate/nima

  3. A diagnostic for electron dynamics in tokamaks. Final report

    International Nuclear Information System (INIS)

    Skiff, F.; Boyd, D.

    1996-12-01

    The diagnostic was installed on TdeV and brought into operation. It was optimized to the extent that time and money permitted. A considerable quantity of data was accumulated and analyzed. Experiments ended in August 1995. The apparatus has been removed from TdeV and returned to the University of Maryland. Each of these activities is detailed in this report

  4. A diagnostic for electron dynamics in tokamaks. Final report

    International Nuclear Information System (INIS)

    Skiff, F.; Boyd, D.

    1997-12-01

    The diagnostic was installed on TdeV and brought into operation. It was optimized to the extent that time and money permitted. A considerable quantity of data was accumulated and analyzed. Experiments ended in August 1995. The apparatus has been removed from TdeV and returned to the University of Maryland. Each of these activities is detailed here. The diagnostic worked very well. Although the distribution functions behaved in ways that were not anticipated and the refractive losses were sometimes higher than projected, the authors were able to adapt to the unexpected. In the authors' estimation, all of the effects listed above are significant, and warrant further study. The diagnostic is ready for use as a tool to study the physics of current drive and current profile modification. A mechanism for steering the launched beams is desirable to cope with the strong variations in refraction which are seen. Phased array launchers seem attractive for this purpose. Tuning of the length of the waveguide run is important to avoid troublesome reflections (return losses). It may be best to build in this capability in a future system. The perpendicular dynamics of the current driven electrons are invisible to us with the present form of the diagnostic. Simultaneous measurements at fundamental and harmonic frequencies would make perpendicular distribution function measurements possible

  5. Edge Thomson scattering diagnostic on COMPASS tokamak: Installation, calibration,operation, improvements

    Czech Academy of Sciences Publication Activity Database

    Böhm, Petr; Aftanas, Milan; Bílková, Petra; Štefániková, Estera; Mikulín, Ondřej; Melich, Radek; Janky, Filip; Havlíček, Josef; Šesták, David; Weinzettl, Vladimír; Stöckel, Jan; Hron, Martin; Pánek, Radomír; Scannell, R.; Frassinetti, L.; Fassina, A.; Naylor, G.; Walsh, M.J.

    2014-01-01

    Roč. 85, č. 11 (2014), 11E431-11E431 ISSN 0034-6748. [Topical Conference on High-Temperature Plasma Diagnostics/20./. Atlanta, Georgia, 01.06.2014-05.06.2014] R&D Projects: GA MŠk(CZ) LM2011021; GA ČR(CZ) GA14-35260S Institutional support: RVO:61389021 Keywords : plasma * tokamak * pedestal * Thomson scattering * diagnostic Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.614, year: 2014 http://scitation.aip.org/content/aip/journal/rsi/85/11/10.1063/1.4893995

  6. Electron cyclotron-electron Bernstein wave emission diagnostics for the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Zajac, Jaromír; Preinhaelter, Josef; Urban, Jakub; Žáček, František; Šesták, David; Nanobashvili, S.

    2010-01-01

    Roč. 81, č. 10 (2010), 10D911-10D911 ISSN 0034-6748. [TOPICAL CONFERENCE ON HIGH-TEMPERATURE PLASMA DIAGNOSTICS/18th./. Wildwood, New Jersey, 16.05.2010-20.05.2010] R&D Projects: GA ČR GA202/08/0419 Institutional research plan: CEZ:AV0Z20430508 Keywords : antenna radiation patterns * antennas in plasma * plasma diagnostics * Tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.598, year: 2010 http://link.aip.org/link/?RSI/81/10D911

  7. Distribution of the In-Vessel Diagnostics in ITER Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    González, Jorge, E-mail: Jorge.Gonzalez@iter.org [Rüecker Lypsa, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Clough, Matthew; Martin, Alex; Woods, Nick; Suarez, Alejandro [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France); Martinez, Gonzalo [Technical University Of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Stefan, Gicquel; Yunxing, Ma [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France)

    2017-01-15

    The ITER In-Vessel Diagnostics have been distributed around the In-Vessel shell to understand burning plasma physics and assist in machine operation. Each diagnostics component has its own requirements, constraints, and even exclusion among them for the highly complex In-Vessel environment. The size of the plasma, the requirement to be able to align the blanket system to the magnetic centre of the machine, the cooling requirements of the blanket system and the size of the pressure vessel itself all add to the difficulties of integrating these systems into the remaining space available. The available space for the cables inside the special trays (in-Vessel looms) is another constraint to allocate In-Vessel electrical sensors. Besides this, there are issues with the Assembly sequences and surface & volumetric neutron heating considerations that have imposed several additional restrictions.

  8. Remote handling installation of diagnostics in the JET Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Allan, P., E-mail: Peter.Allan@ccfe.ac.uk [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Loving, A.B. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Omran, H. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Oxford Technologies Ltd, 7 Nuffield Way, Abingdon OX14 1RJ (United Kingdom); Collins, S.; Thomas, J. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Parsloe, A.; Merrigan, I. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); British Nuclear Services, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Hassall, I. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Oxford Technologies Ltd, 7 Nuffield Way, Abingdon OX14 1RJ (United Kingdom)

    2011-10-15

    The requirement for an upgrade of the diagnostics for the JET ITER Like Wall (ILW) while maintaining personnel exposure to contamination as low as reasonably practicable or ALARP, has necessitated the development of a bespoke set of diagnostic components. These components, by virtue of their design and location, require a versatile yet comprehensive suite of remote handling tools to undertake their in-vessel installation. The installation of the various diagnostic components is covered in multiple tasks. Each task requires careful assessment and design of tools that can successfully interface with the components and comply with the handling and installation requirements. With remote maintenance a requirement, the looms/conduits were designed to be modular with connections which are electrically connected when the module is fitted or conversely disconnected when removed. The shape of each complex and often bulky component is verified during the design phase, to ensure that it can be delivered and installed to its specified location in the torus. This is done by matching the kinematic capabilities of the remote handling system and the path of the component through the torus by using a state of the art virtual reality system.

  9. Control system for 5 MW neutral beam ion source for SST1

    Science.gov (United States)

    Patel, G. B.; Onali, Raja; Sharma, Vivek; Suresh, S.; Tripathi, V.; Bandyopadhyay, M.; Singh, N. P.; Thakkar, Dipal; Gupta, L. N.; Singh, M. J.; Patel, P. J.; Chakraborty, A. K.; Baruah, U. K.; Mattoo, S. K.

    2006-01-01

    This article describes the control system for a 5MW ion source of the NBI (neutral beam injector) for steady-state superconducting tokamak-1 (SST-1). The system uses both hardware and software solutions. It comprises a DAS (data acquisition system) and a control system. The DAS is used to read the voltage and current signals from eight filament heater power supplies and 24 discharge power supplies. The control system is used to adjust the filament heater current in order to achieve an effective control on the discharge current in the plasma box. The system consists of a VME (Verse Module Eurocard) system and C application program running on a VxWorks™ real-time operating system. A PID (proportional, integral, and differential) algorithm is used to control the filament heater current. Experiments using this system have shown that the discharge current can be controlled within 1% accuracy for a PID loop time of 20ms. Response of the control system to the pressure variation of the gas in the chamber has also been studied and compared with the results obtained from those of an uncontrolled system. The present approach increases the flexibility of the control system. It not only eases the control of the plasma but also allows an easy changeover to various operation scenarios.

  10. Fabrication of new joints for SST-1 TF coil winding packs

    International Nuclear Information System (INIS)

    Prasad, Upendra; Sharma, A.N.; Patel, Dipak; Doshi, Kalpesh; Khristi, Yohan; Varmora, Pankaj; Chauhan, Pradeep; Jadeja, S.J.; Gupta, Pratibha; Pradhan, S.

    2013-01-01

    Highlights: • We have carried out work related with sub-nanoohm joints for superconducting Tokamak winding packs. • We have established fine tune QA/QC procedures for sub-nanoohm joints fabrication. • We have optimised welding parameters for cable in conduit conductors for fusion relevant magnets. • We have established precised measurement data acquisition system for low resistance measurements at cryogenic temperature. -- Abstract: The Toroidal Field (TF) magnet system of SST-1 has sixteen NbTi/Cu based coils with about one hundred Inter-Pancake (IP) and Inter-Coil (IC) joints. New box type helium leak tight, low DC resistance joints have been designed, fabricated and tested at 5 K temperature and 10 kA DC transport current. The prototype of this joint has been validated in laboratory as well as on spare TF coil winding pack. Moreover, the performance of these joints has been realised and validated on actual sixteen TF winding packs, the joint resistance of ∼0.5 nΩ repeatedly measured on hundreds of IP joints. The quality of terminations and joints was ensured at various stages of fabrication. The quality of joint box material was ensured by visual inspection, chemical analysis, radiography test, ultrasonic test, eddy current test, etc. This paper describes joint design drivers, joint design detail, prototype joint fabrication processes, quality assurance (QA)/quality control (QC) adopted during prototype and actual joint fabrication process, joint resistance measurement on actual TF coils and analysis of measured joint resistance in detail

  11. Operational experience of SST1 NBI control system with prototype Ion source

    International Nuclear Information System (INIS)

    Patel, V B; Patel, P J; Singh, N P; Tripathi, V; Thakkar, D; Gupta, L N; Prahlad, V; Sharma, S K; Bandyopadyay, M; Chakraborty, A K; Baruah, U K; Mattoo, S K; Patel, G B; Onali, Raja

    2010-01-01

    This paper presents operational experience of integrated control of the arc-filament and High-voltage power supply of Steady State Tokamak (SST)-1 NBI system using Versa Module Europa (VME) system on prototype Ion source. The control algorithm is implemented on the VxWorks operating system using 'C' language. This paper also describes the operating sequence and controls on power supply system. Discharge and Filament power supplies are controlled in such a way so that necessary discharge current can be available in Ion Source. The discharge current is controlled by manipulating the filament current. Close loop control is implemented on each filament power supply with feedback from Discharge Current to control the overall discharge inside the ion source. Necessary actions for shut OFF and subsequent Turn ON are also taken during breakdowns between the Grids of the ion source. Total numbers of breakdowns are also monitored. Shot is terminated, if the breakdown count is higher than the set value. This control system can be programmed to restart High-voltage power supply within 5mS after breakdown occurs. This control system is capable to handle the all types of dynamics in the system. This paper also presents results of experiment.

  12. Data acquisition and control system for the ECE imaging diagnostic on the EAST tokamak

    Science.gov (United States)

    Luo, C.; Lan, T.; Zhu, Y.; Xie, J.; Gao, B.; Liu, W.; Yu, C.; Milne, P. G.; Domier, C. W.; Luhmann, N. C.

    2017-06-01

    A 384-channel electron cyclotron emission imaging (ECEI) system is installed on the experimental advanced superconducting tokamak (EAST) and 7-gigabyte data is produced for each regular discharge of a 10-second pulse. The data acquisition and control (DAC) system for the EAST ECEI diagnostics covers the large data production and embeds the ability to report the data quality instantly after the discharge. The symmetric routing design of the timing signal distributions among the 384 channels provides a low-cost solution to the synchronization of a large number of channels. The application of the load-balance bond service largely reduces the configuration difficulty and the cost in the high-speed data transferring tasks. Benefiting from the various kinds of hardware units with dedicated functionalities, an automated and user interactive DAC work flow is achieved, including the pre-selections of the automation scheme and the observation region, 384-channel data acquisition and local caching, post-discharge imaging data quality evaluation, remote system status monitoring, and inter-discharge imaging system event handling. The system configuration in a specific physics experiment is further optimized through the associated operating software which is enhanced by the input of the tokamak operation status and the region of interest (ROI) from other diagnostics. The DAC system is based on a modularized design and scalable to the long-pulse discharges in the EAST tokamak.

  13. Data acquisition and control system for the ECE imaging diagnostic on the EAST tokamak

    International Nuclear Information System (INIS)

    Luo, C.; Lan, T.; Xie, J.; Gao, B.; Liu, W.; Yu, C.; Zhu, Y.; Domier, C.W.; Luhmann, N.C.; Milne, P.G.

    2017-01-01

    A 384-channel electron cyclotron emission imaging (ECEI) system is installed on the experimental advanced superconducting tokamak (EAST) and 7-gigabyte data is produced for each regular discharge of a 10-second pulse. The data acquisition and control (DAC) system for the EAST ECEI diagnostics covers the large data production and embeds the ability to report the data quality instantly after the discharge. The symmetric routing design of the timing signal distributions among the 384 channels provides a low-cost solution to the synchronization of a large number of channels. The application of the load-balance bond service largely reduces the configuration difficulty and the cost in the high-speed data transferring tasks. Benefiting from the various kinds of hardware units with dedicated functionalities, an automated and user interactive DAC work flow is achieved, including the pre-selections of the automation scheme and the observation region, 384-channel data acquisition and local caching, post-discharge imaging data quality evaluation, remote system status monitoring, and inter-discharge imaging system event handling. The system configuration in a specific physics experiment is further optimized through the associated operating software which is enhanced by the input of the tokamak operation status and the region of interest (ROI) from other diagnostics. The DAC system is based on a modularized design and scalable to the long-pulse discharges in the EAST tokamak.

  14. Neutron emissivity profile camera diagnostics considering present and future tokamaks

    International Nuclear Information System (INIS)

    Forsberg, S.

    2001-12-01

    This thesis describes the neutron profile camera situated at JET. The profile camera is one of the most important neutron emission diagnostic devices operating at JET. It gives useful information of the total neutron yield rate but also about the neutron emissivity distribution. Data analysis was performed in order to compare three different calibration methods. The data was collected from the deuterium campaign, C4, in the beginning of 2001. The thesis also includes a section about the implication of a neutron profile camera for ITER, where the issue regarding interface difficulties is in focus. The ITER JCT (Joint Central Team) proposal of a neutron camera for ITER is studied in some detail

  15. Neutron emissivity profile camera diagnostics considering present and future tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, S. [EURATOM-VR Association, Uppsala (Sweden)

    2001-12-01

    This thesis describes the neutron profile camera situated at JET. The profile camera is one of the most important neutron emission diagnostic devices operating at JET. It gives useful information of the total neutron yield rate but also about the neutron emissivity distribution. Data analysis was performed in order to compare three different calibration methods. The data was collected from the deuterium campaign, C4, in the beginning of 2001. The thesis also includes a section about the implication of a neutron profile camera for ITER, where the issue regarding interface difficulties is in focus. The ITER JCT (Joint Central Team) proposal of a neutron camera for ITER is studied in some detail.

  16. Integration of cryopump instrumentation for SST-1 NBI

    International Nuclear Information System (INIS)

    Bansal, Laxmi Kant; Patel, Paresh J.; Prahlad, V.

    2015-01-01

    A positive ion neutral injector (PINI) capable in delivering 5MW (55kV, 90A) ion beam power is being operated for SST-1 neutral beam injection (NBI). The production and neutralization of the ion beams in the injector requires a gas throughput of 20 torr I/s in the plasma box and 50-100 torr I/s in the neutralizer section. It is necessary to maintain operating pressure of vessel at 10 -5 torr to reduce the re-ionization loss of beam within tolerable limits. Conventional Turbo molecular pumps cannot maintain this vacuum level at required gas feed rate so two cryo condensation pumps are being operated to achieve require vacuum in vessel. In order to monitor and optimize the performance of cryopumps, it is necessary to measure the temperature at various locations in LN 2 and LHe path. It is also required to monitor the level of LHe and LN 2 in cryopumps. Several temperature and level sensors are mounted at various places in cryopumps and integrated with PLC and SCADA based control system. This paper presents the details of sensor mounting, signals conditioning, scheme of their integration with PLC and SCADA and results in detail. (author)

  17. Technological start of T-15 tokamak. The start-up diagnostic complex

    International Nuclear Information System (INIS)

    Notkin, G.E.

    1989-01-01

    The T-15 tokamak with superconducting toroidal winding reached the technological start-up phase. The results of the first operating tests of the main tokamak components are reported. Due to improper function of both the vacuum and the cryogenic system, the nominal parameters of the vacuum and of the toroidal magnetic field have not been achieved. The non-optimum vacuum conditions made the discharge start-up difficult even when a pre-ionizing electron beam and a gyrotron generator were used. The pre-discharge plasma parametes were studied by means of a limited set of plasma diagnostic apparatus. Due to substantially deteriorated vacuum conditions, it was not possible to repeat the only one successful discharge with a current of 100 kA, lasting for 50 ms. (J.U.)

  18. MTX [Microwave Tokamak Experiment] diagnostic and auxiliary systems for confinement, transport, and plasma physics studies

    International Nuclear Information System (INIS)

    Hooper, E.B.; Allen, S.L.; Casper, T.A.; Thomassen, K.I.

    1989-01-01

    This note describes the diagnostics and auxiliary systems on the Microwave Tokamak Experiment (MTX) for confinement, transport, and other plasma physics studies. It is intended as a reference on the installed and planned hardware on the machine for those who need more familiarity with this equipment. Combined with the tokamak itself, these systems define the opportunities and capabilities for experiments in the MTX facility. We also illustrate how these instruments and equipment are to be used in carrying out the MTX Operations Plan. Near term goals for MTX are focussed on the absorption and heating by the microwave beam from the FEL, but the Plan also includes using the facility to study fundamental phenomena in the plasma, to control MHD activity, and to drive current noninductively

  19. EDICAM fast video diagnostic installation on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Szappanos, A.; Berta, M.; Hron, Martin; Pánek, Radomír; Stöckel, Jan; Veres, G.; Weinzettl, Vladimír; Zoletnik, S.; Tulipán, S.

    2010-01-01

    Roč. 85, 3-4 (2010), s. 370-373 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition and Remote Participation for Fusion Research/7th./. Aix – en – Provence, 15.06.2009-19.06.2009] Institutional research plan: CEZ:AV0Z20430508 Keywords : Video diagnostic * Fast data processing * CMOS sensor * Image processing Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.143, year: 2010 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6V3C-4Y0C2FK-1&_user=6542793&_coverDate=07%2F31%2F2010&_rdoc=1&_fmt=high&_orig=search&_origin=search&_sort=d&_docanchor=&view=c&_acct=C000070123&_version=1&_urlVersion=0&_userid=6542793&md5=99eb6704be38e61ac7e2316cb63a7ee9&searchtype=a

  20. Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

    Science.gov (United States)

    Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

    2017-12-01

    The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

  1. The interpretation of tokamak magnetic diagnostics: Status and prospects

    International Nuclear Information System (INIS)

    Braams, B.J.

    1985-09-01

    The analytical theory and the computational methods that are available for the determination of MHD equilibrium characteristics from magnetic measurements on axisymmetric systems are reviewed and developed. The interpretation of these measurements relies to a large extent on two classes of integral relations due to L.E. Zakharov and V.D. Shafranov. Following and extending their work we provide an inventory of useful integral relations, including the contributions due to pressure anisotropy and plasma rotation. Effective methods to evaluate the required integrals from imperfect measurements are considered. A full equilibrium analysis of the magnetic diagnostics implies a determination of the current profile, consistent with the equations of MHD equilibrium, aiming at an optimal fit between the corresponding calculated magnetic field and the measured data. Published approaches to this problem are evaluated, and a novel fast algorithm is proposed. Instead of the full equilibrium analysis several more limited problems are often considered, for which faster methods are available. The determination of only the plasma boundary requires the solution of a Cauchy problem, or similar, for an elliptic equation. The published approaches are compared. Analytical theory provides approximations that are suitable for the rapid estimation of characteristic parameters, related to the plasma current, position, shape of the cross-section, pressure and internal inductance. Very efficient algorithms may be obtained when this theory is employed in conjunction with the method of function parametrization. These algorithms are well suited to real-time control of the plasma. Appendices present a discussion of the boundary conditions for the MHD equilibrium problem, a compendium of analytical solutions to the homogeneous equilibrium equation, and a re-examination of the possibility of determining the current distribution from knowledge of only the shape of the flux surfaces. (orig.)

  2. The diagnostic neutral beam injector with arc-discharge plasma source on the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Karpushov, Alexander N. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland)], E-mail: alexander.karpushov@epfl.ch; Andrebe, Yanis; Duval, Basil P.; Bortolon, Alessandro [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland)

    2009-06-15

    The diagnostic neutral beam injector (DNBI) together with a charge exchange recombination spectroscopy (CXRS) system has been used on the TCV Tokamak as a diagnostic tool for local measurements of plasma ion temperature, velocity and carbon impurity density based on analysis of the beam induced impurity radiation emission since 2000. To improve the performance of the CXRS diagnostic, several upgrades of both the optical system and the neutral beam were performed. An increase of the plasma source size together with beam optimization in 2003 resulted in a twofold increase the beam current. The RF plasma generator was replaced by an arc-discharge plasma source together with a new ion optical system (IOS) in 2006 and subsequent beam optimization is presented herein. This was designed to increase the line brightness of the beam in the CXRS observation region without increasing of the injected power (to avoid plasma perturbation by the beam). The beam characteristics are measured by a multi-chord scanning of Doppler-shifted H{sub {alpha}} emission, thermal measurements on a movable calorimeter and visible optical measurements inside the Tokamak vessel.

  3. Mechanical design and thermo-hydraulic simulation of the infrared thermography diagnostic of the WEST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Micolon, Frédéric, E-mail: frederic.micolon@cea.fr; Courtois, Xavier; Aumeunier, Marie-Hélène; Chenevois, Jean-Pierre; Larroque, Sébastien

    2015-10-15

    The WEST (Tungsten (W) Environment in Steady state Tokamak) project is a partial rebuild of the Tore Supra tokamak to make it an X-point metallic environment machine aimed at testing ITER technologies in relevant plasma environment. For the safe operation of the WEST tokamak, infra-red (IR) thermography is a crucial diagnostic as it is a sound and reliable way to detect hotspots or abnormal heating patterns on the plasma facing components (PFCs). Thus WEST will be fitted with middle/short-IR (1.5–2 μm or 3–5 μm) cameras in the upper port plugs to get a full view of the critical PFCs (in particular the new lower divertor) and radio-frequency (RF) heating antennas and one camera at the equatorial level to monitor the new upper divertor and the first wall. This paper describes the design of the up-to-date optical system along with the hydraulic analysis and the thermal and mechanical finite element analysis conducted to ensure adequate heat extraction capabilities. Boundary conditions and simulation results will be presented and discussed as well as technological solutions retained.

  4. Timing and triggering of the Thomson scattering diagnostics on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Mikulín, Ondřej; Hron, Martin; Böhm, Petr; Naylor, G.; Bílková, Petra; Janky, Filip; Salášek, J.; Pánek, Radomír

    2014-01-01

    Roč. 89, č. 5 (2014), s. 693-697 ISSN 0920-3796. [The 9th Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/9./. Hefei, 06.05.2013-10.05.2013] R&D Projects: GA MŠk 7G10072; GA MŠk(CZ) LM2011021; GA ČR GAP205/11/2470 Institutional support: RVO:61389021 Keywords : Tokamak * Timing and triggering * FPGA * Real-time control * Diagnostics control * Thomson scattering Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.152, year: 2014 http://www.sciencedirect.com/science/article/pii/S0920379614002105#

  5. Performance and analysis of the TVTS diagnostic system on HT-7 tokamak

    International Nuclear Information System (INIS)

    Han Xiaofeng; Shao Chunqiang; Xi Xiaoqi; Zhao Junyu; Qing Zang; Yang Jianhua; Dai Xingxing

    2013-01-01

    A high spatial resolution imaging Thomson scattering diagnostic system was developed in ASIPP. After about one month trial running on the superconducting HT-7 tokamak, the system was proved to be capable of measuring plasma electron temperature. The system setup and data calibration are described in this paper and then the instrument function is studied in detail, as well as the measurement capability, an electron temperature of 50 eV to 2 keV and density beyond 1x10"1"9 m"-"3. Finally, the data processing method and experimental results are presented. (author)

  6. Selected methods of electron-and ion-diagnostics in tokamak scrape-off-layer

    Directory of Open Access Journals (Sweden)

    Sadowski Marek J.

    2015-06-01

    Full Text Available This invited paper considers reasons why exact measurements of fast electron and ion losses in tokamaks, and particularly i n a scrape-off-layer and near a divertor region, are necessary in order to master nuclear fusion energy production. Attention is also paid to direct measurements of escaping fusion products from D-D and D-T reactions, and in particular of fast alphas which might be used for plasma heating. The second part describes the generation of so-called runaway and ripple-born electrons which might induce high energy losses and cause severe damages of internal walls in fusion facilities. Advantages and disadvantages of different diagnostic methods applied for studies of such fast electrons are discussed. Particular attention is paid to development of a direct measuring technique based on the Cherenkov effect which might be induced by fast electrons in appropriate radiators. There are presented various versions of Cherenkov-type probes which have been developed by the NCBJ team and applied in different tokamak experiments. The third part is devoted to direct measurements of fast ions (including those produced by the nuclear fusion reactions which can escape from a high-temperature plasma region. Investigation of fast fusion-produced protons from tokamak discharges is reported. New ion probes, which were developed by the NCBJ team, are also presented. For the first time there is given a detailed description of an ion pinhole camera, which enables irradiation of several nuclear track detectors during a single tokamak discharge, and a miniature Thomson-type mass-spectrometer, which can be used for ion measurements at plasma borders.

  7. High-resolution spectroscopy diagnostics for measuring impurity ion temperature and velocity on the COMPASS tokamak

    International Nuclear Information System (INIS)

    Weinzettl, Vladimir; Shukla, Gaurav; Ghosh, Joydeep; Melich, Radek; Panek, Radomir; Tomes, Matej; Imrisek, Martin; Naydenkova, Diana; Varju, Josef; Pereira, Tiago; Gomes, Rui; Abramovic, Ivana; Jaspers, Roger; Pisarik, Michael; Odstrcil, Tomas; Van Oost, Guido

    2015-01-01

    Highlights: • We built a new diagnostic of poloidal plasma rotation on the COMPASS tokamak. • Improvements in throughput via toroidal integration and fiber optimizations shown. • Poloidal rotation and ion temperature measured in L- and H-mode and during RMP. • Design and parameters of a new CXRS diagnostic for COMPASS are introduced. - Abstract: High-resolution spectroscopy is a powerful tool for the measurement of plasma rotation as well as ion temperature using the Doppler shift of the emitted spectral lines and their Doppler broadening, respectively. Both passive and active diagnostic variants for the COMPASS tokamak are introduced. The passive diagnostic focused on the C III lines at about 465 nm is utilized for the observation of the poloidal plasma rotation. The current set-up of the measuring system is described, including the intended high-throughput optics upgrade. Different options to increase the fiber collection area are mentioned, including a flower-like fiber bundle, and the use of micro-lenses or tapered fibers. Recent measurements of poloidal plasma rotation of the order of 0–6 km/s are shown. The design of the new active diagnostic using a deuterium heating beam and based on charge exchange recombination spectroscopy (C VI line at 529 nm) is introduced. The tool will provide both space (0.5–5 cm) and time (10 ms) resolved toroidal plasma rotation and ion temperature profiles. The results of the Simulation of Spectra code used to examine the feasibility of charge exchange measurements on COMPASS are shown and connected with a selection of the spectrometer coupled with the CCD camera.

  8. High-resolution spectroscopy diagnostics for measuring impurity ion temperature and velocity on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Weinzettl, Vladimir, E-mail: vwei@ipp.cas.cz [Institute of Plasma Physics ASCR, Prague (Czech Republic); Shukla, Gaurav [Institute of Plasma Physics ASCR, Prague (Czech Republic); Department of Applied Physics, Ghent University, Ghent (Belgium); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Ghosh, Joydeep [Institute for Plasma Research, Bhat, Gandhinagar (India); Melich, Radek; Panek, Radomir [Institute of Plasma Physics ASCR, Prague (Czech Republic); Tomes, Matej; Imrisek, Martin; Naydenkova, Diana [Institute of Plasma Physics ASCR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Varju, Josef [Institute of Plasma Physics ASCR, Prague (Czech Republic); Pereira, Tiago [Instituto de Plasmas e Fusão Nuclear, Lisboa (Portugal); Instituto Superior Técnico, Universidade de Lisboa, Lisboa (Portugal); Gomes, Rui [Instituto de Plasmas e Fusão Nuclear, Lisboa (Portugal); Abramovic, Ivana; Jaspers, Roger [Eindhoven University of Technology, Eindhoven (Netherlands); Pisarik, Michael [SQS Vlaknova optika a.s., Nova Paka (Czech Republic); Department of Electromagnetic Field, Faculty of Electrical Engineering, Czech Technical University in Prague (Czech Republic); Odstrcil, Tomas [Max-Planck-Institut fur Plasmaphysik, Garching (Germany); Van Oost, Guido [Department of Applied Physics, Ghent University, Ghent (Belgium)

    2015-10-15

    Highlights: • We built a new diagnostic of poloidal plasma rotation on the COMPASS tokamak. • Improvements in throughput via toroidal integration and fiber optimizations shown. • Poloidal rotation and ion temperature measured in L- and H-mode and during RMP. • Design and parameters of a new CXRS diagnostic for COMPASS are introduced. - Abstract: High-resolution spectroscopy is a powerful tool for the measurement of plasma rotation as well as ion temperature using the Doppler shift of the emitted spectral lines and their Doppler broadening, respectively. Both passive and active diagnostic variants for the COMPASS tokamak are introduced. The passive diagnostic focused on the C III lines at about 465 nm is utilized for the observation of the poloidal plasma rotation. The current set-up of the measuring system is described, including the intended high-throughput optics upgrade. Different options to increase the fiber collection area are mentioned, including a flower-like fiber bundle, and the use of micro-lenses or tapered fibers. Recent measurements of poloidal plasma rotation of the order of 0–6 km/s are shown. The design of the new active diagnostic using a deuterium heating beam and based on charge exchange recombination spectroscopy (C VI line at 529 nm) is introduced. The tool will provide both space (0.5–5 cm) and time (10 ms) resolved toroidal plasma rotation and ion temperature profiles. The results of the Simulation of Spectra code used to examine the feasibility of charge exchange measurements on COMPASS are shown and connected with a selection of the spectrometer coupled with the CCD camera.

  9. Ultrafast two-dimensional lithium beam emission spectroscopy diagnostic on the EAST tokamak

    Science.gov (United States)

    Zoletnik, S.; Hu, G. H.; Tál, B.; Dunai, D.; Anda, G.; Asztalos, O.; Pokol, G. I.; Kálvin, S.; Németh, J.; Krizsanóczi, T.

    2018-06-01

    A diagnostic instrument is described for the Experimental Advanced Superconducting Tokamak (EAST) for the measurement of the edge plasma electron density profile and plasma turbulence properties. An accelerated neutral lithium beam is injected into the tokamak and the Doppler shifted 670.8 nm light emission of the Li2p-2s transition is detected. A novel compact setup is used, where the beam injection and observation take place from the same equatorial diagnostic port and radial-poloidal resolution is achieved with microsecond time resolution. The observation direction is optimized in order to achieve a sufficient Doppler shift of the beam light to be able to separate from the strong edge lithium line emission on this lithium coated device. A 250 kHz beam chopping technique is also demonstrated for the removal of background light. First results show the capability of measuring turbulence and its poloidal flow velocity in the scrape-off layer and edge region and the resolution of details of transient phenomena like edge localized modes with few microsecond time resolution.

  10. Design and implementation of data acquisition system for magnets of SST-1

    Energy Technology Data Exchange (ETDEWEB)

    Doshi, K., E-mail: pushpuk@ipr.res.in; Pradhan, S.; Masand, H.; Khristi, Y.; Dhongde, J.; Sharma, A.; Parghi, B.; Varmora, P.; Prasad, U.; Patel, D.

    2014-05-15

    The magnet system of the Steady-State Superconducting Tokamak-1 at the Institute for Plasma Research, Gandhinagar, India, consists of sixteen toroidal field and nine poloidal field. Superconducting coils together with a pair of resistive PF coils, an air core ohmic transformer and a pair of vertical field coils. These magnets are instrumented with various cryogenic compatible sensors and voltage taps for its monitoring, operation, protection, and control during different machine operational scenarios like cryogenic cool down, current charging cycles including ramp up, flat top, plasma breakdown, dumping/ramp down and warm up. The data acquisition system for these magnet instrumentation have stringent requirement regarding operational flexibility, reliability for continuous long term operation and data visualization during operations. A VME hardware based data acquisition system with ethernet based remote system architecture is implemented for data acquisition and control of the complete magnet operation. Software application is developed in three parts namely an embedded VME target, a network server and a remote client applications. A target board application implemented with real time operating system takes care of hardware configuration and raw data transmission to server application. A java server application manages several activities mainly multiple client communication over ethernet, database interface and data storage. A java based platform independent desktop client application is developed for online and offline data visualization, remote hard ware configuration and many other user interface tasks. The application has two modes of operation to cater to different needs of cool-down and charging operations. This paper describes application architecture, installation and commissioning and operational experience from the recent campaigns of SST-1.

  11. Millimeter-wave imaging diagnostics systems on the EAST tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Y. L.; Xie, J. L., E-mail: jlxie@ustc.edu.cn; Yu, C. X.; Zhao, Z. L.; Gao, B. X.; Chen, D. X.; Liu, W. D.; Liao, W.; Qu, C. M.; Luo, C. [School of Physics, University of Science and Technology of China, Anhui 230026 (China); Hu, X.; Spear, A. G.; Luhmann, N. C.; Domier, C. W.; Chen, M.; Ren, X. [University of California, Davis, California 95616 (United States); Tobias, B. J. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2016-11-15

    Millimeter-wave imaging diagnostics, with large poloidal span and wide radial range, have been developed on the EAST tokamak for visualization of 2D electron temperature and density fluctuations. A 384 channel (24 poloidal × 16 radial) Electron Cyclotron Emission Imaging (ECEI) system in F-band (90-140 GHz) was installed on the EAST tokamak in 2012 to provide 2D electron temperature fluctuation images with high spatial and temporal resolution. A co-located Microwave Imaging Reflectometry (MIR) will be installed for imaging of density fluctuations by December 2016. This “4th generation” MIR system has eight independent frequency illumination beams in W-band (75-110 GHz) driven by fast tuning synthesizers and active multipliers. Both of these advanced millimeter-wave imaging diagnostic systems have applied the latest techniques. A novel design philosophy “general optics structure” has been employed for the design of the ECEI and MIR receiver optics with large aperture. The extended radial and poloidal coverage of ECEI on EAST is made possible by innovations in the design of front-end optics. The front-end optical structures of the two imaging diagnostics, ECEI and MIR, have been integrated into a compact system, including the ECEI receiver and MIR transmitter and receiver. Two imaging systems share the same mid-plane port for simultaneous, co-located 2D fluctuation measurements of electron density and temperature. An intelligent remote-control is utilized in the MIR electronics systems to maintain focusing at the desired radial region even with density variations by remotely tuning the probe frequencies in about 200 μs. A similar intelligent technique has also been applied on the ECEI IF system, with remote configuration of the attenuations for each channel.

  12. Observation of short time-scale spectral emissions at millimeter wavelengths with the new CTS diagnostic on the FTU tokamak

    DEFF Research Database (Denmark)

    Bruschi, A.; Alessi, E.; Bin, W.

    2017-01-01

    On the FTU tokamak, the collective Thomson scattering (CTS) diagnostic was renewed for investigating the possible excitation of parametric decay instabilities (PDI) by electron cyclotron (EC) or CTS probe beams in presence of magnetic islands and measure their effects on the EC power absorption...

  13. RADAR upper hybrid resonance scattering diagnostics of small-scale fluctuations and waves in tokamak plasmas

    International Nuclear Information System (INIS)

    Bulyiginskiy, D.G.; Gurchenko, A.D.; Gusakov, E.Z.; Korkin, V.V.; Larionov, M.M.; Novik, K.M.; Petrov, Yu.V.; Popov, A.Yu.; Saveliev, A.N.; Selenin, V.L.; Stepanov, A.Yu.

    2001-01-01

    The upper hybrid resonance (UHR) scattering technique possessing such merits as one-dimensional probing geometry, enhancement of cross section, and fine localization of scattering region is modified in the new diagnostics under development to achieve wave number resolution. The fluctuation wave number is estimated in the new technique from the scattering signal time delay measurements. The feasibility of the scheme is checked in the proof of principal experiment in a tokamak. The time delay of the UHR scattering signal exceeding 10 ns is observed. The small scale low frequency density fluctuations are investigated in the UHR RADAR backscattering experiment. The UHR cross-polarization scattering signal related to small scale magnetic fluctuations is observed. The lower hybrid (LH) wave propagation and both linear and nonlinear wave conversion are investigated. The small wavelength (λ≤0.02 cm) high number ion Bernstein harmonics, resulting from the linear wave conversion of the LH wave are observed in a tokamak plasma for the first time

  14. Magnetic Diagnostics for Equilibrium Reconstructions in the Presence of Nonaxisymmetric Eddy Current Distributions in Tokamaks

    International Nuclear Information System (INIS)

    Kaita, R.; Kozub, T.; Logan, N.; Majeski, R.; Menard, J.; Zakharov, L.

    2010-01-01

    The lithium tokamak experiment (LTX) is a modest-sized spherical tokamak (R 0 = 0.4 m and a = 0.26 m) designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 C. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.

  15. Performance test results of ion beam transport for SST-1 neutral beam injector

    Energy Technology Data Exchange (ETDEWEB)

    Jana, M R; Mattoo, S K [Institute for Plasma Research Bhat, Gandhinagar-382428, Gujarat (India); Uhlemann, R, E-mail: mukti@ipr.res.i [Forschungszentrum Juelich, Institute fur Energieforschung IEF-4, Plasmaphysik D-52425 Juelich (Germany)

    2010-02-01

    A neutral beam injector is built at IPR to heat the plasma of SST-1 and its upgrade. It delivers a maximum beam power of 1.7 MW for 55 kV Hydrogen beam or 80 kV Deuterium beam. At lower beam voltage, the delivered power falls to 500 kW at 30 kV Hydrogen beam which is adequate to heat SST-1 plasma ions to {approx} 1 keV. Process of acceleration of ions to the required beam voltage, conversion of ions to neutrals and removal of un-neutralized ions and the beam diagnostic systems occupy a large space. The consequence is that linear extent of the neutral beam injector is at least a few meters. Also, port access provides a very narrow duct. Even a very good injector design and fabrication practices keep beam divergence at a very low but finite value. The result is beam transport becomes an important issue. Since a wide area beam is constructed by hundreds of beam lets, it becomes essential they be focused in such a way that beam transport loss is minimized. Horizontal and vertical focal lengths are two parameters, in addition to beam divergence, which give a description of the beam transport. We have obtained these two parameters for our injector by using beam transport code; making several hundred simulation runs by varying optical parameters of the beam. The selected parameters set has been translated into the engineering features of the extractor grid set of the ion source. Aperture displacement technique is used to secure the horizontal beam focusing at 5.4 m. Combination of both aperture displacement and inclining of two grid halves to {approx} 17 mrad are secured for vertical beam focusing at 7 m from earth grid of the ion source. The gaps between the design, engineered and performance tested values usually arise due to lack of exercising control over fabrication processes or due to inaccuracies in the assumption made in the model calculations of beam optics and beam transport. This has been the case with several injectors, notably with JET injector. To overcome

  16. Vertical one-dimensional electron cyclotron emission imaging diagnostic for HT-7 tokamak

    International Nuclear Information System (INIS)

    Wang Jun; Xu Xiaoyuan; Wen Yizhi; Yu Changxuan; Wan Baonian; Luhmann, N.C.; Wang, Jian; Xia, Z.G.

    2005-01-01

    A vertical resolved 16-channel electron cyclotron emission imaging (ECEI) diagnostic has been developed and installed on the HT7 Tokamak for measuring plasma electron cyclotron emission with a temporal resolution of 0.5 us. The system is working on a fixed frequency 97.5 GHz in the first stage. The sample volumes of the system are aligned vertically with a vertical channel spacing of 11 mm, and can be shifted across the plasma cross-section by varying the toroidal magnetic field. The high spatial resolution of the system is achieved by utilizing a low cost linear mixer/receiver array and an optical imaging system. The focus location may be shifted horizontally via translation of one of the optical imaging elements. The detail of the system design and laboratory testing of the ECE Imaging optics are presented, together with HT7 plasma data. (author)

  17. New detection system and signal processing for the tokamak ISTTOK heavy ion beam diagnostic

    International Nuclear Information System (INIS)

    Henriques, R. B.; Nedzelskiy, I. S.; Malaquias, A.; Fernandes, H.

    2012-01-01

    The tokamak ISTTOK heavy ion beam diagnostic (HIBD) operates with a multiple cell array detector (MCAD) that allows for the plasma density and the plasma density fluctuations measurements simultaneously at different sampling volumes across the plasma. To improve the capability of the plasma density fluctuations investigations, a new detection system and new signal conditioning amplifier have been designed and tested. The improvements in MCAD design are presented which allow for nearly complete suppression of the spurious plasma background signal by applying a biasing potential onto special electrodes incorporated into MCAD. The new low cost and small size transimpedance amplifiers are described with the parameters of 400 kHz, 10 7 V/A, 0.4 nA of RMS noise, adequate for the plasma density fluctuations measurements.

  18. New detection system and signal processing for the tokamak ISTTOK heavy ion beam diagnostic.

    Science.gov (United States)

    Henriques, R B; Nedzelskiy, I S; Malaquias, A; Fernandes, H

    2012-10-01

    The tokamak ISTTOK havy ion beam diagnostic (HIBD) operates with a multiple cell array detector (MCAD) that allows for the plasma density and the plasma density fluctuations measurements simultaneously at different sampling volumes across the plasma. To improve the capability of the plasma density fluctuations investigations, a new detection system and new signal conditioning amplifier have been designed and tested. The improvements in MCAD design are presented which allow for nearly complete suppression of the spurious plasma background signal by applying a biasing potential onto special electrodes incorporated into MCAD. The new low cost and small size transimpedance amplifiers are described with the parameters of 400 kHz, 10(7) V/A, 0.4 nA of RMS noise, adequate for the plasma density fluctuations measurements.

  19. New detection system and signal processing for the tokamak ISTTOK heavy ion beam diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Henriques, R. B.; Nedzelskiy, I. S.; Malaquias, A.; Fernandes, H. [Associacao Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, 1049-001 Lisboa (Portugal)

    2012-10-15

    The tokamak ISTTOK heavy ion beam diagnostic (HIBD) operates with a multiple cell array detector (MCAD) that allows for the plasma density and the plasma density fluctuations measurements simultaneously at different sampling volumes across the plasma. To improve the capability of the plasma density fluctuations investigations, a new detection system and new signal conditioning amplifier have been designed and tested. The improvements in MCAD design are presented which allow for nearly complete suppression of the spurious plasma background signal by applying a biasing potential onto special electrodes incorporated into MCAD. The new low cost and small size transimpedance amplifiers are described with the parameters of 400 kHz, 10{sup 7} V/A, 0.4 nA of RMS noise, adequate for the plasma density fluctuations measurements.

  20. Development of a positronium atom diagnostic beam to study transport in tokamaks

    International Nuclear Information System (INIS)

    Surko, C.M.

    1992-01-01

    Anomalous transport is probably the key physics issue in magnetic confinement fusion. It severely limits our ability to create and maintain a dense, hot, fusion plasma. There is also good evidence that there is a fundamental lack of understanding in this area, namely the transport of particles and energy induced by waves and fluctuations in magnetically confined plasmas. The positronium atom beam diagnostic, if successful, can provide a qualitatively new and different way of studying transport in tokamak fusion plasmas. The use of the positron as a thermalized, electron-mass test particle will allow important new tests of current theories of plasma transport. In particular, it could provide unique insights into the potential role of magnetic fluctuations in producing anomalous transport. This is particularly significant in that, at present, there is essentially no in situ probe of these fluctuations or the transport which they produce. Some results of this study are summarized

  1. The circuit of polychromator for Experimental Advanced Superconducting Tokamak edge Thomson scattering diagnostic

    International Nuclear Information System (INIS)

    Zang, Qing; Zhao, Junyu; Chen, Hui; Li, Fengjuan; Hsieh, C. L.

    2013-01-01

    The detector circuit is the core component of filter polychromator which is used for scattering light analysis in Thomson scattering diagnostic, and is responsible for the precision and stability of a system. High signal-to-noise and stability are primary requirements for the diagnostic. Recently, an upgraded detector circuit for weak light detecting in Experimental Advanced Superconducting Tokamak (EAST) edge Thomson scattering system has been designed, which can be used for the measurement of large electron temperature (T e ) gradient and low electron density (n e ). In this new circuit, a thermoelectric-cooled avalanche photodiode with the aid circuit is involved for increasing stability and enhancing signal-to-noise ratio (SNR), especially the circuit will never be influenced by ambient temperature. These features are expected to improve the accuracy of EAST Thomson diagnostic dramatically. Related mechanical construction of the circuit is redesigned as well for heat-sinking and installation. All parameters are optimized, and SNR is dramatically improved. The number of minimum detectable photons is only 10

  2. First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

    Science.gov (United States)

    Bin, W.; Bruschi, A.; D'Arcangelo, O.; Castaldo, C.; De Angeli, M.; Figini, L.; Galperti, C.; Garavaglia, S.; Granucci, G.; Grosso, G.; Korsholm, S. B.; Lontano, M.; Mellera, V.; Minelli, D.; Moro, A.; Nardone, A.; Nielsen, S. K.; Rasmussen, J.; Simonetto, A.; Stejner, M.; Tartari, U.

    2015-10-01

    Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under certain circumstances, the use of the ECRH in fusion devices. An accurate characterization of the conditions for the occurrence of this phenomenon and of its consequences is thus of primary importance. Exploiting the front-steering configuration available with the real-time launcher, the implementation of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system during the very first operations in 2014. The present work has been carried out under an EUROfusion Enabling Research project. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

  3. Recent run-time experience and investigation of impurities in turbines circuit of Helium plant of SST-1

    International Nuclear Information System (INIS)

    Panchal, P.; Panchal, R.; Patel, R.

    2013-01-01

    One of the key sub-systems of Steady State superconducting Tokamak (SST-1) is cryogenic 1.3 kW at 4.5 K Helium refrigerator/liquefier system. The helium plant consists of 3 nos. of screw compressors, oil removal system, purifier and cold-box with 3 turbo expanders (turbines) and helium cold circulator. During the recent SST-1 plasma campaigns, we observed high pressure drop of the order of 3 bar between the wheel outlet of turbine A and the wheel inlet of turbine - B. This was significant higher values of pressures drop across turbines, which reduced the speed of turbine A and B and in turn reduced the overall plant capacity. The helium circuits in the plant have 10-micron filter at the mouth of turbine - B. Initially, major suspects of such high blockage are assumed to be air-impurity, dust particles or collapse of filter. Several breaks in plant operation have been taken to warm up the turbines circuits up to 90 K to remove condensation of air-impurities at filter. Still this exercise did not solve blockage of filter in turbine circuits. A detailed investigation exercise with air/water regeneration and rinsing of cold box as well as purification of helium gas in buffer tanks are carried out to remove air impurities from cold-box. A trial run of cold box was executed in liquefier mode with turbines up to cryogenic temperatures and solved blockage in turbine circuits. The paper describes run-time experience of helium plant with helium impurity in turbine circuits, methods to remove impurity, demonstration of turbine performance and lessons learnt during this operation. (author)

  4. Novel, potent, and radio-iodinatable somatostatin receptor 1 (sst1) selective analogues.

    Science.gov (United States)

    Erchegyi, Judit; Cescato, Renzo; Grace, Christy Rani R; Waser, Beatrice; Piccand, Véronique; Hoyer, Daniel; Riek, Roland; Rivier, Jean E; Reubi, Jean Claude

    2009-05-14

    The proposed sst(1) pharmacophore (J. Med. Chem. 2005, 48, 523-533) derived from the NMR structures of a family of mono- and dicyclic undecamers was used to design octa-, hepta-, and hexamers with high affinity and selectivity for the somatostatin sst(1) receptor. These compounds were tested for their in vitro binding properties to all five somatostatin (SRIF) receptors using receptor autoradiography; those with high SRIF receptor subtype 1 (sst(1)) affinity and selectivity were shown to be agonists when tested functionally in a luciferase reporter gene assay. Des-AA(1,4-6,10,12,13)-[DTyr(2),DAgl(NMe,2naphthoyl)(8),IAmp(9)]-SRIF-Thr-NH(2) (25) was radio-iodinated ((125)I-25) and specifically labeled sst(1)-expressing cells and tissues. 3D NMR structures were calculated for des-AA(1,4-6,10,12,13)-[DPhe(2),DTrp(8),IAmp(9)]-SRIF-Thr-NH(2) (16), des-AA(1,2,4-6,10,12,13)-[DAgl(NMe,2naphthoyl)(8),IAmp(9)]-SRIF-Thr-NH(2) (23), and des-AA(1,2,4-6,10,12,13)-[DAgl(NMe,2naphthoyl)(8),IAmp(9),Tyr(11)]-SRIF-NH(2) (27) in DMSO. Though the analogues have the sst(1) pharmacophore residues at the previously determined distances from each other, the positioning of the aromatic residues in 16, 23, and 27 is different from that described earlier, suggesting an induced fit mechanism for sst(1) binding of these novel, less constrained sst(1)-selective family members.

  5. Identification of Plasma Parameters and Optimization of Magnetic Sensors in the Superconducting Steady-State Tokamak-1 Using Neural Networks

    International Nuclear Information System (INIS)

    Sengupta, A.; Ranjan, P.

    2001-01-01

    In this paper, we examine the possibility of using a multilayered feedforward neural network to extract tokamak plasma parameters from magnetic measurements as an improvement over the traditional methodology of function parametrization. It is also used to optimize the number and locations of the magnetic diagnostics designed for the tokamak. This work has been undertaken with the specific purpose of application of the neural network technique to the newly designed (and currently under fabrication) Superconducting Steady-State Tokamak-1 (SST-1). The magnetic measurements will be utilized to achieve real-time control of plasma shape, position, and some global profiles. A trained neural network is tested, and the results of parameter identification are compared with function parametrization. Both techniques appear well suited for the purpose, but a definite improvement with neural networks is observed. Although simulated measurements are used in this work, confidence regarding the network performance with actual experimental data is ensured by testing the network's noise tolerance with Gaussian noise of up to 10%. Finally, three possible methods of ranking the diagnostics in decreasing order of importance are suggested, and the neural network is used to optimize the number and locations of the magnetic sensors designed for SST-1. The results from the three methods are compared with one another and also with function parametrization. Magnetic probes within the plasma-facing side of the outboard limiter have been ranked high. Function parametrization and one of the neural network methods show a distinct tendency to favor the probes in the remote regions of the vacuum vessel, proving the importance of redundancy. Fault tolerance of the optimized network is tested. The results obtained should, in the long run, help in the decision regarding the final effective set of magnetic diagnostics to be used in SST-1 for reconstruction of the control parameters

  6. High-pressure duo-multichannel soft x-ray spectrometer for tokamak plasma diagnostics

    International Nuclear Information System (INIS)

    Schwob, J.L.; Wouters, A.W.; Suckewer, S.

    1987-03-01

    A high-resolution, time-resolving soft X-ray multichannel spectrometer (SOXMOS) that permits the simultaneous measurement of emission in two different spectral ranges has been developed and tested extensively for tokamak plasma diagnostics. The basic instrument is a high-resolution, interferometrically adjusted, extreme grazing incidence Schwob-Fraenkel duochromator. The instrument is equipped with two multichannel detectors that are adjusted interferometrically and scan along the Rowland circle. Each consists of an MgF 2 coated, funneled microchannel plate, associated with a phosphor screen image intensifier that is coupled to a 1024-element photodiode array by a flexible fibrer optic conduit. The total wavelength coverage of the instrument is 5 to 340 0 A with a measured resolution (FWHM) of about 0.2 A when equipped with a 600 g/mm grating, and 5 to 85 A with a resolution of about 0.06 A using a 2400 g/mm grating. The simultaneous spectral coverage of each detector varies from 15 A at the short wavelength limit to 70 A at the long wavelength limit with the lower dispersion grating. The minimum read-out time for a full spectral portion is 17 ms, but several individual lines can be measured with 1 ms time resolution by selected pixel readout. Higher time resolution can be achieved by replacing one multichannel detector with a single channel electron multiplier detector. Examples of data from the PLT and TFTR tokamaks are presented to illustrate the instrument's versatility, high spectral resolution, and high signal-to-noise ratio even in the 10 A region. 44 refs., 20 figs

  7. Diagnostics upgrade and capability available for physics study on EAST tokamak

    International Nuclear Information System (INIS)

    Hu Liqun

    2013-01-01

    As a consequence of employment of many new techniques and upgrade of EAST superconducting tokamak to enhance divertor plasma performance, significant achievement has been realized in 2012, including 400s long pulse plasma, stationary 35s H-mode and 3.45s H-mode with only ion cyclotron resonant heating (ICRH) etc. To approach steady-state (SS) operation of high-performance plasmas and address key physics on fusion reactor-relevent subjects, recently, capability of the plasma heating and current drive of EAST machine are doubled with total auxiliary injection power over 20 MW, including new methodology of neutral beam injection (NBI) and electron cyclotron resonant heating (ECRH). Most diagnostics have been upgraded to be more compact and integrated due to limited port window and space available, and new advanced neutral-beam related diagnostics have been developed as well, to provide profile of all key parameters for study and understanding critical issues specific to SS high performance plasma. (author)

  8. Measurement of electron density of the plasma in the Tokamak TCABR, through Thomson scattering diagnostic

    International Nuclear Information System (INIS)

    Jeronimo, Leonardo Cunha

    2013-01-01

    Over the last few years is remarkable, so increasingly evident the need for a new source of energy for mankind. One promising option is through nuclear fusion, where the plasma produced in the reactor can be converted into electrical energy. Therefore, knowing the characteristics of this plasma is very important to control it and understand it so desirable. One of the diagnostic options is called Thomson scattering . This is considered the most reliable method for the determination of important plasma parameters such as temperature and electron density, and may also help in the study and explanation of various internal mechanisms. The great advantage lies in the tact that they consist of a direct measurement and nonperturbative. But it is a diagnosis whose installation and execution is admittedly complex, limiting it only a few laboratories in the fíeld of fusion for the world. Among the main difficulties, wc can highlight the fact that the scattered signal is very small, thus requiring a large increase of the incident power. Moreover, the external physical conditions can cause mechanical vibrations that eliminate or minimize them as much as possible, is a great challenge, considering the optical micrometrically very sensitive and needs involved in the system. This work describes the entire process of installation and operation of Thomson scattering diagnostic in tokamak TCABR and through this diagnosis, we work on results of electron temperature, to finally be able to calculate the electron density of the plasma. (author)

  9. Commissioning and Operational Experience with 1 kW Class Helium Refrigerator/Liquefier for SST-1

    Science.gov (United States)

    Dhard, C. P.; Sarkar, B.; Misra, Ruchi; Sahu, A. K.; Tanna, V. L.; Tank, J.; Panchal, P.; Patel, J. C.; Phadke, G. D.; Saxena, Y. C.

    2004-06-01

    The helium refrigerator/liquefier (R/L) for the Steady State Super conducting Tokamak (SST-1) has been developed with very stringent specifications for the different operational modes. The total refrigeration capacity is 650 W at 4.5 K and liquefaction capacity of 200 l/h. A cold circulation pump is used for the forced flow cooling of 300 g/s supercritical helium (SHe) for the magnet system (SCMS). The R/L has been designed also to absorb a 200 W transient heat load of the SCMS. The plant consists of a compressor station, oil removal system, on-line purifier, Main Control Dewar (MCD) with associated heat exchangers, cold circulation pump and warm gas management system. An Integrated Flow Control and Distribution System (IFDCS) has been designed, fabricated and installed for distribution of SHe in the toroidal and poloidal field coils as well as liquid helium for cooling of 10 pairs of current leads. A SCADA based control system has been designed using PLC for R/L as well as IFDCS. The R/L has been commissioned and required parameters were achieved confirming to the process. All the test results and commissioning experiences are discussed in this paper.

  10. Magnetic shielding for FEL microwave electric field diagnostic in MTX tokamak

    International Nuclear Information System (INIS)

    Yamada, Shinichi; Odajima, Kazuo; Ishida, Hiroyasu

    1991-07-01

    A diagnostic system for measurement of microwave electric field from free electron laser (FEL) is in preparation at JAERI under JAERI-DOE collaborative program in the Microwave Tokamak Experiment (MTX) being held at Lawrence Livermore National Laboratory in U.S.A.. That is called LAPPS (Laser Aided Particle Probe Spectroscopy). This is consist of helium neutral beam source, a dye laser and viewing optics. It is required that 1000 gauss of the magnetic field must be shielded to less than 1 gauss in order to operate these LAPPS components. New high performance soft ferrous magnetic material 'FERROPERM' and PERMALLOY are used on this purpose. This paper proposes a new method to estimate a required thickness of the magnetic shielding in a saturated region of B-H curve, that is, 'magnetic shielding calculation by Virtual Divided Layers Method (VDLM)', where the shielding layer is virtually divided in many layers in the calculation. The results are compared with a computer simulation using 'three dimensional static magnetic field code' and with experimental results in a uniform static field. (author)

  11. Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y. L.; Xu, G. S.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Tritz, K. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, Maryland 21218 (United States); Zhu, Y. B. [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States)

    2015-12-15

    A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.

  12. Low-noise cable for diagnostics, control and instrumentation of the ASDEX tokamak fusion experiment

    International Nuclear Information System (INIS)

    Gernhardt, J.

    1988-11-01

    ASDEX (Axially Symmetric Divertor EXperiment) is a large tokamak (R=1.65 m; a=0.4 m) with an air transformer. The relatively large stray field, Bz=10 mT=(100 G); for ρ=5 m, Bz=40 mT=(400 G); for ρ=3 m, BΦ=0.3 T=(3 kG); for ρ=3m, compared with that of an iron transformer, and the cable length 1≤30 m from the experiment to the control room, make mainly the magnetically induced and capacitively coupled noise signal in the cable relatively high. As a result of neutral injection (> 4 MW; 40 kV) and lower hybrid ion cyclotron and Alfven wave heating strong E-fields are produced and noise is coupled into the cables. These magnetic and electric field gradients during the plasma shot vary with time and location. This report tries to show how these noise signals can be reduced without reducing the broadcast frequency of the signal. The Electro Magnetic Compatibility and Interference (EMC, EMI) are discussed. The cost of diagnostic cable, connectors and cable ducts without mounting is approximately DM 700,000.--. (orig.)

  13. The charge exchange recombination diagnostic system on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Gohil, P.; Burrell, K.H.; Groebner, R.J.; Kim, J.; Martin, W.C.; McKee, E.L.; Seraydarian, R.P.

    1991-11-01

    The charge exchange recombination (CER) diagnostic system on the DIII-D tokamak is used to make spatially and temporally resolved measurements of the ion temperature and toroidal and poloidal rotation velocities. This is performed through visible spectroscopic measurements of the Doppler broadened and Doppler shifted HE II 468.6 nm, the CVI 529.1 nm, and the BV 494.5 nm spectral lines which have been excited by charge exchange recombination interactions between the fully stripped ions and the neutral atoms from the heating beams. The plasma viewing optics comprises 32 viewing chords spanning a typical plasma minor radius of 63 cm across the midplane, of which 15 spatial chords span 4.2 cm at the plasma edge just within the separatrix and provide a chord-to-chord spatial resolution of 0.3 cm. Fast camera readout electronics can provide a temporal resolution of 260 μs per time slice, but the effective minimum integration time, at present, is 1 ms which is limited by the detected photon flux from the plasma and the decay times of the phosphors used on the multichannel plate image intensifiers. Significant changes in the edge plasma radial electric field at the L-H transition have been observed, as determined from the CER measurements, and these results are being extensively compared to theories which consider the effects of sheared electric fields on plasma turbulence. 13 refs., 10 figs

  14. Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Li, Y. L.; Xu, G. S.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L.; Tritz, K.; Zhu, Y. B.

    2015-01-01

    A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks

  15. Synchrotron emission diagnostic of full-orbit kinetic simulations of runaway electrons in tokamaks plasmas

    Science.gov (United States)

    Carbajal Gomez, Leopoldo; Del-Castillo-Negrete, Diego

    2017-10-01

    Developing avoidance or mitigation strategies of runaway electrons (RE) for the safe operation of ITER is imperative. Synchrotron radiation (SR) of RE is routinely used in current tokamak experiments to diagnose RE. We present the results of a newly developed camera diagnostic of SR for full-orbit kinetic simulations of RE in DIII-D-like plasmas that simultaneously includes: full-orbit effects, information of the spectral and angular distribution of SR of each electron, and basic geometric optics of a camera. We observe a strong dependence of the SR measured by the camera on the pitch angle distribution of RE, namely we find that crescent shapes of the SR on the camera pictures relate to RE distributions with small pitch angles, while ellipse shapes relate to distributions of RE with larger pitch angles. A weak dependence of the SR measured by the camera with the RE energy, value of the q-profile at the edge, and the chosen range of wavelengths is found. Furthermore, we observe that oversimplifying the angular distribution of the SR changes the synchrotron spectra and overestimates its amplitude. Research sponsored by the LDRD Program of ORNL, managed by UT-Battelle, LLC, for the U. S. DoE.

  16. Plasma Fluctuation Studies in the TCV Tokamak: Modeling of Shaping Effects and Advanced Diagnostic Development

    International Nuclear Information System (INIS)

    Marinoni, A.

    2009-10-01

    One of the most important issues for magnetic-confinement fusion research is the so-called anomalous transport across magnetic field lines, i.e. transport that is in excess of that caused by collisional processes. The need to reduce anomalous transport in order to increase the efficiency of a prospective fusion reactor must be addressed through an investigation of its fundamental underlying causes. This thesis is divided into two distinct components: one experimental and instrumental, and the other theoretical and based on numerical modeling. The experimental part consists of the design and installation of a new diagnostic for core turbulence fluctuations in the TCV tokamak. An extensive conceptual investigation of a number of possible solutions, including Beam Emission Spectroscopy, Reflectometry, Cross Polarization, Collective Scattering and different Imaging techniques, was carried out at first. A number of criteria, such as difficulties in data interpretation, costs, variety of physics issues that could be addressed and expected performance, were used to compare the different techniques for specific application to the TCV tokamak. The expected signal to noise ratio and the required sampling frequency for TCV were estimated on the basis of a large number of linear, local gyrokinetic simulations of plasma fluctuations. This work led to the choice of a Zernike phase contrast imaging system in a tangential launching configuration. The diagnostic was specifically designed to provide information on turbulence features up to now unknown. In particular, it is characterized by an outstanding spatial resolution and by the capability to measure a very broad range of fluctuations, from ion to electron Larmor radius scales, thus covering the major part of the instabilities expected to be at play in TCV. The spectrum accessible covers the wavenumber region from 0.9 cm -1 to 60 cm -1 at 24 radial positions with 3 MHz bandwidth. The diagnostic is an imaging technique and is

  17. Characteristics of solid-target charge-exchange analyzers for energetic ion diagnostics on tokamaks

    International Nuclear Information System (INIS)

    Beiersdorfer, P.; Roquemore, A.L.; Kaita, R.

    1987-05-01

    Compact electrostatic charge-exchange analyzers have been constructed for installation in areas of high magnetic fields and restricted access near tokamak fusion devices. The analyzers employed carbon stripping foils, and have been calibrated for proton energies between 1 and 70 keV. They have been successfully used to study charge-exchange losses in auxiliary-heated tokamak plasmas

  18. GHRSST Level 4 Australian Bureau of Meteorology GAMSSA_28km Global SST:1

    Data.gov (United States)

    National Aeronautics and Space Administration — The GAMSSA v1.0 system blends NAVOCEANO's GAC 9.9 km x 4.4 km resolution AVHRR L2P SST1m data (NOAA-17, NOAA-18 and METOP-A), European Space Agency's 0.17 AATSR skin...

  19. Characterization and commissioning of the SST-1M camera for the Cherenkov Telescope Array

    Czech Academy of Sciences Publication Activity Database

    Aguilar, J.A.; Bilnik, W.; Blocki, J.; Bogacz, L.; Borkowski, J.; Bulik, T.; Cadoux, F.; Christov, A.; Curylo, M.; della Volpe, D.; Dyrda, M.; Favre, Y.; Frankowski, A.; Grudnik, L.; Grudzinska, M.; Heller, M.; Idzkowski, B.; Jamrozy, M.; Janiak, M.; Kasperek, J.; Lalik, K.; Lyard, E.; Mach, E.; Mandát, Dušan; Marszalek, A.; Medina Miranda, L. D.; Michałowski, J.; Moderski, R.; Montaruli, T.; Neronov, A.; Niemiec, J.; Ostrowski, M.; Pasko, P.; Pech, Miroslav; Porcelli, A.; Prandini, E.; Rajda, P.; Rameez, M.; Schioppa, E.; Schovánek, Petr; Seweryn, K.; Skowron, K.; Sliusar, V.; Sowinski, M.; Stawarz, L.; Stodulska, M.; Stodulski, M.; Toscano, S.; Troyano Pujadas, I.; Walter, R.; Wiecek, M.; Zagdanski, A.; Zietara, K.; Zychowski, P.

    2017-01-01

    Roč. 845, Feb (2017), s. 350-354 ISSN 0168-9002 Institutional support: RVO:68378271 Keywords : SiPM * G-APD * CTA * SST-1 M * Gamma-ray Astronomy * FPGA * PhotoDetection * high-speed electronics * ADC Subject RIV: BF - Elementary Particles and High Energy Physics OBOR OECD: Particles and field physics Impact factor: 1.362, year: 2016

  20. Thermo hydraulic and quench propagation characteristics of SST-1 TF coil

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, A.N., E-mail: ansharma@ipr.res.in [Institute for Plasma Research, Gandhinagar (India); Pradhan, S. [Institute for Plasma Research, Gandhinagar (India); Duchateau, J.L. [CEA Cadarache, 13108 St Paul lez Durance Cedex (France); Khristi, Y.; Prasad, U.; Doshi, K.; Varmora, P.; Patel, D.; Tanna, V.L. [Institute for Plasma Research, Gandhinagar (India)

    2014-02-15

    Highlights: • Details of SST-1 TF coils, CICC. • Details of SST-1 TF coil cold test. • Quench analysis of TF magnet. • Flow changes following quench. • Predictive analysis of assembled magnet system. - Abstract: SST-1 toroidal field (TF) magnet system is comprising of sixteen superconducting modified ‘D’ shaped TF coils. During single coil test campaigns spanning from June 10, 2010 till January 24, 2011; the electromagnetic, thermal hydraulic and mechanical performances of each TF magnet have been qualified at its respective nominal operating current of 10,000 A in either two-phase or supercritical helium cooling conditions. During the current charging experiments, few quenches have initiated either as a consequence of irrecoverable normal zones or being induced in some of the TF magnets. Quench evolution in the TF coils have been analyzed in detail in order to understand the thermal hydraulic and quench propagation characteristics of the SST-1 TF magnets. The same were also simulated using 1D code Gandalf. This paper elaborates the details of the analyses and the quench simulation results. A predictive quench propagation analysis of 16 assembled TF magnets system has also been reported in this paper.

  1. Suprathermal electron studies in Tokamak plasmas by means of diagnostic measurements and modeling

    International Nuclear Information System (INIS)

    Kamleitner, J.

    2015-01-01

    To achieve reactor-relevant conditions in a tokamak plasma, auxiliary heating systems are required and can be realized by waves injected in the plasma that heat ions or electrons. Electron cyclotron resonant heating (ECRH) is a very flexible and robust technique featuring localized power deposition and current drive (CD) capabilities. Its fundamental principles are well understood and the application of ECRH is a proven and established tool; electron cyclotron current drive (ECCD) is regularly used to develop advanced scenarios and control magneto-hydrodynamics (MHD) instabilities in the plasma by tailoring the current profile. There remain important open questions, such as the phase space dynamics, the observed radial broadening of the supra-thermal electron distribution function and discrepancies in predicted and experimental CD efficiency. A main goal is to improve the understanding of wave-particle interaction in plasmas and current drive mechanisms. This was accomplished by combined experimental and numerical studies, strongly based on the conjunction of hard X-ray (HXR) Bremsstrahlung measurements and Fokker-Planck modelling, characterizing the supra-thermal electron population. The hard X-ray tomographic spectrometer (HXRS) diagnostic was developed to perform these studies by investigating spatial HXR emission asymmetries in the co- and counter-current directions and within the poloidal plane. The system uses cadmium-telluride detectors and digital acquisition to store the complete time history of incoming photon pulses. An extensive study of digital pulse processing algorithms was performed and its application allows the HXRS to handle high count rates in a noisy tokamak environment. Numerical tools were developed to improve the time resolution by conditional averaging and to obtain local information with the general tomographic inversion package. The interfaces of the LUKE code and the well-established CQL3D Fokker-Planck code to the Tokamak a

  2. New receiving line for the remote-steering antenna of the 140 GHz CTS diagnostics in the FTU Tokamak

    Science.gov (United States)

    D'Arcangelo, O.; Bin, W.; Bruschi, A.; Cappelli, M.; Fanale, F.; Gittini, G.; Pallotta, F.; Rocchi, G.; Tudisco, O.; Garavaglia, S.; Granucci, G.; Moro, A.; Tuccillo, A. A.

    2018-01-01

    A new receiving antenna for collecting signals of the Collective Thomson Scattering (CTS) diagnostics in FTU Tokamak has been recently installed. The squared corrugated section and the precisely defined length make it possible to receive from different directions by remotely steering the receiving mirrors. This type of Remote-Steering (RS) antennas, being studied on FTU for the DEMO Electron Cyclotron Heating (ECH) system launch, is already installed on the W7- X stellarator and will be tested in the next campaign. The transmission of the signal from the antenna in the tokamak hall to the CTS diagnostics hall will be mainly realized by means of oversized circular corrugated waveguides carrying the hybrid HE11 (quasi-gaussian) waveguide mode, with inclusion of a special smooth-waveguide section and a short run of reduced-size square-corrugated waveguide through the tokamak bio-shield. The coupling between different waveguide types is made with ellipsoidal focusing mirrors, using quasi-optical matching formulas between the gaussian-shaped beams in input and output to the waveguides. In this work, after a complete study of feasibility of the overall line, a design for the receiving line will be proposed, in order to realize an executive layout to be used as a guideline for the commissioning phase.

  3. Calculations in the tokamak vacuum region: δWυ and diagnostic capabilities

    International Nuclear Information System (INIS)

    Chance, M.S.

    1993-01-01

    The calculational methods used for the magnetic energy, δW v , in the vacuum region surrounding tokamak plasmas permit a natural extension to more general toroidal configurations, and also for calculating useful ancillary quantities like the simulation of the Mirnov-loop measurements, and the eddy currents generated in the conducting materials outside the plasma discharge. The generalizations include the capabilities for topologically spherical shells, and toroidally symmetric unlinked (with respect to the plasma discharge) solid conductors. The former was motivated by spheromak-like configurations, and the latter for simulating the segmented shell of PBX-M. The collocation method and Green's second identity is used for solving ∇ 2 χ = 0 for the magnetic scalar potential. This lends itself readily to incorporating the necessary boundary conditions, and once χ is known on the surfaces, a second application of Green's identity enables the simulation of the Mirnov-loop measurements. The eddy currents on the surface of the conductors, K c , is given by n x ∇χ. This is useful for visualizing the proposed coil currents needed for stabilizing the external kink modes by feedback mechanisms, and also for estimating the stresses in the shell due to disruptions, etc. The code is modular and can be easily interfaced to various stability codes like the PEST, NOVA, GATO, etc. system of codes, or can be used in a stand-alone mode for diagnostic purposes. Details of the numerical method and results are presented. These include the efficient treatment of the singularities inherently present in the Green's function method

  4. Application of diamond window for infrared laser diagnostics in a tokamak device

    International Nuclear Information System (INIS)

    Kawano, Yasunori; Chiba, Shinichi; Inoue, Akira

    2004-01-01

    Chemical vapor deposited diamond disks have been successfully applied as the vacuum windows for infrared CO 2 laser interferometry and polarimetry used in electron density measurement in the JT-60U tokamak. In comparison with the conventional zinc-selenide windows, the Faraday rotation component of diamond windows was negligible. This results in an improvement of the Faraday rotation measurement of tokamak plasma by polarimetry

  5. Development of high-speed and wide-angle visible observation diagnostics on Experimental Advanced Superconducting Tokamak using catadioptric optics

    International Nuclear Information System (INIS)

    Yang, J. H.; Hu, L. Q.; Zang, Q.; Han, X. F.; Shao, C. Q.; Sun, T. F.; Chen, H.; Wang, T. F.; Li, F. J.; Hu, A. L.; Yang, X. F.

    2013-01-01

    A new wide-angle endoscope for visible light observation on the Experimental Advanced Superconducting Tokamak (EAST) has been recently developed. The head section of the optical system is based on a mirror reflection design that is similar to the International Thermonuclear Experimental Reactor-like wide-angle observation diagnostic on the Joint European Torus. However, the optical system design has been simplified and improved. As a result, the global transmittance of the system is as high as 79.6% in the wavelength range from 380 to 780 nm, and the spatial resolution is <5 mm for the full depth of field (4000 mm). The optical system also has a large relative aperture (1:2.4) and can be applied in high-speed camera diagnostics. As an important diagnostic tool, the optical system has been installed on the HT-7 (Hefei Tokamak-7) for its final experimental campaign, and the experiments confirmed that it can be applied to the investigation of transient processes in plasma, such as ELMy eruptions in H-mode, on EAST

  6. Study on the characters of high voltage charging power supply system for diagnostics neutral beam on HT-7 Tokamak

    International Nuclear Information System (INIS)

    Zhang Jian; Huang Yiyun; Liu Baohua; Guo Wenjun; Shen Xiaoling; Wei Wei

    2011-01-01

    A high voltage power supply system has been developed for the diagnostic neutral beam on the HT-7 experimental Tokamak, and the over-voltage phenomenon of storage capacitor was founded in the experiment. In order to analyse and resolve this problem, the structure and principle of high voltage power supply is described and the primary high voltage charging power supply system is introduced in detail. The phenomenon of over-voltage on the capacitors is also studied with circuit model, and the conclusion is obtained that the leakage inductance is the mA in reason which causes the over-voltage on the capacitors. (authors)

  7. Vacuum system design and tritium inventory for the charge exchange diagnostic on the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Medley, S.S.

    1986-01-01

    The application of charge exchange analyzers for the measurement of ion temperature in fusion plasma experiments requires a direct connection between the diagnostic and plasma-discharge vacuum chambers. Differential pumping of the gas load from the diagnostic stripping cell operated at > or approx. = 10 -3 Torr is required to maintain the analyzer chamber at a pressure of -6 Torr. The migration of gases between the diagnostic and plasma vacuum chambers must be minimized. In particular, introduction of the analyzer stripping cell gas into the plasma chamber having a base pressure of -8 Torr must be suppressed. The charge exchange diagnostic for the Tokamak Fusion Test Reactor (TFTR) is comprised of two analyzer systems designed to contain a total of 18 independent mass/energy analyzers and one diagnostic neutral beam rated at 80 keV, 15 A. The associated arrays of multiple, interconnected vacuum systems were analyzed using the Vacuum System Transient Simulator (Vsts) computer program which models the transient transport of multigas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced costs, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus

  8. Understanding of impurity behavior in SST-1 plasmas using visible spectroscopy

    International Nuclear Information System (INIS)

    Manchanda, Ranjana; Ramaiya, Nilam; Chowdhuri, Malay Bikas; Banerjee, Santanu; Ghosh, Joydeep

    2015-01-01

    Studies of impurity behavior in SST-1 plasma have been carried out using visible spectroscopic systems installed on the tokomak. This has been carried out using a low resolution and broadband survey spectrometer covering a 350-900 nm wavelength range, 0.5 m visible spectrometer having 600 and 1200 grooves/mm grating coupled with CCD camera and interference filter and photomultiplier (PMT) tube based systems. Temporal evolution of the hydrogen (H α , H β ) and impurities emissions like, C II, C III, O I, O II, O III, O V and a visible Continuum at 536.0 nm have been monitored using the PMT based system to understand impurity charge state evolution during plasma discharges. All systems are absolutely calibrated for impurity influx and plasma parameter estimations. Observed spectral lines in the visible range have been identified to recognize the presence of various impurities in the SST-1 plasmas. Comparison of impurities emission has been made for different plasma currents and toroidal magnetic fields. An analysis has been carried out to understand the impurities activities in plasmas of SST-1 tokomak in presence and absence of installed plasma facing components (PFC). Significantly higher carbon emissions have been observed indicating higher carbon content in the plasma with graphite PFCs installed. (author)

  9. Supervisory control and data acquisition system development for superconducting current feeder system of SST-1

    International Nuclear Information System (INIS)

    Patel, R.; Mahesuria, G.; Gupta, N.C.; Sonara, D.; Panchal, R.; Panchal, P.; Tanna, V.L.; Pradhan, S.

    2014-01-01

    The Current Feeders System (CFS) is essentially an optimized bridge between the power supply at room temperature and Super Conducting Magnet System (SCMS) of the SST-1 machine at 4.5 K.CFS is a complex electrical and cryogenic network which consists of ten pairs of 10 KA rating helium Vapor cooled Conventional Current Leads (VCCLs), superconducting (SC) current feeder and associated components. For the safe and reliable operation of CFS, it is equipped with different physical process parameters measuring instruments like flow, pressure, temperature, level, vacuum, voltage taps and final control element like control valves, heaters, vacuum pumps etc. PLC program is developed in ladder language for acquiring and controlling the process parameters. Independent SCADA applications developed in WonderwareIntouch software for data communication from PLC, front-end Graphical User Interface (GUI), auto-manual interface, real time trends, history trends, events and alarm pages. Time synchronized communication established between CFS control system and Industrial SQL server (InSQL) Historian for centralized storage of CFS process parameters which intern provides the CFS process data to SST-1 central control room. SCADA based data acquisition and data retrieval system is found to be satisfactory during the recent SST-1 cool down experiment. This paper describes the SCADA and PLC application development and their communication to InSQL server. (author)

  10. IR and FIR laser polarimetry as a diagnostic tool in high-. beta. and Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, D; Machida, M; Scalabrin, A

    1986-03-01

    The change of the polarization state of an electromagnetic wave (EMW) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the EMW allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-pinch plasmas (lambda approx. 10..mu..m). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-pinch/UNICAMP, using lasers developed at UNICAMP.

  11. Use of soft x-ray diagnostic on the COMPASS tokamak for investigations of sawteeth crash neighborhood and of plasma position using fast inversion methods

    Energy Technology Data Exchange (ETDEWEB)

    Imrisek, M. [Institute of Plasma Physics ASCR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Weinzettl, V.; Mlynar, J.; Panek, R.; Hron, M. [Institute of Plasma Physics ASCR, Prague (Czech Republic); Odstrcil, T. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Odstrcil, M. [Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Optical Research Center, University of Southampton, Southampton (United Kingdom); Ficker, O. [Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Pinzon, J. R. [Institue Jean Lamour, Université de Lorraine, Nancy (France); Ehrlacher, C. [ENS Cachan, Paris (France)

    2014-11-15

    The soft x-ray diagnostic is suitable for monitoring plasma activity in the tokamak core, e.g., sawtooth instability. Moreover, spatially resolved measurements can provide information about plasma position and shape, which can supplement magnetic measurements. In this contribution, fast algorithms with the potential for a real-time use are tested on the data from the COMPASS tokamak. In addition, the soft x-ray data are compared with data from other diagnostics in order to discuss possible connection between sawtooth instability on one side and the transition to higher confinement mode, edge localized modes and productions of runaway electrons on the other side.

  12. Introducing minimum Fisher regularisation tomography to AXUV and soft x-ray diagnostic systems of the COMPASS tokamak

    International Nuclear Information System (INIS)

    Mlynar, J.; Weinzettl, V.; Imrisek, M.; Odstrcil, M.; Havlicek, J.; Janky, F.; Alper, B.; Murari, A.

    2012-01-01

    The contribution focuses on plasma tomography via the minimum Fisher regularisation (MFR) algorithm applied on data from the recently commissioned tomographic diagnostics on the COMPASS tokamak. The MFR expertise is based on previous applications at Joint European Torus (JET), as exemplified in a new case study of the plasma position analyses based on JET soft x-ray (SXR) tomographic reconstruction. Subsequent application of the MFR algorithm on COMPASS data from cameras with absolute extreme ultraviolet (AXUV) photodiodes disclosed a peaked radiating region near the limiter. Moreover, its time evolution indicates transient plasma edge cooling following a radial plasma shift. In the SXR data, MFR demonstrated that a high resolution plasma positioning independent of the magnetic diagnostics would be possible provided that a proper calibration of the cameras on an x-ray source is undertaken.

  13. Development of ion diagnostic system based on electrostatic probe in the boundary plasma of the JFT-2M tokamak

    International Nuclear Information System (INIS)

    Uehara, Kazuya; Kawakami, Tomohide; Amemiya, Hiroshi; Hoethker, K.; Cosler, A.; Bieger, W.

    1995-06-01

    An ion diagnostic system using electrostatic probes for measurements in the JFT-2M tokamak boundary plasma has been developed under the collaboration program between KFA and JAERI. The rotating double probe system, on which the Hoethker double probe and Amemiya asymmetric probe can mounted, are manufactured at KFA workshop while the linear driver to support the rotating double probe, the ion toothbrush probe, the Katsumata probe and the cubic Mach probe are developed at JAERI. This report describes the hardware of this probe system for ion diagnostics in the boundary plasma and preliminary data obtained by means of this system. Furthermore, results on the transport are estimated on the basis of these probe data. (author)

  14. First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

    International Nuclear Information System (INIS)

    Litnovsky, A.; Rudakov, D.L.; De Temmerman, G.; Wienhold, P.; Philipps, V.; Samm, U.; McLean, A.G.; West, W.P.; Wong, C.P.C.; Brooks, N.H.; Watkins, J.G.; Wampler, W.R.; Stangeby, P.C.; Boedo, J.A.; Moyer, R.A.; Allen, S.L.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J.; Boivin, R.L.

    2008-01-01

    Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 o C neither carbon deposition nor degradation of optical properties was detected

  15. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

    Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  16. Implementation of time synchronized cryogenics control system network architecture for SST-1

    Energy Technology Data Exchange (ETDEWEB)

    Patel, Rakesh J., E-mail: rpatel@ipr.res.in; Mahesuria, Gaurang; Panchal, Pradip; Panchal, Rohit; Sonara, Dasarath; Tanna, Vipul; Pradhan, Subrata

    2016-11-15

    Highlights: • SST-1 cryogenics sub-systems are 1.3 kW HRL, LN2 distribution system, current feeders system and 80 K booster system. • GUI developed in SCADA and control program developed in PLC for automation of the above sub-systems. • Implemented the cryogenics control system network to communicate all systems to InSQL server. • InSQL server configured for real time centralized process data acquisition from all connected sub-systems control nodes. • Acquired the process parameters coming from different systems at same time stamp. - Abstract: Under the SST-1 mission mandate, the several cryogenic sub-systems have been developed, upgraded and procured in prior to the SST-1 operation. New developments include 80 K Bubble type thermal shields, LN2 distribution system, LN2 booster system and current feeders system (CFS).Graphical User Interface (GUI) program developed in Wonderware SCADA and control logic program developed in Schneider make PLC for the above sub-systems. Industrial SQL server (InSQL) configured for centralized storage of real time process data coming from various control nodes of cryogenics sub-systems. The cryogenics control system network for communicating all cryogenics sub-system control nodes to InSQL server for centralized data storage and time synchronization among cryogenic sub-systems with centralized InSQL server is successfully implemented. Due to implemented time synchronization among sub-systems control nodes, it is possible to analyze the process parameters coming from different sub-systems at same time stamp. This paper describes the overview of implemented cryogenics control system network architecture for real time cryogenic process data monitor, storage and retrieval.

  17. Implementation of time synchronized cryogenics control system network architecture for SST-1

    International Nuclear Information System (INIS)

    Patel, Rakesh J.; Mahesuria, Gaurang; Panchal, Pradip; Panchal, Rohit; Sonara, Dasarath; Tanna, Vipul; Pradhan, Subrata

    2016-01-01

    Highlights: • SST-1 cryogenics sub-systems are 1.3 kW HRL, LN2 distribution system, current feeders system and 80 K booster system. • GUI developed in SCADA and control program developed in PLC for automation of the above sub-systems. • Implemented the cryogenics control system network to communicate all systems to InSQL server. • InSQL server configured for real time centralized process data acquisition from all connected sub-systems control nodes. • Acquired the process parameters coming from different systems at same time stamp. - Abstract: Under the SST-1 mission mandate, the several cryogenic sub-systems have been developed, upgraded and procured in prior to the SST-1 operation. New developments include 80 K Bubble type thermal shields, LN2 distribution system, LN2 booster system and current feeders system (CFS).Graphical User Interface (GUI) program developed in Wonderware SCADA and control logic program developed in Schneider make PLC for the above sub-systems. Industrial SQL server (InSQL) configured for centralized storage of real time process data coming from various control nodes of cryogenics sub-systems. The cryogenics control system network for communicating all cryogenics sub-system control nodes to InSQL server for centralized data storage and time synchronization among cryogenic sub-systems with centralized InSQL server is successfully implemented. Due to implemented time synchronization among sub-systems control nodes, it is possible to analyze the process parameters coming from different sub-systems at same time stamp. This paper describes the overview of implemented cryogenics control system network architecture for real time cryogenic process data monitor, storage and retrieval.

  18. Simulation of scenarios of LHCD antenna for pre-ionization in SST1 machine

    International Nuclear Information System (INIS)

    Sharma, P.K.; Ambulkar, K.K.; Dalakoti, S.; Virani, C.G.; Parmar, P.R.; Thakur, A.L.

    2013-01-01

    SST1 machine has a continuous vacuum vessel, which inhibits the penetration of Ohmic electric field in to the vessel thereby reducing the peak loop voltage in the machine required for Ohmic breakdown. Alternatively, electron cyclotron resonance (ECR) preionization technique is used for preionization, to assist plasma start-up with lower available loop voltages. In early eighties, lower hybrid current drive (LHCD) system, was also used in PLT machine, for preionization and start-up purpose. The PLT LHCD system was based on 800MHz source and could have provided electric field across large distances because of longer wavelength, thereby assisting gas breakdown. In SST1 machine, the LHCD system is based on 3.7 GHz klystron sources and may not produce favourable conditions for gas breakdown owing to its shorter wavelength. In this paper, we have proposed a novel way to excite LHCD antenna so that electric field variation is created over large spatial distances, conducive for gas breakdown studies. In this scenario, all the elements of the grill antenna are not energized. Out of 32 elements of the grill antenna, only 16 elements are energized. In this special configuration, a periodic arrangement of four adjacent active elements is realized, leaving another set of four elements, adjacent to it, without any power. The CST microwave studio, commercially available software, is used to simulate the above scenario to study the behaviour of electric field produced in this configuration. In this paper we present the modelling aspect of the antenna and the results obtained from the simulation analysis is discussed in details for proposing and planning of preionization experiments on SST1 machine. (author)

  19. Investigation of small-scale tokamak plasma turbulence by correlative UHR backscattering diagnostics

    International Nuclear Information System (INIS)

    Gusakov, E Z; Gurchenko, A D; Altukhov, A B; Bulanin, V V; Esipov, L A; Kantor, M Yu; Kouprienko, D V; Lashkul, S I; Petrov, A V; Stepanov, A Yu

    2006-01-01

    Fine scale turbulence is considered nowadays as a possible candidate for the explanation of anomalous ion and electron energy transport in magnetized fusion plasmas. The unique correlative upper hybrid resonance backscattering (UHR BS) technique is applied at the FT-2 tokamak for investigation of density fluctuations excited in this turbulence. The measurements are carried out in Ohmic discharge at several values of plasma current and density and during current ramp up experiment. The moveable focusing antennas set have been used in experiments allowing probing out of equatorial plane. The radial wave number spectra of the small-scale component of tokamak turbulence are determined from the correlation data with high spatial resolution. Two small-scale modes possessing substantially different phase velocities are observed in plasma under conditions when the threshold for the electron temperature gradient mode excitation is overcome. The possibility of plasma poloidal velocity profile determination using the UHR BS signal is demonstrated

  20. Long pulse characteristics of 5 MW ion source for SST-1 neutral beam injector

    Energy Technology Data Exchange (ETDEWEB)

    Jana, M.R. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)], E-mail: mukti@ipr.res.in; Mattoo, S.K.; Chakraborty, A.K.; Baruah, U.K.; Patel, G.B.; Jayakumar, P.K. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2008-10-15

    We present characteristics of a 5 MW ion source for SST-1 neutral beam injector. Before the source could be tested for its performance, it was conditioned by 480 arc discharges of 1 s and beam extraction of hydrogen species at various beam voltages ranging between 19 kV and 56 kV. Breakdown free beam extraction could be secured only after about 3000 beam second extraction. The ion source is capable of delivering 1.7 MW of neutral beam power at 55 kV with horizontal and vertical focal length of 5.4 m and 7 m respectively. Beam divergence is {approx}0.97 deg. Steady-state beam energy of 31 MJ at 41 kV was achieved during 14 s long beam extraction. We have not noticed any deterioration of beam parameters, including beam divergence during long pulse operation. These results indicate that 0.5 MW of neutral beam power at 30 kV required for heating of plasma in SST-1 can be delivered.

  1. Long pulse characteristics of 5 MW ion source for SST-1 neutral beam injector

    International Nuclear Information System (INIS)

    Jana, M.R.; Mattoo, S.K.; Chakraborty, A.K.; Baruah, U.K.; Patel, G.B.; Jayakumar, P.K.

    2008-01-01

    We present characteristics of a 5 MW ion source for SST-1 neutral beam injector. Before the source could be tested for its performance, it was conditioned by 480 arc discharges of 1 s and beam extraction of hydrogen species at various beam voltages ranging between 19 kV and 56 kV. Breakdown free beam extraction could be secured only after about 3000 beam second extraction. The ion source is capable of delivering 1.7 MW of neutral beam power at 55 kV with horizontal and vertical focal length of 5.4 m and 7 m respectively. Beam divergence is ∼0.97 deg. Steady-state beam energy of 31 MJ at 41 kV was achieved during 14 s long beam extraction. We have not noticed any deterioration of beam parameters, including beam divergence during long pulse operation. These results indicate that 0.5 MW of neutral beam power at 30 kV required for heating of plasma in SST-1 can be delivered.

  2. Camera calibration strategy of the SST-1M prototype of the Cherenokov Telescope Array

    CERN Document Server

    Prandini, E; Lyard, E.; Schioppa, E. jr.; Neronov, A.; Bilnik, W.; Błocki, J.; Bogacz, L.; Bulik, T.; Cadoux, F.; Christov, A.; Curyło, M.; della Volpe, D.; Dyrda, M.; Favre, Y.; Frankowski, A.; Grudnik, Ł.; Grudzińska, M.; Idźkowski, B.; Jamrozy, M.; Janiak, M.; Kasperek, J.; Lalik, K.; Mach, E.; Mandat, D.; Marszałek, A.; Michałowski, J.; Moderski, R.; Montaruli, T.; Niemiec, J.; Ostrowski, M.; Paśko, P.; Pech, M.; Porcelli, A.; Rameez, M.; Rajda, P.; Schovanek, P.; Seweryn, K.; Skowron, K.; Sliusar, V.; Sowiński, M.; Stawarz, Ł.; Stodulska, M.; Stodulski, M.; Toscano, S.; Pujadas, I. Troyano; Walter, R.; Więcek, M.; Zagdański, A.; Ziętara, K.; Żychowski, P.

    2015-01-01

    The SST-1M telescope is one of the prototypes under construction proposed to be part of the future Cherenkov Telescope Array. It uses a standard Davis-Cotton design for the optics and telescope structure, with a dish diameter of 4 meters and a large field-of-view of 9 degrees. The innovative camera design is composed of a photo-detection plane with 1296 pixels including entrance window, light concentrators, Silicon Photomultipliers (SiPMs), and pre-amplifier stages together with a fully digital readout and trigger electronics, DigiCam. In this contribution we give a general description of the analysis chain designed for the SST-1M prototype. In particular we focus on the calibration strategy used to convert the SiPM signals registered by DigiCam to the quantities needed for Cherenkov image analysis. The calibration is based on an online feedback system to stabilize the gain of the SiPMs, as well as dedicated events (dark count, pedestal, and light flasher events) to be taken during the normal operation of the...

  3. Instrumentation for status monitoring and protection of SST-1 superconducting magnets

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, A.N., E-mail: aashoo.sharma@yahoo.com; Prasad, U.; Doshi, K.; Varmora, P.; Khristi, Y.; Patel, D.; Pradhan, S.

    2016-11-15

    Highlights: • Details of status monitoring instrumentation are presented. • Protection instrumentation details are presented. • Instrumentation installation details, signal conditioning and DAQ system details and the results during SST-1 operation are presented. - Abstract: Superconducting magnets of SST-1 are extensively instrumented to continuously monitor the health of magnets during machine cool-down, plasma experiments and also during the machine warm-up phase. These instrumentations include temperature sensors, flow meters, hall probes, strain gages, displacement sensors, pressure sensors and voltage taps. The number of sensors and their locations has been optimized to systematically monitor all important magnet parameters to ensure its safety. In-house developed modular signal conditioning cards have been developed for these instrumentations. The data is acquired on a Versa Module Europa bus based data acquisition system (VME DAQ). This paper gives an overview of selection, installation, laboratory scale validations, and distribution logics of these instrumentations. Results during plasma campaigns and the up-gradation aspects of these instrumentations are also discussed in this paper.

  4. Development of laser-based technology for the routine first wall diagnostic on the tokamak EAST: LIBS and LIAS

    Science.gov (United States)

    Hu, Z.; Gierse, N.; Li, C.; Liu, P.; Zhao, D.; Sun, L.; Oelmann, J.; Nicolai, D.; Wu, D.; Wu, J.; Mao, H.; Ding, F.; Brezinsek, S.; Liang, Y.; Ding, H.; Luo, G.; Linsmeier, C.; EAST Team

    2017-12-01

    A laser based method combined with spectroscopy, such as laser-induced breakdown spectroscopy (LIBS) and laser-induced ablation spectroscopy (LIAS), is a promising technology for plasma-wall interaction studies. In this work, we report the development of in situ laser-based diagnostics (LIBS and LIAS) for the assessment of static and dynamic fuel retention on the first wall without removing the tiles between and during plasma discharges in the Experimental Advanced Superconducting Tokamak (EAST). The fuel retention on the first wall was measured after different wall conditioning methods and daily plasma discharges by in situ LIBS. The result indicates that the LIBS can be a useful tool to predict the wall condition in EAST. With the successful commissioning of a refined timing system for LIAS, an in situ approach to investigate fuel retention is proposed.

  5. CONTROL SYSTEM FOR THE LITHIUM BEAM EDGE PLASMA CURRENT DENSITY DIAGNOSTIC ON THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    PEAVY, J.J.; CARY, W.P; THOMAS, D.M; KELLMAN, D.H.; HOYT, D.M; DELAWARE, S.W.; PRONKO, S.G.E.; HARRIS, T.E.

    2004-03-01

    OAK-B135 An edge plasma current density diagnostic employing a neutralized lithium ion beam system has been installed on the DIII-D tokamak. The lithium beam control system is designed around a GE Fanuc 90-30 series PLC and Cimplicity(reg s ign) HMI (Human Machine Interface) software. The control system operates and supervises a collection of commercial and in-house designed high voltage power supplies for beam acceleration and focusing, filament and bias power supplies for ion creation, neutralization, vacuum, triggering, and safety interlocks. This paper provides an overview of the control system, while highlighting innovative aspects including its remote operation, pulsed source heating and pulsed neutralizer heating, optimizing beam regulation, and beam ramping, ending with a discussion of its performance

  6. Analysis and Performance of the Thomson Scattering Diagnostics on HT-7 Tokamak Based on I-EMCCD

    International Nuclear Information System (INIS)

    Shao Chunqiang; Zhao Junyu; Zang Qing; Han Xiaofeng; Xi Xiaoqi; Yang Jianhua; Chen Hui; Hu Ailan

    2014-01-01

    A visible light imaging Thomson scattering (VIS-TVTS) diagnostic system has been developed for the measurement of plasma electron temperature on the HT-7 tokamak. The system contains a Nd:YAG laser (λ = 532 nm, repetition rate 10 Hz, total pulse duration ≍ 10 ns, pulse energy > 1.0 J), a grating spectrometer, an image intensifier (I.I.) lens coupled with an electron multiplying CCD (EMCCD) and a data acquisition and analysis system. In this paper, the measurement capability of the system is analyzed. In addition to the performance of the system, the capability of measuring plasma electron temperature has been proved. The profile of electron temperature is presented with a spatial resolution of about 0.96 cm (seven points) near the center of the plasma

  7. High-resolution spectroscopy diagnostics for measuring impurity ion temperature and velocity on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Weinzettl, Vladimír; Shukla, G.; Ghosh, J.; Melich, Radek; Pánek, Radomír; Tomeš, Matěj; Imríšek, Martin; Naydenkova, Diana; Varju, Jozef; Pereira, T.; Gomes, R.; Abramovic, I.; Jaspers, R.; Písařík, M.; Odstrčil, T.; Van Oost, G.

    96-97, October (2015), s. 1006-1011 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA ČR(CZ) GA14-35260S; GA ČR GAP205/11/2341; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Tokamak * Plasma spectroscopy * Plasma rotation * Ion temperature * CXRS Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://www.sciencedirect.com/science/article/pii/S0920379615002355

  8. Cool Down Experiences with the SST-1 Helium Cryogenics System before and after Current Feeders System Modification

    Science.gov (United States)

    Patel, R.; Panchal, P.; Panchal, R.; Tank, J.; Mahesuriya, G.; Sonara, D.; Srikanth, G. L. N.; Garg, A.; Bairagi, N.; Christian, D.; Patel, K.; Shah, P.; Nimavat, H.; Sharma, R.; Patel, J. C.; Gupta, N. C.; Prasad, U.; Sharma, A. N.; Tanna, V. L.; Pradhan, S.

    The SST-1 machine comprises a superconducting magnet system (SCMS), which includes TF and PF magnets. In order to charge the SCMS, we need superconducting current feeders consisting of SC feeders and vapor cooled current leads (VCCLs). We have installed all 10 (+/-) pairs of VCCLs for the TF and PF systems. While conducting initial engineering validation of the SST-1 machine, our prime objective was to produce circular plasma using only the TF system. During the SST-1 campaign I to VI, we have to stop the PF magnets cooling in order to get the cryo- stable conditions for current charging of the TF magnets system. In that case, the cooling of the PF current leads is not essential. It has been also observed that after aborting the PF system cooling, there was a limited experimental window of TF operation. Therefore, in the recent SST-1 campaign-VII, we removed the PF current leads (9 pairs) and kept only single (+/-) pair of the 10,000 A rated VCCLs to realize the charging of the TF system for the extended window of operation. We have observed a better cryogenic stability in the TF magnets after modifications in the CFS. In this paper, we report the comparison of the cool down performance for the SST-1 machine operation before and after modifications of the current feeders system.

  9. Ion temperature profiles along a hydrogen diagnostic beam in a TORE SUPRA tokamak plasma

    International Nuclear Information System (INIS)

    Romannikov, A.; Petrov, Yu.; Platts, P.; Khess, V.; Khutter, T.; Farzhon, Zh.; Moro, F.

    2002-01-01

    By means of corpuscular diagnostics one studies temperature of ions along a diagnostic hydrogen beam. Paper presents comparison of temperature of plasma (deuterium) basic ions measures by means of the active corpuscular diagnostics with temperature of C + carbon ions along a beam. One studies behavior peculiarities of T i ion temperature profiles for TORE-SUPRA different modes, such as: formation of plane and even hollow T i profiles for ohmic modes, variation of T i profiles under operation of an ergodic diverter, difference of temperature of basic ions measured by means of the active corpuscular diagnostics from C +5 temperature. Paper offers clear explanation of these peculiarities [ru

  10. Silicon drift detector based X-ray spectroscopy diagnostic system for the study of non-thermal electrons at Aditya tokamak.

    Science.gov (United States)

    Purohit, S; Joisa, Y S; Raval, J V; Ghosh, J; Tanna, R; Shukla, B K; Bhatt, S B

    2014-11-01

    Silicon drift detector based X-ray spectrometer diagnostic was developed to study the non-thermal electron for Aditya tokamak plasma. The diagnostic was mounted on a radial mid plane port at the Aditya. The objective of diagnostic includes the estimation of the non-thermal electron temperature for the ohmically heated plasma. Bi-Maxwellian plasma model was adopted for the temperature estimation. Along with that the study of high Z impurity line radiation from the ECR pre-ionization experiments was also aimed. The performance and first experimental results from the new X-ray spectrometer system are presented.

  11. Machine Control System of Steady State Superconducting Tokamak-1

    Energy Technology Data Exchange (ETDEWEB)

    Masand, Harish, E-mail: harish@ipr.res.in; Kumar, Aveg; Bhandarkar, M.; Mahajan, K.; Gulati, H.; Dhongde, J.; Patel, K.; Chudasma, H.; Pradhan, S.

    2016-11-15

    Highlights: • Central Control System. • SST-1. • Machine Control System. - Abstract: Central Control System (CCS) of the Steady State Superconducting Tokamak-1 (SST-1) controls and monitors around 25 plant and experiment subsystems of SST-1 located remotely from the Central-Control room. Machine Control System (MCS) is a supervisory system that sits on the top of the CCS hierarchy and implements the CCS state diagram. MCS ensures the software interlock between the SST-1 subsystems with the CCS, any subsystem communication failure or its local error does not prohibit the execution of the MCS and in-turn the CCS operation. MCS also periodically monitors the subsystem’s status and their vital process parameters throughout the campaign. It also provides the platform for the Central Control operator to visualize and exchange remotely the operational and experimental configuration parameters with the sub-systems. MCS remains operational 24 × 7 from the commencement to the termination of the SST-1 campaign. The developed MCS has performed robustly and flawlessly during all the last campaigns of SST-1 carried out so far. This paper will describe various aspects of the development of MCS.

  12. Development of a Cherenkov-type diagnostic system to study runaway electrons within the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Rabinski, M.; Jakubowski, L.; Malinowski, K.; Sadowski, M. J.; Zebrowski, J.; Jakubowski, M.J.; Mirowski, R.; Weinzettl, Vladimír; Ficker, Ondřej; Mlynář, Jan; Pánek, Radomír; Papřok, Richard; Vlainic, Milos

    2017-01-01

    Roč. 12, October (2017), č. článku C10014. ISSN 1748-0221. [European Conference on Plasma Diagnostics (ECPD2017)/2./. Bordeaux, 18.04.2017-21.04.2017] R&D Projects: GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : Nuclear instruments and methods for hot plasma diagnostics * Plasma diagnostics - probes Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.220, year: 2016 http://iopscience.iop.org/article/10.1088/1748-0221/12/10/C10014

  13. Up gradation of VME based data acquisition for SST-1 superconducting magnets

    International Nuclear Information System (INIS)

    Varmora, Pankaj; Parghi, Bhadresh; Banaudha, Moni; Prasad, Upendra

    2017-01-01

    SST-1 magnet system consists of sixteen Toroidal Field (TF) coils and nine Poloidal Field (PF) superconducting coils along with a pair of vertical field coils and air core ohmic transformer. These magnets are instrumented with various cryogenic compatible sensors and voltage taps for its monitoring, operation, protection, and control during different machine operational scenarios like cryogenic cool down, current charging cycles including ramp up, flat top, plasma breakdown, dumping/ramp down and warm up. A VME hardware based data acquisition system has been developed for data monitoring, acquisition and control of magnet operation. A java platform based client and server utility has been developed for this data acquisition system. Upgradation of this java software utility has been carried out with enhance features, fast operating performance and new tools additions. Upgradation features includes larger data file sizes, highlights of critical data indicators, new file generation, online mass flow calculations etc. This poster describes basis hardware details, upgradation of previous software utility, testing and troubleshooting during software development. (author)

  14. Conceptual design of dump resistor for superconducting CS of SST-1

    International Nuclear Information System (INIS)

    Roy, Swati; Pradhan, Subrata; Panchal, Arun

    2015-01-01

    During the upgradation of SST-1, the resistive central solenoid (CS) coil has been planned to be replaced with Nb 3 Sn based superconducting coil. The superconducting CS will store upto 3.5MJ of magnetic energy per operation cycle with operating current upto 14kA. In case of coil quench, the energy stored in the coils is to be extracted rapidly with a time constant of 1.5s. This will be achieved by inserting a 20m Ohm dump resistor in series with the superconducting CS which is normally shorted by circuit breakers. As a vital part of the superconducting CS quench protection system, a conceptual design of the 20m Ohm dump resistor has been proposed. In this paper, the required design aspects and a dimensional layout of the dump resistor for the new superconducting CS has been presented. Natural air circulation is proposed as cooling method for this dump resistor. The basic structure of the proposed dump resistor comprises of stainless steel grids connected in series in the shape of meander to minimize the stray inductance and increase the surface area for cooling. The entire dump resistor will be an array of such grids connected in series and parallel to meet electrical as well as thermal parameters. The maximum temperature of the proposed dump resistor is upto 350 °C during dump 3.5MJ energy. The proposed design permits indigenous fabrication of the dump resistor using commercially available welding techniques. (author)

  15. Conceptual design of Dump resistor for Superconducting CS of SST-1

    Science.gov (United States)

    Roy, Swati; Raj, Piyush; Panchal, Arun; Pradhan, Subrata

    2017-04-01

    Under upgradation activities for SST-1, the existing resistive central solenoid (CS) coil will be replaced with Nb3Sn based superconducting coil. Design of Central solenoid had been completed and some of the initiative has already taken for its manufacturing. The superconducting CS will store upto 3 MJ of magnetic energy per operation cycle with operating current upto 14 kA. During quench, energy stored in the coils has to be extracted rapidly with a time constant of 1.5 s by inserting a 20 mΩ dump resistor in series with the superconducting CS which is normally shorted by circuit breakers. As a critical part of the superconducting CS quench protection system, a conceptual design of the 20 mΩ dump resistor has been proposed. The required design aspects and a dimensional layout of the dump resistor for the new superconducting CS has been presented and discussed. The basic structure of the proposed dump resistor comprises of stainless steel grids connected in series in the form of meander to minimize the stray inductance and increase the surface area for cooling. Such an array of grids connected in series and parallel will cater to the electrical as well as thermal parameters. It will be cooled by natural convection. During operation, the estimated maximum temperature of the proposed dump resistor will raise upto 600 K.

  16. Immunohistochemical detection of somatostatin receptor subtypes sst1 and sst2A in human somatostatin receptor positive tumors

    NARCIS (Netherlands)

    L.J. Hofland (Leo); Q. Liu; P.M. van Koetsveld (Peter); J. Zuijderwijk; F. van der Ham (Frieda); R.R. de Krijger (Ronald); A. Schonbrunn; S.W.J. Lamberts (Steven)

    1999-01-01

    textabstractAlthough in situ hybridization has been used to examine the distribution of messenger RNA for somatostatin receptor subtypes (sst) in human tumors, the cellular localization of sst1 and sst2A receptors has not been reported. In this study, we describe the

  17. On the survivability of diagnostic windows in the CIT [Compact Ignition Tokamak] reactor

    International Nuclear Information System (INIS)

    Taylor, A.

    1988-11-01

    The problem of radiation induced stresses in CIT diagnostic windows is discussed. Existing data indicate windows of existing design will probably survive if placed on the periphery of the reactor. There is a lack of adequate engineering data upon which the design and survivability of windows can be based. 22 refs., 5 figs., 2 tabs

  18. Proposed experiment to investigate use of heated optical fibers for tokamak diagnostics during D-T discharges

    International Nuclear Information System (INIS)

    Tighe, W.; Morgan, P.; Griscom, D.; Adler, H.; Cylinder, D.; Johnson, D.; Palladino, D.; Ramsey, A.

    1995-02-01

    A collaborative JET/TFTR study has been undertaken to investigate attenuation and luminescence effects due to neutron irradiation of optical fibers heated to 400 degrees C. It is expected that a significant improvement in fiber behavior will be observed due to thermal annealing. This technique may be important for use in fiber-related, tokamak diagnostics exposed to high neutron flux. The study will make use of aluminum jacketed, 600 μm diameter, all silica (F-doped cladding) fibers in lengths of 150 m. The fibers are prepared in 1 foot coils. Of the coils to be irradiated, one is heated constantly to 400 degrees C, a second is not heated, and a third is heated periodically. A fourth fiber coil is not to be irradiated. Spectrally and temporally resolved transmission and luminescence data under neutron irradiation during D-T discharges on TFTR will be obtained. An investigation of permanent and short term effects will be made. Experimental details along with initial results will be presented

  19. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  20. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  1. Ultra high vacuum compatible microwave beam launcher for ECRH in SST - 1

    International Nuclear Information System (INIS)

    Shukla, B.K.; Sathyanarayana, K.; Biswas, P.; Pragnesh, D.; Bora, D.

    2005-01-01

    Microwave beam launcher for Electron Cyclotron Resonance Heating (ECRH) system is used to focus the microwave beam at plasma center of SST -1. The beam launcher consists of an ultra high vacuum (UHV) compatible mirror box with two mirrors mounted in it. One mirror is focusing mirror while other one is a plane mirror. The total volume of the launcher is ∼ 60000 cc and the total surface area exposed to UHV is around ∼ 1.0x10 4 cm 2 . The mirrors are cooled with water for high power and long pulse operation. UHV compatible SS hoses provide flexible cooling connection to the mirrors. Flexible cooling connection helps in adjustment and steering of the mirrors. SS hoses are welded at both the ends and this is necessary to avoid any flange connection inside ultra high vacuum. The system has been tested for UHV compatibility. The leak rate is checked with helium leak detector and found better than l x 10 -9 mbar.lt/s. The system has been baked to 150 deg C for ∼14 hours and the ultimate vacuum achieved with turbomolecular pump (TMP) is ∼ 5x10 -9 mbar. The mirror assembly is tested for leak in pressurized condition using a sniffer probe. The mirrors of the launcher along with the welded bellow are pressurized with helium gas up to a water equivalent pressure of ∼3kg/cm 2 . No increase in the background (∼-10 -6 mbar.lt/s) of the sniffer probes has been observed during the test. The plane mirror is connected with two UHV linear motion feedthroughs with suitable hinges and smooth movement is checked in vacuum. (author)

  2. Challenges of video monitoring for phenomenological diagnostics in present and future tokamaks

    International Nuclear Information System (INIS)

    Martin, Vincent; Mondana, Victor; Travere, Jean-Marcel

    2011-01-01

    With the development of heterogeneous camera networks working at different wavelengths and frame rates and covering a large surface of vacuum vessel, the visual observation of a large variety of plasma and thermal phenomena (e.g., hot spots, ELMs, MARFE, arcs, dusts, etc.) becomes possible. In the domain of machine protection, a phenomenological diagnostic is a key-element towards plasma/thermal event dangerousness assessment during real time operation. It is also of primary importance to automate the extraction and the storage of phenomena information for further off-line event retrieval and analysis, thus leading to a better use of massive image data bases for plasma physics studies. To this end, efforts have been devoted to the development of image processing algorithms dedicated to the recognition of specific events. But a need arises now for the integration of techniques developed so far in both hardware and software directions. We present in this paper our latest results in the field of real time phenomena recognition and management through our image understanding software platform. T his platform has been validated on Tore Supra during operation and is under evaluation for the foreseen imaging diagnostic of ITER. (authors)

  3. Performance, diagnostics, controls and plans for the gyrotron system on the DIII-D tokamak

    Directory of Open Access Journals (Sweden)

    Ponce D.M.

    2012-09-01

    Full Text Available The DIII-D ECH complex is being upgraded with three new depressed collector gyrotrons. The performance of the existing system has been very good. As more gyrotrons having higher power are added to the system, diagnostics of gyrotron operation, optimization of the performance and qualification of components for higher power become more important. A new FPGA-based gyrotron control system is being installed, additional capabilities for rapid real time variation of the rf injection angles by the DIII-D Plasma Control System are being tested and infrastructure enhancements are being completed. Longer term plans continue to include ECH as a major component in the DIII-D heating and current drive capabilities.

  4. Application of optical fluctuation diagnostics to transport studies in high temperature tokamak plasmas

    International Nuclear Information System (INIS)

    Fonck, R.J.

    1991-01-01

    First data was obtained from the TFTR BES system in May 1990, after the prototype 4-channel system with photomultiplier detectors from PBX-M was transferred to the TFTR Hot Cell. The system was expanded to 16 spatial channels during the Summer 1990 run period, and data of interest was obtained near the end of the run in October 1990. The most attention has been given to a radial profile data set obtained for an L-mode plasma condition in TFTR, where we characterized the nature of the long-wavelength turbulence seen by the BES diagnostic. We summarize here highlights of the initial run with BES or TFTR. We also give a progress report on support activities at UW

  5. Instrument to synchronize Thomson scattering diagnostic measurements with MHD acitivity in a tokamak

    International Nuclear Information System (INIS)

    Wintenberg, A.L.

    1985-04-01

    An instrument to synchronize the firing of a ruby laser for a Thomson scattering diagnostic with plasma oscillations was designed, developed, and evaluated. The instrument will fire the laser at a user-selected phase of an input sine or sawtooth wave with an accuracy of +-15 0 . Allowable frequencies range from 20 to 500 Hz for a sawtooth and from 1 to 30 kHz for a sine wave. The instrument also allows synchronization with a sine wave to be enabled by a preselected sawtooth phase. The instrument uses analog signal processing circuits to separate the signal components, remove unwanted components, and produce zero-phase synchronization pulses. The instrument measures the period between zero-phase pulses in order to produce phase synchronization pulses delayed a fraction of the period from the zero-phase pulses. The laser is fired by the phase synchronization pulse. Unwanted signal components are attenuated by bandpass filters. A digitally controlled self-adjusting bandpass filter for sine processing. The instrument was used to investigate the variation of the electron temperature profile with the phase of the x-ray signal from an Impurity Studies Experiment (ISX-B) plasma exhibiting magnetohydrodynamic (MHD) activity

  6. The general optics structure of millimeter-wave imaging diagnostic on TOKAMAK

    International Nuclear Information System (INIS)

    Zhu, Y.; Xie, J.; Liu, W.D.; Luo, C.; Zhao, Z.; Chen, D.; Domier, C.W.; Luhmann, N.C. Jr.; Chen, M.; Hu, X.

    2016-01-01

    Advanced imaging optics techniques have significantly improved the performance of millimeter-wave imaging diagnostics, such as Electron Cyclotron Emission imaging and Microwave Imaging of Reflectometry. The fundamental functions of millimeter-wave imaging optics are focusing, collecting the emission or reflected microwave signal from the target area in the plasma and focusing the emitted (reflected) signal on the detector array. The location of the observation area can be changed using the focus lens. Another important function of the imaging optics is zooming. The size of the observation area in poloidal direction can be adjusted by the zoom lenses and the poloidal spatial resolution is determined by the level of zoom. The field curvature adjustment lenses are employed to adjust the shape of the image plane in the poloidal direction to reduce crosstalk between neighboring channels. The incident angle on each channel is controlled using the specific surface type of the front-side lenses to increase the signal-to-noise ratio. All functions are decoupled with the minimum number of lenses. Successful applications are given

  7. 1-channel wireless acquisition system for magnetic diagnostics of Aditya-U Tokamak

    International Nuclear Information System (INIS)

    Dash, Suvendu Kumar; Mahapatra, Sakuntala; Ali, Shaik Mohammad; Raju, Daniel

    2017-01-01

    In recent years Data Acquisition Systems have proficient advances mainly due to the reduction in cost and gaining functionally of systems based on microcontrollers and microcomputers. This paper invites a prototype model of one channel wireless data acquisition system. The system contains a ratiometric linear Hall Effect sensor, embedded system with Atmeg328 microcontroller for both transmitter and receiver and wireless transceiver module NRF24L01+. The readings from the ratiometric linear Hall Effect sensor, IC - A1301KUA-T are digitized by in built A/D converter present in the embedded system then they are sent to the wireless transceiver NRF24L01+. NRF24L01+ performs GFSK modulation technique for transmission of the digital data. When the bar magnet is kept close to the sensor we found 448 gauss (0.0448Tesla). We received the same transmitted data without any error. Moreover we show the results from Hall Effect sensors mounted in 3-axis perpendicular to each other and observations of Hall Effect sensor in presence of High voltage. Helmholtz coil experimental results validated the Hall Effect Sensor used it for magnetic diagnostics. (author)

  8. Characterization and commissioning of the SST-1M camera for the Cherenkov Telescope Array

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar, J.A. [Université Libre Bruxelles, Faculté des Sciences, Avenue Franklin Roosevelt 50, 1050 Brussels (Belgium); DPNC - Université de Genéve, 24 Quai Ernest Ansermet, Genéve (Switzerland); Department of Information Technologies, Jagiellonian University, ul. prof. Stanisława Łojasiewicza 11, 30–348 Kraków (Poland); Bilnik, W. [AGH University of Science and Technology, al.Mickiewicza 30, Kraków (Poland); Department of Information Technologies, Jagiellonian University, ul. prof. Stanisława Łojasiewicza 11, 30–348 Kraków (Poland); Błocki, J. [Instytut Fizyki Jadrowej im. H. Niewodniczańskiego Polskiej Akademii Nauk, ul. Radzikowskiego 152, 31–342 Kraków (Poland); Department of Information Technologies, Jagiellonian University, ul. prof. Stanisława Łojasiewicza 11, 30–348 Kraków (Poland); Bogacz, L. [Astronomical Observatory, Jagiellonian University, ul. Orla 171, 30–244 Kraków (Poland); Department of Information Technologies, Jagiellonian University, ul. prof. Stanisława Łojasiewicza 11, 30–348 Kraków (Poland); and others

    2017-02-11

    The Cherenkov Telescope Array (CTA), the next generation very high energy gamma-rays observatory, will consist of three types of telescopes: large (LST), medium (MST) and small (SST) size telescopes. The SSTs are dedicated to the observation of gamma-rays with energy between a few TeV and a few hundreds of TeV. The SST array is expected to have 70 telescopes of different designs. The single-mirror small size telescope (SST-1 M) is one of the proposed telescope designs under consideration for the SST array. It will be equipped with a 4 m diameter segmented mirror dish and with an innovative camera based on silicon photomultipliers (SiPMs). The challenge is not only to build a telescope with exceptional performance but to do it foreseeing its mass production. To address both of these challenges, the camera adopts innovative solutions both for the optical system and readout. The Photo-Detection Plane (PDP) of the camera is composed of 1296 pixels, each made of a hollow, hexagonal light guide coupled to a hexagonal SiPM designed by the University of Geneva and Hamamatsu. As no commercial ASIC would satisfy the CTA requirements when coupled to such a large sensor, dedicated preamplifier electronics have been designed. The readout electronics also use an innovative approach in gamma-ray astronomy by adopting a fully digital approach. All signals coming from the PDP are digitized in a 250 MHz Fast ADC and stored in ring buffers waiting for a trigger decision to send them to the pre-processing server where calibration and higher level triggers will decide whether the data are stored. The latest generation of FPGAs is used to achieve high data rates and also to exploit all the flexibility of the system. As an example each event can be flagged according to its trigger pattern. All of these features have been demonstrated in laboratory measurements on realistic elements and the results of these measurements will be presented in this contribution.

  9. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

    The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

  10. Diagnostic of the spatial and velocity distribution of alpha particles in tokamak fusion reactor using beat-wave generated lower hybrid wave. Progress report, 1994-1995

    International Nuclear Information System (INIS)

    Hwang, D.Q.; Horton, R.D.; Evans, R.

    1995-01-01

    The alpha particle population from fusion reactions in a DT tokamak reactor can have dramatic effects on the pressure profiles, energetic particle confinement, and the overall stability of the plasma; thus leading to important design consideration of a fusion reactor based on the tokamak concept. In order to fully understand the effects of the alpha population, a non-invasive diagnostic technique suitable for use in a reacting plasma environment needs to be developed to map out both the spatial and velocity distribution of the alphas. The proposed experimental goals for the eventual demonstration of LH wave interaction with a fast ion population is given in the reduced 3 year plan in table 1. At present time the authors are approaching the 8th month in their first year of this project. Up to now, their main effort has been concentrated in the operation of the two beat wave sources in burst mode. The second priority in the experimental project is the probe diagnostics and computer aided data acquisition system. The progress made so far is given, and they are ready to perform the beat-wave generated lower hybrid wave experiment. Some theoretical calculation had been reported at APS meetings. More refined theoretical models are being constructed in collaboration with Drs. J. Rogers and E. Valeo at PPPL

  11. A large channel count multi client data acquisition system for superconducting magnet system of SST-1

    International Nuclear Information System (INIS)

    Doshi, K.; Pradhan, S.; Masand, H.; Khristi, Y.; Dhongde, J.; Sharma, A.; Parghi, B.; Varmora, P.; Prasad, U.; Patel, D.

    2012-01-01

    The magnet system of the Steady-state Superconducting Tokamak-1 at the Institute for Plasma Research, Gandhinagar, India, consists of sixteen Toroidal field and nine Poloidal field Superconducting coils together with a pair of resistive PF coils, an air core ohmic transformer and a pair of vertical field coils. These coils are instrumented with various cryogenic grade sensors and voltage taps to monitor its operating status and health during different operational scenarios. A VME based data acquisition system with remote system architecture is implemented for data acquisition and control of the complete magnet operation. Client-Server based architecture is implemented with remote hardware configuration and continuous online/offline monitoring. A JAVA based platform independent client application is developed for data analysis and data plotting. The server has multiple data pipeline architecture to send data to storage database, online plotting application, numerical display screen, and run time calculation. This paper describes software architecture, design and implementation of the data acquisition system. (author)

  12. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  13. The simultaneous measurements of core and outer core density fluctuations in L-H transition using CO2 laser collective scattering diagnostic in the EAST superconducting tokamak

    International Nuclear Information System (INIS)

    Cao, G.M.; Li, Y.D.; Zhang, X.D.; Sun, P.J.; Hu, L.Q.; Li, J.G.; Wu, G.J.

    2013-01-01

    The H-mode is the projected basic operation scenario for the ITER tokamak. The turbulence de-correlation by the synergistic effect of zonal flow and equilibrium ExB flow shear is believed to be the reason for L-H transition, however, the detailed physical mechanism has not been identified so far. Tangential multi-channel CO 2 laser collective scattering diagnostic system (mainly k r measurement) was first installed to investigate electron density fluctuations on EAST tokamak. The measurements in a spontaneous dithering L-H transition show that in core plasma (0 < r/a < 0.5) the low-frequency fluctuations strengthen greatly before L-H transition; meanwhile in outer core plasma (0.2 < r/a < 1) the low-frequency fluctuations strengthen slightly. Bispectral analysis reveals that the coupling strength between low- and high-frequency fluctuations in both core and outer core plasma strengthens greatly before the transition, but the latter is greater than the former. The results indicate that the low-frequency fluctuations of the core and outer core plasma play active, but different, roles in the spontaneous L-H transition. (author)

  14. Investigation of plasma turbulence and local electric field in the T-10 tokamak and TJ-II stellarator by HIBP diagnostic (Review)

    International Nuclear Information System (INIS)

    Krupnik, L.I.; Chmuga, A.A.; Komarov, A.D.; Kozachek, A.S.; Zhezhera, A.I.; Melnikov, A.V.; Eliseev, L.G.; Lysenko, S.E.; Mavrin, V.A.; Perfilov, S.V.; Hidalgo, C.; Ascasibar, E.; Estrada, T.; Ochando, M.A.; Pablos, J.L.; Pedrosa, M.A.; Tabares, F.

    2011-01-01

    Direct study of the electric potential and its fluctuations for comparable plasma conditions in the T-10 tokamak and TJ-II stellarator by HIBP diagnostics has been performed. The following similar features of potential were found: the scale of several hundred Volts; the negative sign for densities n e >1x10 19 m -3 and comparable values in spite of the different heating methods. When ne or τ E rises, the potential evolves to negative values. During ECR heating and associated T e rise, τ E degrades and the potential evolves to positive direction. Oscillations of potential and density in the range of Geodesic Acoustic Modes in T-10 and Alfven Eigenmodes in TJ-II were observed.

  15. Adopted Methodology for Cool-Down of SST-1 Superconducting Magnet System: Operational Experience with the Helium Refrigerator

    Science.gov (United States)

    Sahu, A. K.; Sarkar, B.; Panchal, P.; Tank, J.; Bhattacharya, R.; Panchal, R.; Tanna, V. L.; Patel, R.; Shukla, P.; Patel, J. C.; Singh, M.; Sonara, D.; Sharma, R.; Duggar, R.; Saxena, Y. C.

    2008-03-01

    The 1.3 kW at 4.5 K helium refrigerator / liquefier (HRL) was commissioned during the year 2003. The HRL was operated with its different modes as per the functional requirements of the experiments. The superconducting magnets system (SCMS) of SST-1 was successfully cooled down to 4.5 K. The actual loads were different from the originally predicted boundary conditions and an adjustment in the thermodynamic balance of the refrigerator was necessary. This led to enhanced capacity, which was achieved without any additional hardware. The required control system for the HRL was tuned to achieve the stable thermodynamic balance, while keeping the turbines' operating parameters at optimized conditions. An extra mass flow rate requirement was met by exploiting the margin available with the compressor station. The methodology adopted to modify the capacity of the HRL, the safety precautions and experience of SCMS cool down to 4.5 K, are discussed.

  16. Development of the optical system for the SST-1M telescope of the Cherenkov Telescope Array observatory

    CERN Document Server

    Ostrowski, Michael; Błocki, J.; Bogacz, L.; Bulik, T.; Cadoux, F.; Christov, A.; Curyło, M.; della Volpe, D.; Dyrda, M.; Favre, Y.; Frankowski, A.; Grudnik, Ł.; Grudzińska, M.; Heller, M.; Idźkowski, B.; Jamrozy, M.; Janiak, M.; Kasperek, J.; Lalik, K.; Lyard, E.; Mach, E.; Mandat, D.; Marszałek, A.; Michałowski, J.; Moderski, R.; Montaruli, T.; Neronov, A.; Niemiec, J.; Paśko, P.; Pech, M.; Porcelli, A.; Prandini, E.; Pueschel, E.; Rajda, P.; Rameez, M.; Schioppa, E. jr; Schovanek, P.; Skowron, K.; Sliusar, V.; Sowiński, M.; Stawarz, Ł.; Stodulska, M.; Stodulski, M.; Toscano, S.; Troyano Pujadas, I.; Walter, R.; Wiȩcek, M.; Zagdański, A.; Ziȩtara, K.; Żychowski, P.; Barciński, T.; Karczewski, M.; Kukliński, J. Nicolau; Płatos, Ł.; Rataj, M.; Wawer, P.; Wawrzaszek, R.

    2016-01-01

    The prototype of a Davies-Cotton small size telescope (SST-1M) has been designed and developed by a consortium of Polish and Swiss institutions and proposed for the Cherenkov Telescope Array (CTA) observatory. The main purpose of the optical system is to focus the Cherenkov light emitted by extensive air showers in the atmosphere onto the focal plane detectors. The main component of the system is a dish consisting of 18 hexagonal mirrors with a total effective collection area of 6.47 m2 (including the shadowing and estimated mirror reflectivity). Such a solution was chosen taking into account the analysis of the Cherenkov light propagation and based on optical simulations. The proper curvature and stability of the dish is ensured by the mirror alignment system and the isostatic interface to the telescope structure. Here we present the design of the optical subsystem together with the performance measurements of its components.

  17. Research using small tokamaks

    International Nuclear Information System (INIS)

    1993-01-01

    This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs

  18. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1990-04-01

    This paper discusses the following work on the text tokamak: data systems; particle confinement; impurity transport; plasma rotation; runaway electrons; electron cyclotron heating; FIR system; transient transport; internal turbulence; edge turbulence; ion temperature; EML experiments; impurity pellet experiments; MHD experiments and analysis; TEXT Upgrade; and Upgrade diagnostics

  19. Development of plasma diagnostics technologies - Measurement of transport= parameters in tokamak edge plasma by using electric transport probes

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kyu Sun; Chang, Do Hee; Sim, Yeon Gun; Kim, Jin Hee [Hanyang University, Seoul (Korea, Republic of)

    1995-08-01

    Electric transport probe system is developed for the measurement of electron temperature, floating potential, plasma density and flow velocity of= edge plasmas in the KT-2 medium size tokamak. Experiments have been performed in KT-1 small size tokamak. Electric transport probe is composed of a single probe(SP) and a Mach probe (MP). SP is used for the measurements of electron density, floating potential, and plasma density and measured values are {approx} 3*10{sup 11}/cm{sup -3}, -20 volts, 15 {approx} 25 eV. For the most discharges, respectively. MP is for the measurements of toroidal(M{sub T}) and poloidal(M{sub P}) flow velocities, and density, which are M{sub T} {approx_equal} .0.85, M{sub P} {approx_equal}. 0.17, n. {approx_equal} 2.1*10{sup 11} cm{sup -3}, respectively. A triple probe is also developed for the direct reading of T{sub e} and n{sub e}, and is used for DC, RF, and RF+DC plasma in APL of Hanyang university. 38 refs., 36 figs. (author)

  20. Application of Thomson scattering at 1.06μm as a diagnostic for spatial profile measurements of electron temperature and density on the TCV tokamak

    International Nuclear Information System (INIS)

    Franke, S.

    1997-04-01

    The variable configuration tokamak, TCV, in operation at CRPP since the end of 1991, is a particularly challenging machine with regard to the experimental system that must provide essential information regarding properties of confined plasmas with strongly shaped, non-circular cross-sections. The importance of the energy confinement issue in a machine designed specifically for the investigation of the effect of plasma shape on confinement and stability is self-evident, as is the necessity for a diagnostic capable of providing the profiles of electron temperature and density required for evaluation of this confinement. For TCV, a comprehensive Thomson Scattering (TS) diagnostic was the natural choice, specifically owing to the resulting spatially localized and time resolved measurement. The details of the system installed on TCV, together with the results obtained from the diagnostic comprise the subject matter of this thesis. A first version of the diagnostic was equipped with only ten observation volumes. In this case, adequate spatial resolution can only be maintained if measurements are limited to plasmas located in the upper half of the highly elongated TCV vacuum vessel. The system has recently been upgraded through the addition of a further fifteen observation volumes, together with major technical improvements in the scattered light detection system. This new version now permits TS observations in all TCV plasma configurations, including equilibria produced in the lower and upper halves of the vacuum vessel and the highly elongated plasmas now routinely created. Whilst a description of the new detection system along with some results obtained using the extended set of observation volumes are included, this thesis reports principally on the hardware details of and the interpretation of data from the original, ten observation volume system. (author) figs., tabs., 75 refs

  1. Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, C. M., E-mail: coopercm@fusion.gat.com [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Pace, D. C.; Paz-Soldan, C.; Eidietis, N. W. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Commaux, N.; Shiraki, D. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California, San Diego, La Jolla, California 92093-0533 (United States)

    2016-11-15

    A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5–100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead “pinhole camera” mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20 000 pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.

  2. Applying the new gamma ray imager diagnostic to measurements of runaway electron Bremsstrahlung radiation in the DIII-D Tokamak (invited)

    International Nuclear Information System (INIS)

    Cooper, C. M.; Pace, D. C.; Paz-Soldan, C.; Eidietis, N. W.; Commaux, N.; Shiraki, D.; Hollmann, E. M.

    2016-01-01

    A new gamma ray imager (GRI) is developed to probe the electron distribution function with 2D spatial resolution during runaway electron (RE) experiments at the DIII-D tokamak. The diagnostic is sensitive to 0.5–100 MeV gamma rays, allowing characterization of the RE distribution function evolution during RE growth and dissipation. The GRI consists of a lead “pinhole camera” mounted on the DIII-D midplane with 123 honeycombed tangential chords 20 cm wide that span the vessel interior. Up to 30 bismuth germanate (BGO) scintillation detectors capture RE bremsstrahlung radiation for Pulse Height Analysis (PHA) capable of discriminating up to 20 000 pulses per second. Digital signal processing routines combining shaping filters are performed during PHA to reject noise and record gamma ray energy. The GRI setup and PHA algorithms will be described and initial data from experiments will be presented. A synthetic diagnostic is developed to generate the gamma ray spectrum of a GRI channel given the plasma information and a prescribed distribution function. Magnetic reconstructions of the plasma are used to calculate the angle between every GRI sightline and orient and discriminate gamma rays emitted by a field-aligned RE distribution function.

  3. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

    For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

  4. 68Ga-DOTATOC PET/CT and somatostatin receptor (sst1-sst5) expression in normal human tissue: correlation of sst2 mRNA and SUVmax

    International Nuclear Information System (INIS)

    Boy, Christian; Poeppel, Thorsten D.; Jentzen, Walter; Brandau, Wolfgang; Bockisch, Andreas; Heusner, Till A.; Antoch, Gerald; Redmann-Bischofs, Anja; Unger, Nicole; Mann, Klaus; Petersenn, Stephan

    2011-01-01

    By targeting somatostatin receptors (sst) radiopeptides have been established for both diagnosis and therapy. For physiologically normal human tissues the study provides a normative database of maximum standardized uptake value (SUV max ) and sst mRNA. A total of 120 patients were subjected to diagnostic 68 Ga-DOTATOC positron emission tomography (PET)/CT (age range 19-83 years). SUV max values were measured in physiologically normal tissues defined by normal morphology, absence of surgical intervention and absence of metastatic spread during clinical follow-up. Expression of sst subtypes (sst1-sst5) was measured independently in pooled adult normal human tissue by real-time reverse transcriptase polymerase chain reaction (RT-PCR). SUV max revealed a region-specific pattern (e.g., mean ± SD, spleen 31.1 ± 10.9, kidney 16.9 ± 5.3, liver 12.8 ± 3.6, stomach 7.0 ± 3.1, head of pancreas 6.2 ± 2.3, small bowel 4.8 ± 1.8, thyroid 4.7 ± 2.2, bone 3.9 ± 1.3, large bowel 2.9 ± 0.8, muscle 2.1 ± 0.5, parotid gland 1.9 ± 0.6, axillary lymph node 0.8 ± 0.3 and lung 0.7 ± 0.3). SUV max was age independent. Gender differences were evident within the thyroid (female/male: 3.7 ± 1.6/5.5 ± 2.4, p max values exclusively correlated with sst2 expression (r = 0.846, p max with the expression of the other four subtypes. In normal human tissues 68 Ga-DOTATOC imaging has been related to the expression of sst2 at the level of mRNA. The novel normative database may improve diagnostics, monitoring and therapy of sst-expressing tumours or inflammation on a molecular basis. (orig.)

  5. Application and Continued Development of Thin Faraday Collectors as a Lost Ion Diagnostic for Tokamak Fusion Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    F. Ed Cecil

    2011-06-30

    This report summarizes the accomplishment of sixteen years of work toward the development of thin foil Faraday collectors as a lost energetic ion diagnostic for high temperature magnetic confinement fusion plasmas. Following initial, proof of principle accelerator based studies, devices have been tested on TFTR, NSTX, ALCATOR, DIII-D, and JET (KA-1 and KA-2). The reference numbers refer to the attached list of publications. The JET diagnostic KA-2 continues in operation and hopefully will provide valuable diagnostic information during a possible d-t campaign on JET in the coming years. A thin Faraday foil spectrometer, by virtue of its radiation hardness, may likewise provide a solution to the very challenging problem of lost alpha particle measurements on ITER and other future burning plasma machines.

  6. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  7. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

    With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

  8. The software-defined fast post-processing for GEM soft x-ray diagnostics in the Tungsten Environment in Steady-state Tokamak thermal fusion reactor

    Science.gov (United States)

    Krawczyk, Rafał Dominik; Czarski, Tomasz; Linczuk, Paweł; Wojeński, Andrzej; Kolasiński, Piotr; GÄ ska, Michał; Chernyshova, Maryna; Mazon, Didier; Jardin, Axel; Malard, Philippe; Poźniak, Krzysztof; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol

    2018-06-01

    This article presents a novel software-defined server-based solutions that were introduced in the fast, real-time computation systems for soft X-ray diagnostics for the WEST (Tungsten Environment in Steady-state Tokamak) reactor in Cadarache, France. The objective of the research was to provide a fast processing of data at high throughput and with low latencies for investigating the interplay between the particle transport and magnetohydrodynamic activity. The long-term objective is to implement in the future a fast feedback signal in the reactor control mechanisms to sustain the fusion reaction. The implemented electronic measurement device is anticipated to be deployed in the WEST. A standalone software-defined computation engine was designed to handle data collected at high rates in the server back-end of the system. Signals are obtained from the front-end field-programmable gate array mezzanine cards that acquire and perform a selection from the gas electron multiplier detector. A fast, authorial library for plasma diagnostics was written in C++. It originated from reference offline MATLAB implementations. They were redesigned for runtime analysis during the experiment in the novel online modes of operation. The implementation allowed the benchmarking, evaluation, and optimization of plasma processing algorithms with the possibility to check the consistency with reference computations written in MATLAB. The back-end software and hardware architecture are presented with data evaluation mechanisms. The online modes of operation for the WEST are discussed. The results concerning the performance of the processing and the introduced functionality are presented.

  9. Overview of erosion–deposition diagnostic tools for the ITER-Like Wall in the JET tokamak

    International Nuclear Information System (INIS)

    Rubel, M.; Coad, J.P.; Widdowson, A.; Matthews, G.F.; Esser, H.G.; Hirai, T.; Likonen, J.; Linke, J.; Lungu, C.P.; Mayer, M.; Pedrick, L.; Ruset, C.

    2013-01-01

    This paper presents scientific and technical issues related to the development of erosion–deposition diagnostic tools for JET operated with the ITER-Like Wall: beryllium and tungsten marker tiles and several types of wall probes installed in the main chamber and in the divertor. Markers tiles are the standard limiter and divertor components additionally coated first with a thin sandwich of Ni–Be and Mo–W for, beryllium and tungsten markers, respectively. Both types of markers are embedded in regular arrays of limiter and divertor tiles. Coated W–Be probes are also inserted in the Be-covered Inconel cladding tiles on the central column. Other types of erosion–deposition diagnostic tools are: rotating collectors, deposition traps, louver clips, quartz microbalance and mirrors for the First Mirror Test at JET for ITER. The specific role of these tools is discussed in detail

  10. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  11. Immersive virtual walk-through development for tokamak using active head mounted display

    International Nuclear Information System (INIS)

    Dutta, Pramit

    2015-01-01

    A fully immersive virtual walk-through of the SST-1 tokamak has been developed. The virtual walkthrough renders the virtual model of SST-1 tokamak through a active stereoscopic head mounted display to visualize the virtual environment. All locations inside and outside of the reactor can be accessed and reviewed. Such a virtual walkthrough provides a 1:1 scale visualization of all components of the tokamak. To achieve such a virtual model, the graphical details of the tokamak CAD model are enhanced. Such enhancements are provided to improve lighting conditions at various locations, texturing of components to have a realistic visual effect and 360° rendering for ease of access. The graphical enhancements also include the redefinition of the facets to optimize the surface triangles to remove lags in display during visual rendering. Two separate algorithms are developed to interact with the virtual model. A fly-by algorithm, developed using C#, uses inputs from a commercial joystick to navigate within the virtual environment. The second algorithm uses the IR and gyroscopic tracking system of the head mounted display to render view as per the current pose of the user within the virtual environment and the direction of view. Such a virtual walk-thorough can be used extensively for design review and integration, review of new components, operator training for remote handling, operations, upgrades of tokamak, etc. (author)

  12. Dust Measurements in Tokamaks

    International Nuclear Information System (INIS)

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-01-01

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 (micro)m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics

  13. Local gas injection as a scrape-off layer diagnostic on the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Jablonski, D.F.

    1996-05-01

    A capillary puffing array has been installed on Alcator C-Mod which allows localized introduction of gaseous species in the scrape-off layer. This system has been utilized in experiments to elucidate both global and local properties of edge transport. Deuterium fueling and recycling impurity screening are observed to be characterized by non-dimensional screening efficiencies which are independent of the location of introduction. In contrast, the behavior of non-recycling impurities is seen to be characterized by a screening time which is dependent on puff location. The work of this thesis has focused on the use of the capillary array with a camera system which can view impurity line emission plumes formed in the region of an injection location. The ionic plumes observed extend along the magnetic field line with a comet-like asymmetry, indicative of background plasma ion flow. The flow is observed to be towards the nearest strike-point, independent of x-point location, magnetic field direction, and other plasma parameters. While the axes of the plumes are generally along the field line, deviations are seen which indicate cross-field ion drifts. A quasi-two dimensional fluid model has been constructed to use the plume shapes of the first charge state impurity ions to extract information about the local background plasma, specifically the temperature, parallel flow velocity, and radial electric field. Through comparisons of model results with those of a three dimensional Monte Carlo code, and comparisons of plume extracted parameters with scanning probe measurements, the efficacy of the model is demonstrated. Plume analysis not only leads to understandings of local edge impurity transport, but also presents a novel diagnostic technique

  14. Local gas injection as a scrape-off layer diagnostic on the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jablonski, David F. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1996-05-01

    A capillary puffing array has been installed on Alcator C-Mod which allows localized introduction of gaseous species in the scrape-off layer. This system has been utilized in experiments to elucidate both global and local properties of edge transport. Deuterium fueling and recycling impurity screening are observed to be characterized by non-dimensional screening efficiencies which are independent of the location of introduction. In contrast, the behavior of non-recycling impurities is seen to be characterized by a screening time which is dependent on puff location. The work of this thesis has focused on the use of the capillary array with a camera system which can view impurity line emission plumes formed in the region of an injection location. The ionic plumes observed extend along the magnetic field line with a comet-like asymmetry, indicative of background plasma ion flow. The flow is observed to be towards the nearest strike-point, independent of x-point location, magnetic field direction, and other plasma parameters. While the axes of the plumes are generally along the field line, deviations are seen which indicate cross-field ion drifts. A quasi-two dimensional fluid model has been constructed to use the plume shapes of the first charge state impurity ions to extract information about the local background plasma, specifically the temperature, parallel flow velocity, and radial electric field. Through comparisons of model results with those of a three dimensional Monte Carlo code, and comparisons of plume extracted parameters with scanning probe measurements, the efficacy of the model is demonstrated. Plume analysis not only leads to understandings of local edge impurity transport, but also presents a novel diagnostic technique.

  15. A novel ultra-thin 3D detector-For plasma diagnostics at JET and ITER tokamaks

    International Nuclear Information System (INIS)

    Garcia, Francisco; Pelligrini, G.; Balbuena, J.; Lozano, M.; Orava, R.; Ullan, M.

    2009-01-01

    A novel ultra-thin silicon detector called U3DTHIN has been designed and built for applications that range from Neutral Particle Analyzers (NPA) used in Corpuscular Diagnostics of High Temperature Plasma to very low X-ray spectroscopy. The main purpose of this detector is to provide a state-of-the-art solution to upgrade the current detector system of the NPAs at JET and also to pave the road for the future detection systems of the ITER experimental reactor. Currently the NPAs use a very thin scintillator-photomultiplier tube [F. Garcia, S.S. Kozlovsky, D.V. Balin, Background Properties of CEM, MCP and PMT detectors at n-γ irradiation. Preprint PNPI-2392, Gatchina, 2000, p. 9 ; F. Garcia, S.S. Kozlovsky, V.V. Ianovsky, Scintillation Detectors with Low Sensitivity to n-γ Background. Preprint PNPI-2391, Gatchina, 2000, p. 8 ], and their main drawbacks are poor energy resolution, intrinsic scintillator nonlinearity, and relative low count rate capability and finally poor signal-to-background discrimination for the low-energy channels. The proposed new U3DTHIN detector is based on very thin sensitive substrate, which will provide nearly 100% detection efficiency for ions and at the same time very low sensitivity for neutron and gamma backgrounds. To achieve a very fast collection of the charge carriers generated by the incident ions, a 3D electrode structure [S. Parker, C. Kenney, J. Segal, Nucl. Instr. and Meth. A 395 (1997) 328 ; G. Pellegrini, P. Roy, A. Al-Ajili, R. Bates, L. Haddad, M. Horn, K. Mathieson, J. Melone, V. O'Shea, K.M. Smith, Nucl. Instr. and Meth. A 487 (2002) 19 ] has been introduced in the sensitive volume of the detector. The geometry of the electrode is known to be rad-hard. One of the most innovative features of these detectors is the optimal combination of the thin entrance window and the sensitive substrate thickness, which allows a very large dynamic range for ion detection. GEANT4 simulations were performed to find the losses of energy in

  16. A study on the heating and diagnostic of a tokamak plasma by electromagnetic waves of the electron cyclotron range of frequencies

    International Nuclear Information System (INIS)

    Hoshino, Katsumichi

    1989-09-01

    A study on the heating and diagnosis of tokamak plasma by electromagnetic waves of electron cyclotron range of frequency is summarized. The main results obtained are as follows. On the engineering and technology, the technology of injecting high frequency, large power millimeter waves into tokamak plasma was established by carrying out the design, manufacture and test of a 60 GHz, 400 kW high frequency heating system, and the design, manufacture and test of a heterodyne type electron cyclotron radiation multi-channel mealsuring system were carried out, and the technology of measuring the radiation from tokamak plasma with the time resolution of 10 μs in multi-channel was established. On nuclear fusion reactor core engineering and plasma physics, the high efficiency electron heating of tokamak plasma by the incidence of fundamental irregular and regular waves at electron cyclotron frequency was verified. The discovery and analysis of the heating by electrostatic waves arising due to mode transformation from electromagnetic waves in upper hybrid resonance layer were carried out. By the incidence of second harmonic waves, the high efficiency electron heating of tokamak plasma was verified, and the heating characteristics were clarified. And others. (K.I.) 179 refs

  17. I.R. and F.I.R. laser polarimetry as a diagnostic tool in high-β and Tokamak plasmas

    International Nuclear Information System (INIS)

    Pereira, D.; Machida, M.; Scalabrin, A.

    1986-01-01

    The change of the polarization state of an electromagnetic wave (E.M.W.) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the E.M.W. allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-Pinches plasmas (lambda approx. 10μm). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-Pinch/UNICAMP, using lasers developed at UNICAMP. (Author) [pt

  18. Residual gas analysis of a cryostat vacuum chamber during the cool down of SST - 1 superconducting magnet field coil

    International Nuclear Information System (INIS)

    Semwal, P.; Joshi, K.S.; Thankey, P.L.; Pathan, F.S.; Raval, D.C.; Patel, R.J.; Pathak, H.A.

    2005-01-01

    One of the most important feature of Steady state Superconducting Tokamak -1 (SST-l) is the Nb-Ti superconducting magnet field coils. The coils will be kept in a high vacuum chamber (Cryostat) and liquid Helium will be flown through it to cool it down to its critical temperature of 4.5K. The coil along with its hydraulics has four types of joints (1) Stainless Steel (S.S.) to Copper (Cu) weld joints (2) S. S. to S. S. weld joints (3) Cu to Cu brazed joints and (4) G-10 to S. S. joints with Sti-cast as the binding material. The joints were leak tested with a Helium mass spectrometer leak detector in vacuum as well as in sniffer mode. However during the cool-down of the coil, these joints may develop leaks. This would deteriorate the vacuum inside the cryostat and coil cool-down would subsequently become more difficult. To study the effect of cooling on the vacuum condition of the Cryostat, a dummy Cryostat chamber was fabricated and a toroidal Field (TF) magnet was kept inside this chamber and cooled down to 4.5 K.A residual gas analyzer (RGA) was connected to the Cryostat chamber to study the behaviour of major gases inside this chamber with temperature. An analysis of the RGA data acquired during the coo-down has been presented in this chamber. (author)

  19. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    1982-03-01

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  20. Plasma diagnostics for tokamaks and stellarators. Proceedings of the IV Course and Workshop on Magnetic Confinement Fusion. UIMP Santander (Spain), June 1992

    International Nuclear Information System (INIS)

    Stott, P. E.; Sanchez, J.

    1994-01-01

    A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: Magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Scattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma. (Author) 451 refs

  1. Integration of -70kV, 22A high voltage power supply with solid state crowbar and the LHCD system of SST-1

    International Nuclear Information System (INIS)

    Rajan Babu, N.; Virani, C.G.; Dalakoti, S.; Sharma, P.K.; Ambulkar, K.K.; Parmar, P.R.; Thakur, A.L.; Dhorajiya, Pragnesh

    2015-01-01

    LHCD system is a important system for the steady state operation of the SST-1 machine. Four numbers of klystrons of 3.7 GHz are used as a microwave source to produce 2 MW of microwave power. This power is launched into the machine to achieve the steady state operation of the SST-1 Machine. A -70kV, 22A high voltage power supply and a solid-state crowbar are procured and tested and validated for its performance separately. Both of the system are integrated and tested for its integrated performance for the safe and reliable test of the klystron tube. A 10J wire test is conducted for the optimum value of the series resistor. This test will validate the integrated performance of power supply, Crowbar and the interlocking circuit. This paper details the optimization of the ballast resistor from 150 ohms to 40 ohms and its successful integration with the klystron tube for its 500kW CW operation. Some operational experience is also shared

  2. Steady state operation of tokamaks. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2000-10-01

    The first IAEA Technical Committee Meeting (TCM) on Steady State Operation of Tokamaks was organized to discuss the operations of present long-pulse tokamaks (TRIAM-1M, TORE SUPRA, MT-7, HT-7M, HL-1M) and the plans for future steady-state tokamaks such as SST-1, CIEL, and HT-7U. This meeting, held from 13-15 October 1998, was hosted by the Academia Sinica Institute of Plasma Physics (ASIPP), Hefei, China. Participants from China, France, India, Japan, the Russian Federation, and the IAEA participated in the meeting. There were 18 individual presentations plus general discussions on many topics, including superconducting magnet systems, cryogenics, plasma position control, non-inductive current drive, auxiliary heating, plasma-wall interactions, high heat flux components, particle control, and data acquisition

  3. Diagnostics

    DEFF Research Database (Denmark)

    Donné, A.J.H.; Costley, A.E.; Barnsley, R.

    2007-01-01

    of the measurements—time and spatial resolutions, etc—will in some cases be more stringent. Many of the measurements will be used in the real time control of the plasma driving a requirement for very high reliability in the systems (diagnostics) that provide the measurements. The implementation of diagnostic systems...... on ITER is a substantial challenge. Because of the harsh environment (high levels of neutron and gamma fluxes, neutron heating, particle bombardment) diagnostic system selection and design has to cope with a range of phenomena not previously encountered in diagnostic design. Extensive design and R......&D is needed to prepare the systems. In some cases the environmental difficulties are so severe that new diagnostic techniques are required. The starting point in the development of diagnostics for ITER is to define the measurement requirements and develop their justification. It is necessary to include all...

  4. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  5. High Beta Tokamak research

    International Nuclear Information System (INIS)

    Navratil, G.A.; Mauel, M.E.; Ivers, T.H.; Sankar, M.K.V.; Eisner, E.; Gates, D.; Garofalo, A.; Kombargi, R.; Maurer, D.; Nadle, D.; Xiao, Q.

    1993-01-01

    During the past 6 months, experiments have been conducted with the HBT-EP tokamak in order to (1) test and evaluate diagnostic systems, (2) establish basic machine operation, (3) document MHD behavior as a function of global discharge parameters, (4) investigate conditions leading to passive stabilization of MHD instabilities, and (5) quantify the external saddle coil current required for DC mode locking. In addition, the development and installation of new hardware systems has occurred. A prototype saddle coil was installed and tested. A five-position (n,m) = (1,2) external helical saddle coil was attached for mode-locking experiments. And, fabrication of the 32-channel UV tomography and the multipass Thomson scattering diagnostics have begun in preparation for installation later this year

  6. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11-12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of

  7. Development of thin foil Faraday collector as a lost alpha particle diagnostic for high yield D-T tokamak fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Van Belle, P; Jarvis, O N; Sadler, G J [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Cecil, F E [Colorado School of Mines, Golden, CO (United States)

    1994-07-01

    Alpha particle confinement is necessary for ignition of a D-T tokamak fusion plasma and for first wall protection. Due to high radiation backgrounds and temperatures, scintillators and semiconductor detectors may not be used to study alpha particles which are lost to the first wall during the D-T programs on JET and ITER. An alternative method of charged particle spectrometry capable of operation in these harsh environments, is proposed: it consists of thin foils of electrically isolated conductors with the flux of alpha particles determined by the positive current flowing from the foils. 2 refs., 3 figs.

  8. Evaluation of the optical design of laser Thomson scattering diagnostics for high-temperature EAST tokamak and low-temperature MAP-II divertor simulator

    International Nuclear Information System (INIS)

    Kado, Shinichiro; Scotti, Filippo; Xi Xiaoqi; Zhao Junyu

    2009-01-01

    The optical design of the laser Thomson scattering (LTS) system for EAST tokamak is now on-going. Based on the Visible YAG laser TVTS system developed in the MAP-II (material and plasma) steady-state linear divertor/edge plasma simulator at the University of Tokyo, the required specification and the applicability of the VIS-YAG-TVTS system was evaluated in terms of the photon number to be collected by the fiber light-guide to a spectrometer and the reciprocal linear dispersion of the spectrometer. Then, the possible design of the optical system was proposed. (author)

  9. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-05-01

    The technical reports in this document were presented at the IAEA Technical Committee Meeting ''Research on Small Tokamaks'', September 1990, in three sessions, viz., (1) Plasma Modes, Control, and Internal Phenomena, (2) Edge Phenomena, and (3) Advanced Configurations and New Facilities. In Section (1) experiments at controlling low mode number modes, feedback control using external coils, lower-hybrid current drive for the stabilization of sawtooth activity and continuous (1,1) mode, and unmodulated and fast modulated ECRH mode stabilization experiments were reported, as well as the relation to disruptions and transport of low m,n modes and magnetic island growth; static magnetic perturbations by helical windings causing mode locking and sawtooth suppression; island widths and frequency of the m=2 tearing mode; ultra-fast cooling due to pellet injection; and, finally, some papers on advanced diagnostics, i.e., lithium-beam activated charge-exchange spectroscopy, and detection through laser scattering of discrete Alfven waves. In Section (2), experimental edge physics results from a number of machines were presented (positive biasing on HYBTOK II enhancing the radial electric field and improving confinement; lower hybrid current drive on CASTOR improving global particle confinement, good current drive efficiency in HT-6B showing stabilization of sawteeth and Mirnov oscillations), as well as diagnostic developments (multi-chord time resolved soft and ultra-soft X-ray plasma radiation detection on MT-1; measurements on electron capture cross sections in multi-charged ion-atom collisions; development of a diagnostic neutral beam on Phaedrus-T). Theoretical papers discussed the influence of sheared flow and/or active feedback on edge microstability, large edge electric fields, and two-fluid modelling of non-ambipolar scrape-off layers. Section (3) contained (i) a proposal to construct a spherical tokamak ''Proto-Eta'', (ii) an analysis of ultra-low-q and runaway

  10. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

    The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)

  11. Prospects for Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.

    1995-01-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

  12. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  13. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  14. Comprehensive numerical modelling of tokamaks

    International Nuclear Information System (INIS)

    Cohen, R.H.; Cohen, B.I.; Dubois, P.F.

    1991-01-01

    We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell

  15. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  16. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    1978-08-01

    Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system

  17. Anomalous periodic disruptions in tokamak plasma

    International Nuclear Information System (INIS)

    Montvai, A.; Tegze, M.; Valyi, I.

    1982-09-01

    Anomalously strong, periodic instabilities were observed in the MT-1 tokamak. Characteristics of these instabilities were partly similar to those of internal disruptions, but there were features making them different from the normal relaxational oscillations. Basic characteristics of the phenomenon were studied with the aid of generally used diagnostics. (author)

  18. Tokamak devices: towards controlled fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

  19. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Dickens, J.K.; Hill, N.W.; Hou, F.S.; McConnell, J.W.; Spencer, R.R.; Tsang, F.Y.

    1985-08-01

    A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report

  20. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  1. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  2. Advanced Tokamak Stability Theory

    Science.gov (United States)

    Zheng, Linjin

    2015-03-01

    The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

  3. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

    The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

  4. Tokamak concept innovations

    International Nuclear Information System (INIS)

    1986-04-01

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  5. Time - resolved thermography at Tokamak T-10

    International Nuclear Information System (INIS)

    Grunow, C.; Guenther, K.; Lingertat, J.; Chicherov, V.M.; Evstigneev, S.A.; Zvonkov, S.N.

    1987-01-01

    Thermographic experiments were performed at T-10 tokamak to investigate the thermic coupling of plasma and the limiter. The limiter is an internal equipment of the vacuum vessel of tokamak-type fusion devices and the interaction of plasma with limiter results a high thermal load of limiter for short time. In according to improve the limiter design the temperature distribution on the limiter surface was measured by a time-resolved thermographic method. Typical isotherms and temperature increment curves are presented. This measurement can be used as a systematic plasma diagnostic method because the limiter is installed in the tokamak whereas special additional probes often disturb the plasma discharge. (D.Gy.) 3 refs.; 7 figs

  6. Use of soft x-ray diagnostic on the COMPASS tokamak for investigations of sawteeth crash neighborhood and of plasma position using fast inversion methods

    Czech Academy of Sciences Publication Activity Database

    Imríšek, Martin; Weinzettl, Vladimír; Mlynář, Jan; Odstrčil, T.; Odstrčil, M.; Ficker, O.; Pinzon, J.R.; Ehrlacher, C.; Pánek, Radomír; Hron, Martin

    2014-01-01

    Roč. 85, č. 11 (2014), 11E433-11E433 ISSN 0034-6748. [Topical Conference on High-Temperature Plasma Diagnostics/20./. Atlanta, 01.06.2014-05.06.2014] R&D Projects: GA ČR GAP205/10/2055; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Inversion methods * Plasma positions * Sawteeth * Soft X-ray Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.614, year: 2014 http://dx.doi.org/10.1063/1.4894528

  7. Tokamak control simulator

    International Nuclear Information System (INIS)

    Edelbaum, T.N.; Serben, S.; Var, R.E.

    1976-01-01

    A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks

  8. Parametric study of ohmic discharges in the TCA tokamak

    International Nuclear Information System (INIS)

    De Chambrier, A.; Collins, G.A.; Heym, A.; Hofmann, F.; Hollenstein, Ch.; Joye, B.; Keller, R.; Lietti, A.; Lister, J.B.; Moret, J.-M.; Nowak, S.; O'Rourke, J.; Pochelon, A.; Simm, W.

    1983-01-01

    The study of the energy confinement in a tokamak is an important aspect in the characterisation of its performance. The TCA tokamak has been in operation now for more than two years and the state of the machine and of its diagnostics have permitted such work to be performed. The authors describe the proper method for this type of approach and then present the results concerning the energy confinement of the electrons and ions. (Auth./G.T.H.)

  9. Experimental investigations of driven Alfven wave resonances in a tokamak plasma using carbon dioxide laser interferometry

    International Nuclear Information System (INIS)

    Evans, T.E.

    1984-09-01

    The first direct observation of the internal structure of driven global Alfven eigenmodes in a tokamak plasma is presented. A carbon dioxide laser scattering/interferometer has been designed, built, and installed on the PRETEXT tokamak. By using this diagnostic system in the interferometer configuration, we have for the first time, thoroughly investigated the resonance conditions required for, and the spatial wave field structure of, driven plasma eigenmodes at frequencies below the ion cyclotron frequency in a confined, high temperature, tokamak plasma

  10. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

    Wesson, John; Campbell, D.J.; Connor, J.W.

    1997-01-01

    It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

  11. KTM Tokamak operation scenarios software infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, V.; Baystrukov, K.; Golobkov, YU.; Ovchinnikov, A.; Meaentsev, A.; Merkulov, S.; Lee, A. [National Research Tomsk Polytechnic University, Tomsk (Russian Federation); Tazhibayeva, I.; Shapovalov, G. [National Nuclear Center (NNC), Kurchatov (Kazakhstan)

    2014-10-15

    One of the largest problems for tokamak devices such as Kazakhstan Tokamak for Material Testing (KTM) is the operation scenarios' development and execution. Operation scenarios may be varied often, so a convenient hardware and software solution is required for scenario management and execution. Dozens of diagnostic and control subsystems with numerous configuration settings may be used in an experiment, so it is required to automate the subsystem configuration process to coordinate changes of the related settings and to prevent errors. Most of the diagnostic and control subsystems software at KTM was unified using an extra software layer, describing the hardware abstraction interface. The experiment sequence was described using a command language. The whole infrastructure was brought together by a universal communication protocol supporting various media, including Ethernet and serial links. The operation sequence execution infrastructure was used at KTM to carry out plasma experiments.

  12. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  13. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  14. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1977-01-01

    The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

  15. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.

  16. Development of Atomic Beam Probe for tokamaks

    Czech Academy of Sciences Publication Activity Database

    Berta, M.; Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S.; Havlíček, Josef; Háček, Pavel

    2013-01-01

    Roč. 88, č. 11 (2013), s. 2875-2880 ISSN 0920-3796 R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : ABP * Plasma diagnostics * COMPASS tokamak * Current density * Plasma density profile measurement Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613005048#

  17. Distributed digital real-time control system for the TCV tokamak and its applications

    NARCIS (Netherlands)

    Anand, H.; Galperti, C.; Coda, S.; Duval, B.P.; Felici, F.; Blanken, T.; Maljaars, E.; Moret, J.M.; Sauter, O.; Goodman, T.P.; Kim, D.

    2017-01-01

    A key feature of the new digital plasma control system installed on the TCV (Tokamak à Configuration Variable) tokamak is its possibility to rapidly design, test and deploy real-time algorithms. It accommodates hundreds of diagnostic inputs and actuator outputs, and offers the possibility to design

  18. The role of high speed photography in plasma instability research on the AEC tokamak

    International Nuclear Information System (INIS)

    Fletcher, J.D.; Coster, D.P.; De Villiers, J.A.M.; Kotze, P.B.; Nothnagel, G.; O'Mahony, J.R.; Roberts, D.E.; Sherwell, D.

    1986-01-01

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions in fusion research devices like tokamaks. Such a system has been installed on the AEC tokamak. This paper reports some preliminary results obtained during typical plasma discharges

  19. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  20. Internal Magnetic Configuration Measured by ECE Imaging on EAST Tokamak

    International Nuclear Information System (INIS)

    Xu Ming; Wen Yizhi; Xie Jinlin; Yu Changxuan; Gao Bingxi; Xu Xiaoyuan; Liu Wandong; Hu Liqun; Sun Youwen; Qian Jinping; Wan Baonian

    2013-01-01

    ECE imaging (electron cyclotron emission imaging) is an important diagnostic which can give 2D imaging of temperature fluctuation in the core of tokamak. A method based on ECE imaging is introduced which can give the information of the position of magnetic axis and the structure of internal magnetic surface for EAST tokamak. The EFIT equilibrium reconstruction is not reliable due to the absence of important core diagnostic at the initial phase for EAST, so the information given by ECE imaging could help to improve the accuracy of EFIT equilibrium reconstruction. (magnetically confined plasma)

  1. The scientific program of the Tokamak de Varennes

    International Nuclear Information System (INIS)

    Daughney, C.C.

    1989-01-01

    The Tokamak de Varennes (TdeV) is the principal research tool of the Centre canadien de fusion magnetique (CCFM). This article places the Tokamak de Varennes within the framework of the Canadian National Fusion Program (NFP) and describes the scientific program of the TdeV as it was presented at the April 1989 meeting of the CCFM Advisory Committee. The CCFM scientific plant contains three main elements: tokamak development, research on transport and equilibrium in plasmas, and research on the plasma-wall problem. Phase I of the experimental program, commissioning the tokamak and the diagnostic systems, has been completed. Phase II of the experimental program will begin in December 1989 with the plasma boundary defined by a magnetic divertor and the power supplies and vacuum system capable of creating a sequence of one-second plasma pulses. (3 figs., 3 refs.) (L.L.)

  2. Development of two series ingnitron based crowbar protection system for 42 GHz and 82.6 GHz gyroton in SST-1

    International Nuclear Information System (INIS)

    Dhorajiya, Pragnesh; Dalakoti, Shefali; Patel, Harshida; Ingle, Krunal; Patel, Jatin; Sathyanarayana, K.; Rajanbabu; Shukla, B.K.

    2013-01-01

    Gyrotrons are used to generate the high power at microwave frequency that is used to heat the plasma inside a Tokamak. A conventional high voltage power supply is used for the testing of 82.6 GHz, 200 kW/CW and 42 GHz, 500 kW/500ms gyrotrons at our institute. Its maximum operating cathode parameters are -55 kV DC, 20 A. Like any other High RF power tubes gyrotrons need to be protected against arc faults within the tube. If the energy dumped in such arc fault is more than the critical crater energy of the tube, irreparable damage can occur inside the RF tube or microwave tube and rendering it useless. The specified maximum fault energy for the 42 GHz and 82.6 GHz gyrotrons is 10 joules. When conventional HVDC power supplies feed high power RF tubes or microwave tubes, a reliable crowbar protection is required which is tested separately to limit the energy to the tube in case of any type of fault to assure the tube safety. Two series ignitron (NL-37248) based crowbar system developed in-house is used to limit the arc fault energy under the acceptance level by diverting the fault current from the load or Gyrotron. Fault current diversion and interruption are initiated by the sensing element and protection system. The required protection cards are designed and developed in-house and required performance is achieved. With this crowbar system the high voltage switch-off to the gyrotron is achieved within 5 μsec after occurrence of critical faults. The crowbar is tested for voltage hold-off up to 80 kV DC. This paper presents the critical requirement of the time delay for the fault sensing and crowbar trigger generation and necessary protections that are incorporated with the ignitron switch crowbar like over voltage, pulsed over current and continuous over current. The crowbar system developed in-house, tested at rated value. The results obtained during the stand-alone tests and commissioning tests are also mentioned. Using this crowbar system the high voltage power

  3. X-ray diagnostics of tokamak plasmas

    International Nuclear Information System (INIS)

    Bitter, M.; Hsuan, H.; Hill, K.W.; Zarnstorff, M.

    1993-04-01

    This paper discusses the influence that salinity has on the bioavailability of the two largest classes of contaminants, trace metals and organic compounds will be discussed. Although data on contaminant toxicity will be used to draw inferences about chemical availability, this discussion will focus on the properties that contaminants are likely to exhibit in waters of varying salinities. In addition, information on physiological changes that are affected by salinity will be used to illustrate how biological effects can alter the apparent availability of contaminants

  4. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

  5. An electrostatic detector for dust measurement on HT-7 tokamak

    International Nuclear Information System (INIS)

    Ling, B.L.; Zhang, X.D.; Ti, A.; Gao, X.

    2007-01-01

    An electrostatic dust detector has been successfully developed to measure dust event in situ and in real time on the HT-7 tokamak. For measuring dust near the edge plasmas and preventing interference of electrons and ions, the shielding plates were designed and installed around the dust detector. The electric signal of dust has been successfully measured during LHCD discharges on HT-7 tokamak. The measured dust signal was in good agreement with bursts appeared on multi-channel H α radiation and on multi-channel ECE diagnostics. Diagnostics of the spectrum and the measurement of impurity emission during dust bursts were studied in detail. It is interesting that there is a delay between dust bursts and CIII line emission. It is observed that the delay time between dust signal and measured CIII line emission is about 0.3 ms in the HT-7 tokamak

  6. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-03-01

    This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

  7. Design and construction of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, G.S.

    2001-01-01

    The extensive design effort has been focused on two major aspects of the KSTAR project mission, steady-state operation capability and 'advanced tokamak' physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long-pulse current-drive and heating systems. The 'advanced tokamak' aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizers, internal control coils , and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR experimental facility with cryogenic system and de-ionized water-cooling and main power systems has been designed, and the construction work has been on-going for completion in year 2004. (author)

  8. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  9. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1993-04-01

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

  10. Magnetic ''islandography'' in tokamaks

    International Nuclear Information System (INIS)

    Callen, J.D.; Waddell, B.V.; Hicks, H.R.

    1978-09-01

    Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

  11. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

    This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

  12. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

  13. Internal disruption in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1990-01-01

    A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described

  14. Overview of Tokamak Results

    International Nuclear Information System (INIS)

    Unterberg, Bernhard; Samm, Ulrich

    2004-01-01

    An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas

  15. Internal disruptions in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1990-01-01

    Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described

  16. High beta tokamaks

    International Nuclear Information System (INIS)

    Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.

    1978-01-01

    MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby

  17. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1989-01-01

    A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

  18. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Pare, V.K.

    1983-01-01

    Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

  19. Compact tokamak reactors

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1997-01-01

    The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

  20. Tokamaks (Second Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Stott, Peter [JET, UK (United Kingdom)

    1998-10-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  1. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  2. The design of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, G.S.; Kim, J.; Hwang, S.M.

    1999-01-01

    The Korea superconducting tokamak advanced research (KSTAR) project is the major effort of the Korean national fusion program (KNFP) to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. Major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 Tesla, and plasma current 2 MA with a strongly shaped plasma cross-section and double-null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beam, ion cyclotron waves, lower hybrid waves, and electron-cyclotron waves for flexible profile control. A comprehensive set of diagnostics is planned for plasma control and performance evaluation and physics understanding. The project has completed its conceptual design phase and moved to the engineering design phase. The target date of the first plasma is set for year 2002. (orig.)

  3. Recent Activities on the Experimental Research Programme Using Small Tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M. P.; Bosco, E. del; Malaquias, A.; Mank, G.; Oost, G. van

    2006-01-01

    A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project (CRP) is discussed in this paper. Besides the presentation of the recent activities on the experimental research programme using small tokamaks and scientific results achieved at the participating laboratories, information is provided about the organisation of the co-ordinated research project. Future plans of the co-ordinated activities within the CRP are discussed

  4. Data processing system for spectroscopy at Novillo Tokamak

    International Nuclear Information System (INIS)

    Ortega C, G.; Gaytan G, E.

    1998-01-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  5. Research using small tokamaks. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-09-01

    The technical reports in these proceedings were presented at the IAEA Technical Committee Meeting on research Using Small Tokamaks, held in Ahmedabad, India, 6-7 December 1995. The purpose of this annual meeting is to provide a forum for the exchange of information on various small and medium sized plasma experiments, not only for tokamaks. The potential benefits of these research programmes are to: test theories, such as effects of the plasma rotation; check empirical scalings, such as density limits; develop fusion technology hardware; develop plasma diagnostics; such as tomography; and to train scientists, engineers, technicians, and students, particularly in developing IAEA Member States

  6. Present status of TCA/BR Tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Galvao, R.M.O.; Tuszel, A.G.

    1997-01-01

    The TCA tokamak is being partially reconstructed and reassembled in the Plasma Laboratory of The University of Sao Paulo, and afterwards it will be named TCA/BR. The first discharges are expected by June/July of next year. The main scientific objectives envisaged for the machine are: Alfven wave heating and current drive, confinement improvement, disruptions and turbulence. In this paper we also describe: (i) the present status of the project; (ii) the diagnostic system; (iii) the control and data acquisition system; (iv) the RF system for the excitation of Alfven waves, that are being developed, and also the results of predictive transport simulations of its performance. (author)

  7. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

    1980-06-01

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  8. Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  9. Coherent structures in tokamak plasmas workshop: Proceedings

    International Nuclear Information System (INIS)

    Koniges, A.E.; Craddock, G.G.

    1992-08-01

    Coherent structures have the potential to impact a variety of theoretical and experimental aspects of tokamak plasma confinement. This includes the basic processes controlling plasma transport, propagation and efficiency of external mechanisms such as wave heating and the accuracy of plasma diagnostics. While the role of coherent structures in fluid dynamics is better understood, this is a new topic for consideration by plasma physicists. This informal workshop arose out of the need to identify the magnitude of structures in tokamaks and in doing so, to bring together for the first time the surprisingly large number of plasma researchers currently involved in work relating to coherent structures. The primary purpose of the workshop, in addition to the dissemination of information, was to develop formal and informal collaborations, set the stage for future formation of a coherent structures working group or focus area under the heading of the Tokamak Transport Task Force, and to evaluate the need for future workshops on coherent structures. The workshop was concentrated in four basic areas with a keynote talk in each area as well as 10 additional presentations. The issues of discussion in each of these areas was as follows: Theory - Develop a definition of structures and coherent as it applies to plasmas. Experiment - Review current experiments looking for structures in tokamaks, discuss experimental procedures for finding structures, discuss new experiments and techniques. Fluids - Determine how best to utilize the resource of information available from the fluids community both on the theoretical and experimental issues pertaining to coherent structures in plasmas. Computation - Discuss computational aspects of studying coherent structures in plasmas as they relate to both experimental detection and theoretical modeling

  10. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  11. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

    1981-08-01

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  12. Theory of tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.

    1989-01-01

    The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.

  13. Axisymmetric tokamak scapeoff transport

    International Nuclear Information System (INIS)

    Singer, C.E.; Langer, W.D.

    1982-08-01

    We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed

  14. Key features and progress of the KSTAR tokamak engineering

    International Nuclear Information System (INIS)

    Bak, J.S.; Choi, C.H.; Oh, Y.K.

    2003-01-01

    Substantial progress of the KSTAR tokamak engineering has been made on major tokamak structures, superconducting magnets, in-vessel components, diagnostic system, heating system, and power supplies. The engineering design has been elaborated to the extent necessary to allow a realistic assessment of its feasibility, performance, and cost. The prototype fabrication has been carried out to establish the reliable fabrication technologies and to confirm the validation of analyses employed for the KSTAR design. The completion of experimental building with beneficial occupancy for machine assembly was accomplished in Sep. 2002. The construction of special utility such as cryo-plant, de-ionized water-cooling system, and main power station will begin upon completion of building construction. The commissioning, construction, fabrication, and assembly of the whole facility will be going on by the end of 2005. This paper describes the main design features and engineering progress of the KSTAR tokamak, and elaborates the work currently underway. (author)

  15. Density limits in Tokamaks

    International Nuclear Information System (INIS)

    Tendler, M.

    1984-06-01

    The energy loss from a tokamak plasma due to neutral hydrogen radiation and recycling is of great importance for the energy balance at the periphery. It is shown that the requirement for thermal equilibrium implies a constraint on the maximum attainable edge density. The relation to other density limits is discussed. The average plasma density is shown to be a strong function of the refuelling deposition profile. (author)

  16. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.

    1984-05-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

  17. Modular tokamak magnetic system

    International Nuclear Information System (INIS)

    Yang, T.F.

    1988-01-01

    This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil

  18. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.; California Univ., Los Angeles

    1984-01-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)

  19. TPX tokamak construction management

    International Nuclear Information System (INIS)

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-01-01

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly

  20. Development of atomic beam probe for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Berta, M., E-mail: bertam@sze.hu [Széchenyi István University, EURATOM Association, Győr (Hungary); Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, EURATOM Association, Budapest (Hungary); Havlícek, J.; Háček, P. [Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics (Czech Republic)

    2013-11-15

    Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies.

  1. Development of atomic beam probe for tokamaks

    International Nuclear Information System (INIS)

    Berta, M.; Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S.; Havlícek, J.; Háček, P.

    2013-01-01

    Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies

  2. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1994-01-01

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  3. The density limit in Tokamaks

    International Nuclear Information System (INIS)

    Alladio, F.

    1985-01-01

    A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs

  4. Optical visualization of electric and magnetic field perturbations in tokamak discharges by hydrogen pellet injection

    International Nuclear Information System (INIS)

    Drawin, H.W.; Dubois, M.A.

    1992-04-01

    Two-dimensional intensity distribution mappings of photographs of pellet ablation cloud trajectories in the TFR and TS tokamaks reveal irregular shapes of the luminous striations. The observed features are not well understood, but can be described and interpreted as the first direct visual observation of pre-existing electric and/or magnetic field perturbations in the hot core of tokamak plasmas. It is suggested to use pellet injection as a diagnostic tool for the study of plasma structures and transport phenomena

  5. [Fusion research/tokamak]. Final report, 1 May 1988 - 30 April 1994

    International Nuclear Information System (INIS)

    1994-01-01

    The objectives of the Fusion Research Center Program are: (1) to advance /the transport studies of tokamaks, including the development and maintenance of the Magnetic Fusion Energy Database, and (2) to provide theoretical interpretation, modeling and equilibrium and stability studies for the text-upgrade tokamak. Work is described on five basic categories: (1) magnetic fusion energy database; (2) computational support and numerical modeling; (3) support for TEXT-upgrade and diagnostics; (4) transport studies; and (5) Alfven waves

  6. Mode particle resonances during near-tangential neutral beam injection in large tokamaks

    International Nuclear Information System (INIS)

    Kaita, R.; White, R.B.; Morris, A.W.; Fredrickson, E.D.; McGuire, K.M.; Medley, S.S.; Scott, S.D.

    1988-01-01

    Coherent magnetohydrodynamic modes have been observed during neutral beam injection in TFTR and JET. Periodic bursts of oscillations were detected with several plasma diagnostics, and Fokker-Planck calculations show that the populations of trapped particles in both tokamaks are sufficient to account for fishbone destabilization. Estimates of mode parameters are in reasonable agreement with the experiments, and they indicate that the fishbone mode may continue to affect the performance of intensely heated tokamaks. 13 refs., 1 fig., 1 tab

  7. Deposit of thin films for Tokamaks conditioning

    International Nuclear Information System (INIS)

    Valencia A, R.

    2006-01-01

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature ( -6 to 4.5 x 10 -6 Ω-m, thus taking the Z ef value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow

  8. G1SST, 1km blended SST

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — A daily, global Sea Surface Temperature (SST) data set is produced at 1-km (also known as ultra-high resolution) by the JPL ROMS (Regional Ocean Modeling System)...

  9. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    Hendel, H.W.; Jassby, D.L.

    1989-11-01

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  10. Tokamak instrumentation and controls

    International Nuclear Information System (INIS)

    Becraft, W.R.; Bettis, E.S.; Houlberg, W.A.; Onega, R.J.; Stone, R.S.

    1979-02-01

    The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine

  11. Demonstration tokamak power plant

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

  12. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    Minardi, E.

    1989-01-01

    The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

  13. Topology of tokamak orbits

    International Nuclear Information System (INIS)

    Rome, J.A.; Peng, Y.K.M.

    1978-09-01

    Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well

  14. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  15. Modelling dust transport in tokamaks

    International Nuclear Information System (INIS)

    Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

    2008-01-01

    The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

  16. Design and operation of Alvand II C tokamak

    International Nuclear Information System (INIS)

    Avakian, M.; Azodi, H.; Naraghi, M.; Tabatabaie, H.; Rezvani, B.; Shahnazi, Sh.; Behrozinia, J.; Solaimani, M.

    1984-01-01

    ALVAND IIC is a research tokamak having a circular cross section with major radius of 45.5 cm and minor radius of 12.6 cm, which has been designed and constructed in the plasma physics division of NRC of AEOI. It is used for developing various diagnostics for low beta plasmas and study of the physical characteristics of these plasmas. This paper discusses the design of ALVAND IIC. (Author)

  17. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    International Nuclear Information System (INIS)

    Castracane, J.

    2001-01-01

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies

  18. New design of microwave interferometer for tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Varavin, Mykyta; Zajac, Jaromír; Žáček, František; Nanobashvili, S.I.; Ermak, G.; Varavin, A.; Vasilev, A.; Stumbra, M.; Vetoshko, A.; Fateev, A.; Shevchenko, V.

    2014-01-01

    Roč. 73, č. 10 (2014), s. 935-942 ISSN 0040-2508 Institutional support: RVO:61389021 Keywords : Microwave interferometry * Plasma diagnostics * Tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www.dl.begellhouse.com/journals/0632a9d54950b268,0f0577fd35766f78,6e6cc5d616079cba.html

  19. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Castracane, J.

    2001-01-04

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.

  20. Tokamak building-design considerations for a large tokamak device

    International Nuclear Information System (INIS)

    Barrett, R.J.; Thomson, S.L.

    1981-01-01

    Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

  1. Validation of equilibrium tools on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Urban, J., E-mail: urban@ipp.cas.cz [Institute of Plasma Physics ASCR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Appel, L.C. [CCFE, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); Artaud, J.F. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Faugeras, B. [Laboratoire J.A. Dieudonné, UMR 7351, Université de Nice Sophia-Antipolis, Parc Valrose, 06108 Nice Cedex 02 (France); Havlicek, J. [Institute of Plasma Physics ASCR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Komm, M. [Institute of Plasma Physics ASCR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Lupelli, I. [CCFE, Culham Science Centre, Abingdon, Oxfordshire (United Kingdom); Peterka, M. [Institute of Plasma Physics ASCR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic)

    2015-10-15

    Highlights: • Three equilibrium codes—EFIT++, FREEBIE and VacTH—have been successfully set up and validated on COMPASS. • FREEBIE can predictively calculate the equilibrium and corresponding poloidal field coil currents. • EFIT++ can reconstruct equilibria generated by FREEBIE from synthetic, optionally noisy diagnostic data. • VacTH is a promising tool for real time plasma shape reconstruction. • Optimized parameters are estimated for EFIT++ and VacTH by a statistical analysis. - Abstract: Various MHD (magnetohydrodynamic) equilibrium tools, some of which being recently developed or considerably updated, are used on the COMPASS tokamak at IPP Prague. MHD equilibrium is a fundamental property of the tokamak plasma, whose knowledge is required for many diagnostics and modelling tools. Proper benchmarking and validation of equilibrium tools is thus key for interpreting and planning tokamak experiments. We present here benchmarks and comparisons to experimental data of the EFIT++ reconstruction code (Appel et al., 2006), the free-boundary equilibrium code FREEBIE (Artaud and Kim, 2012), and a rapid plasma boundary reconstruction code VacTH (Faugeras et al., 2014). We demonstrate that FREEBIE can calculate the equilibrium and corresponding poloidal field (PF) coils currents consistently with EFIT++ reconstructions from experimental data. Both EFIT++ and VacTH can reconstruct equilibria generated by FREEBIE from synthetic, optionally noisy diagnostic data. Hence, VacTH is suitable for real-time control. Optimum reconstruction parameters are estimated.

  2. Natural current profiles in tokamaks

    International Nuclear Information System (INIS)

    Biskamp, D.

    1986-01-01

    It is proposed that a certain class of equilibrium, which follow from an elementary variational principle, are the natural current profiles in tokamaks, to which actual discharge profiles tend to relax. (orig.)

  3. Alcator C-Mod Tokamak

    Data.gov (United States)

    Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...

  4. JUST: Joint Upgraded Spherical Tokamak

    International Nuclear Information System (INIS)

    Azizov, E.A.; Dvorkin, N.Ya.; Filatov, O.G.

    1997-01-01

    The main goals, ideas and the programme of JUST, spherical tokamak (ST) for the plasma burn investigation, are presented. The place and prospects of JUST in thermonuclear investigations are discussed. (author)

  5. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.; Saramad, S.

    2012-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

  6. New directions in tokamak reactors

    International Nuclear Information System (INIS)

    Baker, C.C.

    1985-01-01

    New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs

  7. The Tokamak IST-TOK

    International Nuclear Information System (INIS)

    Varandas, C.A.F.; Cabral, J.A.C.; Manso, M.E.

    1991-01-01

    A small tokamak is under construction at the Portuguese Technical Superior Institute. The main objective is to create a home based laboratory in which an independent scientific program might be developed. (L.C.J.A.). 14 refs, 6 figs

  8. Overview of the Tokamak de Varennes program

    International Nuclear Information System (INIS)

    Pacher, H.D.

    1986-01-01

    The Tokamak de Varennes will be the major Canadian experiment in the magnetic fusion domain. It has a toroidal field of 1.5 tesla, major radius of 0.85 m, a minor radius of 0.25 m, and will study long pulses, up to 30 seconds duration. Initially, a series of successive plasma pulses, each of the order of seconds, will yield a duty factor of over 50 percent. During this phase, the major emphasis will be on the study of impurity generation, transport, and control, plasma-wall interactions and material properties. The program will include studies of fast current rampdown and the resultant current profile modifications. The development of advanced diagnostics will also be undertaken. To attain a higher duty factor with continuous plasma operation, noninductive current drive by radio=frequency will be added as an early upgrade. This will introduce current drive investigations such as transformer recharge and profile relaxation, and enhance the wall and materials study program. In this context, the Tokamak de Varennes will concentrate on the study of impurity exhaust and retention as well as net erosion of the limiter and neutralization plate materials

  9. MDSplus integration at TCABR tokamak: Current status

    International Nuclear Information System (INIS)

    Sá, W.P. de; Ronchi, G.

    2016-01-01

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  10. MDSplus integration at TCABR tokamak: Current status

    Energy Technology Data Exchange (ETDEWEB)

    Sá, W.P. de, E-mail: pires@if.usp.br; Ronchi, G., E-mail: gronchi@if.usp.br

    2016-11-15

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  11. Neutron measurement techniques for tokamak plasmas

    International Nuclear Information System (INIS)

    Jarvis, O.N.

    1994-01-01

    The present article reviews the neutron measurement techniques that are currently being applied to the study of tokamak plasmas. The range of neutron energies of primary interest is limited to narrow bands around 2.5 and 14 MeV, and the variety of measurements that can be made for plasma diagnostic purposes is also restricted. To characterize the plasma as a neutron source, it is necessary only to measure the total neutron emission, the relative neutron emissivity as a function of position throughout the plasma, and the energy spectra of the emitted neutrons. In principle, such measurements might be expected to be relatively easy. That this is not the case is, in part, attributable to practical problems of accessibility to a harsh environment but is mostly a consequence of the time-scale on which the measurements have to be made and of the wide range of neutron emission intensities that have to be covered: for tokamak studies, the time-scale is of the order of 1 to 100 ms and the neutron intensity ranges from 10 12 to 10 19 s -1 . (author)

  12. Numerical Tokamak Project code comparison

    International Nuclear Information System (INIS)

    Waltz, R.E.; Cohen, B.I.; Beer, M.A.

    1994-01-01

    The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

  13. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group

    2003-01-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  14. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    McWilliams, R.

    1988-01-01

    The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  15. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-01-01

    The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

  16. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

    2003-07-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  17. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

    2003-12-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  18. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes

    2003-01-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  19. Anomalous transport in tokamaks

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1989-01-01

    A review is presented of what is known about anomalous transport in tokamaks. It is generally thought that this anomalous transport is the result of fluctuations in various plasma parameters. In the plasma edge detailed measurements of the quantities required to directly determine the fluctuation driven fluxes are available. The total flux of particles is well explained by the measured electrostatic fluctuation driven flux. However, a satisfactory model to explain the origin of the fluctuations has not been identified. The processes responsible for determining the edge energy flux are less clear, but electrostatic convection plays an important part. In the confinement region experimental observations are presently restricted to measurements of density and potential fluctuations and their correlations. The characteristics of the measured fluctuations are discussed and compared with the predictions of various models. Comparisons between measured particle, electron heat and ion heat fluxes, and those fluxes predicted to result from the measured fluctuations, are made. Magnetic fluctuations is discussed

  20. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-09-01

    A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  1. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-01-01

    A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  2. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  3. Plasma turbulence in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)

  4. Disruptions in Tokamaks

    International Nuclear Information System (INIS)

    Bondeson, A.

    1987-01-01

    This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)

  5. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  6. Gamma ray imager on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.; Van Zeeland, M. A.; Watkins, M. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Cooper, C. M. [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California-San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Riso, V. [State University of New York-Buffalo, 12 Capen Hall, Buffalo, New York 14260-1660 (United States)

    2016-04-15

    A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1–60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.

  7. Real-time horizontal position control for Aditya-upgrade tokamak

    International Nuclear Information System (INIS)

    Kumar, Rohit; Ghosh, Joydeep; Tanna, Rakesh L.

    2015-01-01

    Position of plasma column is required to be controlled in real time for improved operation of any tokamak. A PID based system for real-time horizontal plasma position control has been designed for Aditya Upgrade tokamak. Modelling of transfer functions of actuators, plasma and diagnostic system are carried out for ADITYA-U tokamak. The PID controller is optimized using MATLAB-SIMULINK for horizontal position control. Further feed-forward loop is implemented where disturbance due to density variation is suppressed, which results in improved performance as compared to conventional PID operation. In this paper the detailed design of the whole system for real time control of plasma horizontal position in Aditya Upgrade tokamak is presented. (author)

  8. Operational region and sawteeth oscillation in the EAST tokamak

    International Nuclear Information System (INIS)

    Liu, H Q; Gao, X; Zhao, J Y; Hu, L Q; Jie, Y X; Ling, B L; Xu, Q; Ti, A; Ming, T F; Yang, Y; Wu, Z W; Wang, J; Xu, G S; Gao, W; Zhong, G Q; Zang, Q; Shi, Y J; Shen, B; Zhou, Q; Li, Y D; Gong, X Z; Hu, J S; Sun, Y W; Zhao, Y P; Luo, J R; Mao, J S; Weng, P D; Wan, Y X; Zhang, X D; Wan, B N; Li, J

    2007-01-01

    The first plasma discharges were successfully achieved on the experimental advanced superconducting tokamak (EAST) in 2006. The sawteeth behaviours were observed by means of soft x-ray diagnostics and ECE signals in the EAST. The displacement and radius of the q = 1 surface was studied and compared with the result of equilibrium calculation. The density sawtooth oscillation was also observed by the HCN laser interferometer diagnostics. The structure of the EAST operational region was studied in detail. Plasma performance was obviously improved by the boronization and wall conditioning. It was observed that lower q a and a wider stable operating region is extended by the GDC boronization

  9. Stability at high performance in the MAST spherical tokamak

    International Nuclear Information System (INIS)

    Buttery, R.J.; Akers, R.; Arends, E. =

    2003-01-01

    The development of reliable H-modes on MAST, together with advances in heating power and a range of powerful diagnostics, has provided a platform to enable MAST to address some of he most important issues of tokamak stability. In particular the high β potential of the ST is highlighted with stable operation at β N ∼5-6 , β T ∼ 16% and β p as high as 1.9, confirmed by a range of profile diagnostics. Calculations indicate that β N levels are in the vicinity of no-wall stability limits. Studies have provided the first identification of the Neoclassical Tearing Mode (NTM) in the ST, using its behaviour to quantitatively validate predictions of NTM theory, previously only applied to conventional tokamaks. Experiments have demonstrated that sawteeth play a strong role in triggering NTMs - by avoiding large sawteeth much higher β N can, and has, been reached. Further studies have confirmed the NTM's significance, with large islands observed using the 300 point Thomson diagnostic, and locking of large n=1 modes frequently leading to disruptions. H-mode plasmas are also limited by ELMs, with confinement degraded as ELM frequency rises. However, unlike the conventional tokamak, the ELMs in high performing regimes on MAST (H IPB98Y2 ∼1) appear to be type III in nature. Modelling identifies instability to peeling modes, consistent with a type III interpretation, and shows considerable scope to raise pressure gradients (despite n=∞ ballooning theory predictions of instability) before ballooning type modes (perhaps associated with type I ELMs) occur. Finally sawteeth are shown not to remove the q=1 surface in the ST - other promising models are being explored. Thus research on MAST is not only demonstrating stable operation at high performance levels, and developing methods to control instabilities; it is also providing detailed tests of the stability physics and models applicable to conventional tokamaks, such as ITER. (author)

  10. ITER diagnostic system: Vacuum interface

    International Nuclear Information System (INIS)

    Patel, K.M.; Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L.; Drevon, J.M.; Encheva, A.; Kashchuk, Y.; Maquet, Ph.; Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J.

    2013-01-01

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10 −7 Pa, irrespective of plasma operation, and a leak rate of less than 10 −10 Pa m 3 s −1 . In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions will be described

  11. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

    Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

    2013-10-15

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

  12. Preliminary project of s Thomson scattering system for the ETE tokamak; Projeto preliminar de um sistema de espalhamento Thomson para o Tokamak ETE

    Energy Technology Data Exchange (ETDEWEB)

    Berni, Luiz Angelo

    1997-12-31

    This report presents the preliminary project of the injection and laser light block system for the Thomson (ET) scattering diagnostic to be implanted at the ETE spheric tokamak of the Instituto Nacional de Pesquisas Espaciais (INPE/LAP). Also, a scanning system for the optics of scattered light 4 refs., 26 figs.

  13. Instrumentation and controls of an ignited tokamak

    International Nuclear Information System (INIS)

    Becraft, W.R.; Golzy, J.; Houlberg, W.A.; Kukielka, C.A.; Onega, R.J.; Raju, G.V.S.; Stone, R.S.

    1980-10-01

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented

  14. Properties of the tokamak edge plasma

    International Nuclear Information System (INIS)

    Wolff, H.

    1988-01-01

    A short review of some features of the edge plasma in limiter tokamaks is given. The limits of the simple one-dimensional scrape-off layer (SOL) model and the relation between the core plasma are discussed. Multifaceted asymmetric radiation from the edge (MARFE) phenomena and detached plasma are closely connected with the particle and energy balance of the SOL. Their occurrence is based on the relation of plasma parameters of the edge plasma to those of the core. Important problems of plasma wall interactions are the detection of the impurity sources and sinks and the study of the impurity transport and shielding. The non-uniform character of plasma wall interactions and their dependence on the discharge performance still renders difficult any theoretical forecast of impurity distribution and transport and calls for better diagnostics. (author)

  15. CXSFIT Code Application to Process Charge-Exchange Recombination Spectroscopy Data at the T-10 Tokamak

    Science.gov (United States)

    Serov, S. V.; Tugarinov, S. N.; Klyuchnikov, L. A.; Krupin, V. A.; von Hellermann, M.

    2017-12-01

    The applicability of the CXSFIT code to process experimental data from Charge-eXchange Recombination Spectroscopy (CXRS) diagnostics at the T-10 tokamak is studied with a view to its further use for processing experimental data at the ITER facility. The design and operating principle of the CXRS diagnostics are described. The main methods for processing the CXRS spectra of the 5291-Å line of C5+ ions at the T-10 tokamak (with and without subtraction of parasitic emission from the edge plasma) are analyzed. The method of averaging the CXRS spectra over several shots, which is used at the T-10 tokamak to increase the signal-to-noise ratio, is described. The approximation of the spectrum by a set of Gaussian components is used to identify the active CXRS line in the measured spectrum. Using the CXSFIT code, the ion temperature in ohmic discharges and discharges with auxiliary electron cyclotron resonance heating (ECRH) at the T-10 tokamak is calculated from the CXRS spectra of the 5291-Å line. The time behavior of the ion temperature profile in different ohmic heating modes is studied. The temperature profile dependence on the ECRH power is measured, and the dynamics of ECR removal of carbon nuclei from the T-10 plasma is described. Experimental data from the CXRS diagnostics at T-10 substantially contribute to the implementation of physical programs of studies on heat and particle transport in tokamak plasmas and investigation of geodesic acoustic mode properties.

  16. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Hively, L.M.; Sigmar, D.J.

    1989-09-01

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

  17. Tokamak engineering test reactor

    International Nuclear Information System (INIS)

    Conn, R.W.; Jassby, D.L.

    1975-07-01

    The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

  18. START: the creation of a spherical tokamak

    International Nuclear Information System (INIS)

    Sykes, Alan

    1992-01-01

    The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)

  19. Heavy Neutral Beam Probe for edge plasma analysis in tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    The Heavy Neutral Beam Probe project presented in this document is part of an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadian de Fusion Magnetique. The overall objective of the effort is to apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes facility in Montreal, Canada. To achieve this goal, a research and development project was started in December, 1990 to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present, satisfactory progress has been achieved. The ion gun is fully operational with the neutralizer in the final assembly stage in preparation for testing. The beam diagnostics have been completed and mounted in the computer automated test stand. The analyzer design and detailed trajectory calculations are nearing completion to allow for the vacuum interface construction. The CAMAC based data acquisition system hardware was integrated into the test stand. Part of this hardware is a component of the Tokamak de Varennes' contribution to the collaboration. Next steps on the critical path include the beginning of the neutralization tests and the start of the analyzer construction. Anticipated installation of the diagnostic on the tokamak is Spring 1992

  20. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  1. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  2. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  3. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  4. Tokamak Physics Experiment (TPX) design

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1995-01-01

    TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

  5. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

    Mardani, M.; Amrollahi, R.; Koohestani, S.

    2012-01-01

    Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

  6. LIDAR Thomson scattering for advanced tokamaks. Final report

    International Nuclear Information System (INIS)

    Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

    1996-01-01

    The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured

  7. Technology Issues of Burning Plasma Diagnostics

    International Nuclear Information System (INIS)

    Kaye, A. S.

    2008-01-01

    The ITER Tokamak will require many diagnostics both for safe and reliable operation of the machine and for understanding of the physics underlying the performance. The design of these diagnostics raises many challenging technical issues not faced on smaller machines. These arise partly from the increase demands on established diagnostics arising from the increased size, higher magnetic field, large heating power, and in particular the dramatically longer pulse duration of ITER, which make issue such as power loading on first wall components more challenging. The demands on reliability and availability of the machine in order to achieve the objectives within the agreed time schedule also place severe additional demands on the design, quality assurance and maintainability of diagnostics. ITER will produce many orders of magnitude more neutrons than previous Tokamaks and will be a licensed nuclear facility. This has important implications for the traceability, quality assurance and availability of safety critical diagnostics, and for the control of the design and procurement of all diagnostics. The high neutron flux/fluence also constrains the design of diagnostics, which must offer shielding consistent with the allowable dose rates on critical components of the Tokamak, and themselves be tolerant of the radiation level at the diagnostic. This paper presents an overview of the more critical issues for ITER diagnostics

  8. Resistive instabilities in tokamaks

    International Nuclear Information System (INIS)

    Rutherford, P.H.

    1985-10-01

    Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed

  9. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  10. Classical tokamak transport theory

    International Nuclear Information System (INIS)

    Nocentini, Aldo

    1982-01-01

    A qualitative treatment of the classical transport theory of a magnetically confined, toroidal, axisymmetric, two-species plasma is presented. The 'weakly collisional' ('banana' and 'plateau') and 'collision dominated' ('Pfirsch-Schlueter' and 'highly collisional') regimes, as well as the Ware effect are discussed. The method used to evaluate the diffusion coffieicnts of particles and heat in the weakly collisional regime is based on stochastic argument, that requires an analysis of the characteristic collision frequencies and lengths for particles moving in a tokamak-like magnetic field. The same method is used to evaluate the Ware effect. In the collision dominated regime on the other hand, the particle and heat fluxes across the magnetic field lines are dominated by macroscopic effects so that, although it is possible to present them as diffusion (in fact, the fluxes turn out to be proportional to the density and temperature gradients), a macroscopic treatment is more appropriate. Hence, fluid equations are used to inveatigate the collision dominated regime, to which particular attention is devoted, having been shown relatively recently that it is more complicated than the usual Pfirsch-Schlueter regime. The whole analysis presented here is qualitative, aiming to point out the relevant physical mechanisms involved in the various regimes more than to develop a rigorous mathematical derivation of the diffusion coefficients, for which appropriate references are given. (author)

  11. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  12. The role of the spherical tokamak in clarifying tokamak physics

    International Nuclear Information System (INIS)

    Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.

    1999-01-01

    The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)

  13. Maintenance concept development for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Macdonald, D.

    1988-01-01

    The Compact Ignition Tokamak (CIT), located at the Princeton Plasma Physics Laboratory, will be the next major experimental machine in the US Fusion Program. Its use of deuterium-tritium (D-T) fuel requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist of removing and repairing such components as diagnostic equipment modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the vacuum vessel includes both bridge-mounted and floor-mounted manipulator systems. Additionally, decontamination (decon) equipment, hot cell repair facilities, and equipment for handling and packaging solid radioactive waste (rad-waste) are being developed. Recent design activities have focused on establishing maintenance system interfaces with the facility design, developing manipulator system requirements, and using mock-ups to support the tokamak configuration design. 3 refs., 8 figs

  14. New results from the Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.; Ananiev, A.S.; Amoskov, V.M.

    2003-01-01

    New results from the Globus-M spherical tokamak are presented. High plasma current of 0.36 MA, high toroidal magnetic field of 0.55 T and other important plasma characteristics were achieved. Described are the operational space and plasma stability limits in the OH regime. The factors limiting operational space (MHD instabilities, runaway electrons, etc.) are discussed. New experiments on plasma fuelling are described. First results of experiments with a coaxial plasma gun injector are presented. Initial results of a plasma - wall interaction study are outlined. First results obtained with new diagnostic tools installed on the tokamak are presented. An auxiliary heating system test was performed. Preliminary results of simulations and experiments are given. (author)

  15. New results from Globus-M spherical tokamak

    International Nuclear Information System (INIS)

    Gusev, V.K.

    2002-01-01

    New results from Globus-M spherical tokamak (ST) are presented. Reported are the achievements of high plasma current of 0.36 MA and high toroidal magnetic field of 0.55 T. Plasma column stability in Globus-M is conserved at low edge safety factors and high plasma densities. Achieved lowest safety factor was q(cyl) 19 m -3 . New methods of density increase are discussed. Low-density boarder of operational space is investigated. Runaway electrons properties and conditions of their generation are investigated. Results look promising for STs. Plasma-wall interaction study was performed. Silicon probes were installed into vacuum vessel. They were exposed to boronization, first, and then deposited film interacted with plasma. Discussed are film properties. Briefly described are new diagnostic tools installed on tokamak. Status and preliminary results obtained with auxiliary heating systems are shown. (author)

  16. Real time equilibrium reconstruction for tokamak discharge control

    International Nuclear Information System (INIS)

    Ferron, J.R.; Walker, M.L.; Lao, L.L.; St John, H.E.; Humphreys, D.A.; Leuer, J.A.

    1998-01-01

    A practical method for performing a tokamak equilibrium reconstruction in real time for arbitrary time varying discharge shapes and current profiles is described. An approximate solution to the Grad-Shafranov equilibrium relation is found which best fits the diagnostic measurements. Thus, a solution for the spatial distribution of poloidal flux and toroidal current density is available in real time that is consistent with plasma force balance, allowing accurate evaluation of parameters such as discharge shape and safety factor profile. The equilibrium solutions are produced at a rate sufficient for discharge control. This equilibrium reconstruction algorithm has been implemented on the digital plasma control system for the DIII-D tokamak. The first application of real time equilibrium reconstruction to discharge shape control is described. (author)

  17. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.

    2001-01-01

    A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

  18. Large aspect ratio tokamak study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Sardella, C.; Wiseman, G.W.

    1979-01-01

    The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured

  19. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (n e and T e ) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  20. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

    1981-02-01

    Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

  1. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1994-01-01

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  2. Thermonuclear Tokamak plasmas in the presence of fusion alpha particles

    International Nuclear Information System (INIS)

    Anderson, D.; Hamnen, H.; Lisak, M.

    1988-01-01

    In this overview, we have focused on several results of the thermonuclear plasma research pertaining to the alpha particle physics and diagnostics in a fusion tokamak plasma. As regards the discussion of alpha particle effects, two distinct classes of phenomena have been distinguished: the simpler class containing phenomena exhibited by individual alpha particles under the influence of bulk plasma properties and, the more complex class including collective effects which become important for increasing alpha particle density. We have also discussed several possibilities to investigate alpha particle effects by simulation experiments using an equivalent population of highly energetic ions in the plasma. Generally, we find that the present theoretical knowledge on the role of fusion alpha particles in a fusion tokamak plasma is incomplete. There are still uncertainties and partial lack of quantitative results in this area. Consequently, further theoretical work and, as far a possible, simulation experiments are needed to improve the situation. Concerning the alpha particle diagnostics, the various diagnostic techniques and the status of their development have been discussed in two different contexts: the escaping alpha particles and the confined alpha particles in the fusion plasma. A general conclusion is that many of the different diagnostic methods for alpha particle measurements require further major development. (authors)

  3. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  4. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    Zacek, F.; Badalec, J.; Jakubka, J.; Kryska, L.; Preinhaelter, J.; Stoeckel, J.; Valovic, M.; Nanobashvili, S.; Weixelbaum, L.; Wenzel, U.; Spineanu, F.; Vlad, M.

    1990-10-01

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

  5. Electron and current density measurements on tokamak plasmas

    International Nuclear Information System (INIS)

    Lammeren, A.C.A.P. van.

    1991-01-01

    The first part of this thesis describes the Thomson-scattering diagnostic as it was present at the TORTUR tokamak. For the first time with this diagnostic a complete tangential scattering spectrum was recorded during one single laser pulse. From this scattering spectrum the local current density was derived. Small deviations from the expected gaussian scattering spectrum were observed indicating the non-Maxwellian character of the electron-velocity distribution. The second part of this thesis describes the multi-channel interferometer/ polarimeter diagnostic which was constructed, build and operated on the Rijnhuizen Tokamak Project (RTP) tokamak. The diagnostic was operated routinely, yielding the development of the density profiles for every discharge. When ECRH (Electron Cyclotron Resonance Heating) is switched on the density profile broadens, the central density decreases and the total density increases, the opposite takes place when ECRH is switched off. The influence of MHD (magnetohydrodynamics) activity on the density was clearly observable. In the central region of the plasma it was measured that in hydrogen discharges the so-called sawtooth collapse is preceded by an m=1 instability which grows rapidly. An increase in radius of this m=1 mode of 1.5 cm just before the crash is observed. In hydrogen discharges the sawtooth induced density pulse shows an asymmetry for the high- and low-field side propagation. This asymmetry disappeared for helium discharges. From the location of the maximum density variations during an m=2 mode the position of the q=2 surface is derived. The density profiles are measured during the energy quench phase of a plasma disruption. A fast flattening and broadening of the density profile is observed. (author). 95 refs.; 66 figs.; 7 tabs

  6. Reversed field pinch diagnostics

    International Nuclear Information System (INIS)

    Weber, P.G.

    1986-01-01

    The Reversed Field Pinch (RFP) is a toroidal, axisymmetric magnetic confinement configuration characterized by a magnetic field configuration in which the toroidal magnetic field is of similar strength to the poloidal field, and is reversed at the edge compared to the center. The RFP routinely operates at high beta, and is a strong candidate for a compact fusion device. Relevant attributes of the configuration will be presented, together with an overview of present and planned experiments and their diagnostics. RFP diagnostics are in many ways similar to those of other magnetic confinement devices (such as tokamaks); these lectures will point out pertinent differences, and will present some diagnostics which provide special insights into unique attributes of the RFP

  7. Corpuscular plasma diagnostics

    International Nuclear Information System (INIS)

    Afrosimov, V.; Petrov, M.

    1984-01-01

    An elementary explanation is presented of the physical principles and important methods of corpuscular plasma diagnostics. The invaluable role of corpuscular methods for measuring the hot plasma ion component in thermonuclear facilities, especially hydrogen ions in tokamaks, is emphasized. All corpuscular methods employ analysis of fast neutral atoms and therefore the mechanism of their creation inside a hot plasma is explained first. The ammount of information obtainable from spectra of fast neutrals is discussed. Multichannel analyzers developed at the FTI A.F. Ioffe in Leningrad are described in detail. Classical passive corpuscular diagnostics are examined as are active methods using artifitial beams of hydrogen atoms. The method used for obtaining local values of ion temperature and density is explained. Corpuscular spectroscopic diagnostics and its application for measuring impurities is mentioned. (J.U.)

  8. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  9. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...

  10. Experimental test of far-infrared polarimetry for Faraday rotation measurements on the TFR 600 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Soltwisch, H [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Plasmaphysik; Association Euratom-Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.)); Equipe, T F.R. [Association Euratom-CEA sur la Fusion, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France). Dept. de Recherches sur la Fusion Controlee

    1981-09-01

    The results are reported on the feasibility of using far-infrared polarimetry for Faraday rotation diagnostic measurements on the TRF Tokamak. Precise quantitative results were not obtained but a satisfactory agreement with a simple theoretical model leads to a good understanding of the experimental limitations of the method.

  11. Preliminary project of s Thomson scattering system for the ETE tokamak

    International Nuclear Information System (INIS)

    Berni, Luiz Angelo

    1997-01-01

    This report presents the preliminary project of the injection and laser light block system for the Thomson (ET) scattering diagnostic to be implanted at the ETE spheric tokamak of the Instituto Nacional de Pesquisas Espaciais (INPE/LAP). Also, a scanning system for the optics of scattered light

  12. Fast-ion transport induced by Alfvén eigenmodes in the ASDEX Upgrade tokamak

    DEFF Research Database (Denmark)

    Garcia-Munoz, M.; Classen, I.G.J.; Geiger, B.

    2011-01-01

    A comprehensive suite of diagnostics has allowed detailed measurements of the Alfvén eigenmode (AE) spatial structure and subsequent fast-ion transport in the ASDEX Upgrade (AUG) tokamak [1]. Reversed shear Alfvén eigenmodes (RSAEs) and toroidal induced Alfvén eigenmodes (TAEs) have been driven u...

  13. Injection of pellets into the TCA tokamak

    International Nuclear Information System (INIS)

    Martin, Y.

    1993-05-01

    This thesis presents experimental results from the analysis of the ablation process of pellets injected into the TCA tokamak. The determination of scaling laws relating the pellet penetration to the pellet and plasma parameters preceding injection, were used to improve the understanding of the interaction of the pellet with the plasma since a) the pellet and plasma conditions preceding injection were varied over a large range, and b) the estimation of the penetration depth takes into account the influence of striations in the deposition profile. Over 400 pellets with a range of sizes and speeds were injected into a range of plasma parameters in order to create a database from which the scaling laws could be deduced. The ablation characteristics were principally measured with two CCD video cameras, which provided good spatial resolution, and two filtered photomultiplier tubes, which provided good temporal resolution of the light emitted from the pellet ablation cloud. In the text, the traditional methods of analysing these diagnostics are examined with special reference to the presumptions that a) the pellet velocity is constant in the plasma, and b) the light intensity determined from the ablation cloud is proportional to the ablation rate. After successive data reduction from the database, in order to separate the effects of varying different parameters, the main observations were that, a) the pellet penetration varies as the square root of the pellet velocity, b) the scaling laws for the other parameters strongly depend on whether the pellet has sufficient velocity to reach the q=1 rational magnetic surface in the tokamak. (author) 45 refs

  14. Energy losses on tokamak startup

    International Nuclear Information System (INIS)

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1983-01-01

    During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells

  15. Integral torque balance in tokamaks

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2011-01-01

    The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.

  16. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately

  17. Tokamak and RFP ignition requirements

    International Nuclear Information System (INIS)

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

  18. Tokamak impurity-control techniques

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1980-01-01

    A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)

  19. An enhanced tokamak startup model

    Science.gov (United States)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

    The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

  20. Multimegawatt neutral beams for tokamaks

    International Nuclear Information System (INIS)

    Kunkel, W.B.

    1979-03-01

    Most of the large magnetic confinement experiments today and in the near future use high-power neutral-beam injectors to heat the plasma. This review briefly describes this remarkable technique and summarizes recent results as well as near term expectations. Progress has been so encouraging that it seems probable that tokamaks will achieve scientific breakeven before 1990

  1. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254 ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  2. Interferometric investigation methods of plasma spatial characteristics on stellarators and tokamaks in submillimeter region

    International Nuclear Information System (INIS)

    Berezhnyj, V.L.; Kononenko, V.I.; Epishin, V.A.; Topkov, A.N.

    1992-01-01

    The review of interferometric methods of plasma investigation in the wave submillimeter range is given. The diagnostic schemes in stellarators and tokamaks designed for experienced thermonuclear reactors and also the perspective ones, which are still out of practice, are shown. The methods of these diagnostics, their physical principles, the main possibilities and restrictions at changes of electron density, magnetic fields (currents) and their spatial distributions are described. 105 refs.; 9 figs.; 2 tables. (author)

  3. Magnetic measurements using array of integrated Hall sensors on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Ďuran, Ivan; Hronová-Bilyková, Olena; Stöckel, Jan; Sentkerestiová, J.; Havlíček, Josef

    2008-01-01

    Roč. 79, č. 10 (2008), 10F123-10F123 ISSN 0034-6748. [Topical Conference on High-Temperature Plasma Diagnostics/17th./. Albuquerque, 11.05.2008-15.05.2008] R&D Projects: GA MPO 2A-1TP1/101 Institutional research plan: CEZ:AV0Z20430508 Keywords : Galvanomagnetic Sensor * Fusion Reactor * Magnetic Diagnostics * CASTOR tokamak Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.738, year: 2008

  4. Low cost alternative of high speed visible light camera for tokamak experiments

    Czech Academy of Sciences Publication Activity Database

    Odstrčil, T.; Odstrčil, Michal; Grover, O.; Svoboda, V.; Ďuran, Ivan; Mlynář, Jan

    2012-01-01

    Roč. 83, č. 10 (2012), 10E505-10E505 ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostic * high speed camera * GOLEM Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://dx.doi.org/10.1063/1.4731003

  5. Overview of the ITER Tokamak complex building and integration of plant systems toward construction

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, Jean-Jacques, E-mail: jean-jacques.cordier@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2015-10-15

    The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed

  6. Characterization of disruptions in the Microwave Tokamak Experiment, MTX

    International Nuclear Information System (INIS)

    Hooper, E.B.; Makowski, M.A.

    1990-01-01

    The Microwave Tokamak Experiment (MTX) has a substantial number of fast diagnostics, especially for electrons, as part of its mission for pulsed, high-power electron cyclotron heating. As part of its contribution to ITER R ampersand D, these diagnostics are being used to characterize disruptions in MTX. This report is the first of two, with the second planned for submittal in September 1990, at the end of the ITER conceptual design activity. Here, we analyze the characteristics of disruptions during normal operation of MTX, discuss some new data pertaining to the ''Granetz limit,'' and describe preliminary data on ramped density shorts which will be used for fast measurements on density limit disruptions. The final report will discuss measurements using the fast diagnostics to characterize the disruption

  7. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-01-01

    This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

  8. The LIDAR Thomson Scattering Diagnostic on JET

    DEFF Research Database (Denmark)

    Salzmann, H.; Bundgaard, J.; Gadd, A.

    1988-01-01

    By combining the time‐of‐flight or LIDAR principle with a Thomson backscatter diagnostic, spatial profiles of the electron temperature and density are measured in a magnetically confined fusion plasma. This technique was realized for the first time on the JET tokamak. A ruby laser (3‐J pulse ener...

  9. Design of neutron diagnostic for MTX

    International Nuclear Information System (INIS)

    Ogawa, Toshihide; Oasa, Kazumi; Hoshino, Katsumichi; Odajima, Kazuo; Maeda, Hikosuke

    1990-07-01

    A neutron diagnostic system was designed for the Microwave Tokamak Experiment being carried out at the lawrence Livermore National Laboratory. High speed measurements are important to this experiment. Plastic scintillator is used for this fast response detection of neutron. Proportional counters and fission counters are used for the total neutron emission rate measurements. (author)

  10. Progress in Neutron Diagnostics at JET

    Czech Academy of Sciences Publication Activity Database

    Mlynář, Jan; Bonheure, G.; Murari, A.; Bertalot, L.; Angelone, M.; Pillon, M.; Conroy, S.; Ericsson, G.; Kaellne, J.; Popovichev, S.

    2006-01-01

    Roč. 56, suppl.B (2006), B118-B124 ISSN 0011-4626. [Symposium on Plasma Physics and Technology /22nd./. Praha, 26.6.2006-29.6.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * JET * fusion * neutrons diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.568, year: 2006

  11. Gated integrator PXI-DAQ system for Thomson scattering diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Patel, Kiran, E-mail: kkpatel@ipr.res.in; Pillai, Vishal; Singh, Neha; Thomas, Jinto; Kumar, Ajai

    2017-06-15

    Gated Integrator (GI) PXI based data acquisition (DAQ) system has been designed and developed for the ease of acquiring fast Thomson Scattered signals (∼50 ns pulse width). The DAQ system consists of in-house designed and developed GI modules and PXI-1405 chassis with several PXI-DAQ modules. The performance of the developed system has been validated during the SST-1 campaigns. The dynamic range of the GI module depends on the integrating capacitor (C{sub i}) and the modules have been calibrated using 12 pF and 27 pF integrating capacitors. The developed GI module based data acquisition system consists of sixty four channels for simultaneous sampling using eight PXI based digitization modules having eight channels per module. The error estimation and functional tests of this unit are carried out using standard source and also with the fast detectors used for Thomson scattering diagnostics. User friendly Graphical User Interface (GUI) has been developed using LabVIEW on Windows platform to control and acquire the Thomson scattering signal. A robust, easy to operate and maintain with low power consumption, having higher dynamic range with very good sensitivity and cost effective DAQ system is developed and tested for the SST-1 Thomson scattering diagnostics.

  12. Wavelength calibration of x-ray imaging crystal spectrometer on Joint Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Yan, W.; Chen, Z. Y.; Jin, W.; Huang, D. W.; Ding, Y. H.; Li, J. C.; Zhang, X. Q.; Zhuang, G.; Lee, S. G.; Shi, Y. J.

    2014-01-01

    The wavelength calibration of x-ray imaging crystal spectrometer is a key issue for the measurements of plasma rotation. For the lack of available standard radiation source near 3.95 Å and there is no other diagnostics to measure the core rotation for inter-calibration, an indirect method by using tokamak plasma itself has been applied on joint Texas experimental tokamak. It is found that the core toroidal rotation velocity is not zero during locked mode phase. This is consistent with the observation of small oscillations on soft x-ray signals and electron cyclotron emission during locked-mode phase

  13. Overview of wall probes for erosion and deposition studies in the TEXTOR tokamak

    Directory of Open Access Journals (Sweden)

    M. Rubel

    2017-05-01

    Full Text Available An overview of diagnostic tools – test limiters and collector probes – used over the years for material migration studies in the TEXTOR tokamak is presented. Probe transfer systems are shown and their technical capabilities are described. This is accompanied by a brief presentation of selected results and conclusions from the research on material erosion – deposition processes including tests of candidate materials (e.g. W, Mo, carbon-based composites for plasma-facing components in controlled fusion devices. The use of tracer techniques and methods for analysis of materials retrieved from the tokamak are summarized. The impact of research on the reactor wall technology is addressed.

  14. Simultaneous measurement of line electron density and Faraday rotation in the ISX-B tokamak

    International Nuclear Information System (INIS)

    Hutchinson, D.P.; Ma, C.H.; Staats, P.A.; Vander Sluis, K.L.

    1981-01-01

    A new diagnostic system utilizing a submillimetre-wave, phase-modulated polarimeter/interferometer has been used to simultaneously measure the time evolution of the line-averaged electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The measurements, performed along four chords of the plasma column, have been correlated with poloidal field changes associated with a ramp in the Ohmic-heating current and by neutral-beam injection. These are the first simultaneous measurements of line electron density and Faraday rotation to be made along a chord of submillimetre laser beam in a tokamak plasma. (author)

  15. ICIT contribution to JET gamma-ray diagnostics enhancement

    International Nuclear Information System (INIS)

    Soare, S.; Curuia, M.; Zoita, V.

    2010-01-01

    Full text: Gamma-ray emission of tokamak plasmas is the result of the interaction of fast ions (fusion reaction products, including alpha particles, NBI ions, ICRH-accelerated ions) with main plasma impurities (e.g., carbon, beryllium). Gamma-ray diagnostics involve both gamma-ray imaging (cameras) and gamma-ray spectrometry (spectrometers). For the JET tokamak, gamma-ray diagnostics have been used to provide information on the characteristics of the fast ion population in plasmas. Two gamma-ray diagnostics enhancements project have been launched by JET and the MEdC/EURATOM Association has agreed to lead both of them with ICIT as projects leader. (authors)

  16. Analysis of disruptive instabilities in Aditya tokamak discharges

    International Nuclear Information System (INIS)

    Chattopadhyay, Asim Kumar; Anand, Arun; Rao, C.V.S.; Joisa, Shankar; Aditya team

    2006-01-01

    Major disruptions and sawteeth oscillations (internal disruptions) are routinely observed in ohmically heated Aditya tokamak discharges and their characteristics have been investigated with the help of soft x-ray (SXR) tomography along with other diagnostics. The SXR tomography is carried out with the help of single array of detectors assuming rigid rotation of the modes to analyse the mode structure of sawtooth internal disruptions. Coupling of m/n = 2/1 and m/n=1/1 modes could be the main mechanism for the major disruption. Sawteeth periods were measured and compared with the scaling laws and found to be in good agreement. (author)

  17. Evaluation of the plasma parameters in COMPASS tokamak divertor area

    Czech Academy of Sciences Publication Activity Database

    Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud

    2012-01-01

    Roč. 356, č. 1 (2012), s. 012007 ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron, and IonTechnologies(VEIT2011)/17./. Sunny Beach, 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf

  18. On Use of Semiconductor Detector Arrays on COMPASS Tokamak

    Czech Academy of Sciences Publication Activity Database

    Weinzettl, Vladimír; Imríšek, Martin; Havlíček, Josef; Mlynář, Jan; Naydenkova, Diana; Háček, Pavel; Hron, Martin; Janky, Filip; Sarychev, D.; Berta, M.; Bencze, A.; Szabolics, T.

    -, č. 71 (2012), s. 844-850 ISSN 2010-376X. [ICPP 2012 : International Conference on Plasma Physics. Venice, 14.11.2012-16.11.2012] R&D Projects: GA ČR GA202/09/1467; GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : bolometry * plasma diagnostics * soft X-rays * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics https://www.waset.org/journals/waset/v71/v71-143.pdf

  19. Multiple view fan beam polarimetry on Tokamak devices

    International Nuclear Information System (INIS)

    Geck, W.R.; Domier, C.W.; Luhmann, N.C.

    1997-01-01

    A polarimeter diagnostic is under development which utilizes several fan beams to accumulate line integrated Faraday rotation data in a Tokamak plasma. The utilization of a fan beam configuration over that of conventional vertical view polarimeter systems significantly reduces access requirements. The high angular separation inherent in a fan beam implementation increases plasma coverage and eliminates the necessity of assumed plasma symmetries to generate high quality current density profiles. Codes have been developed to generate these high-resolution two-dimensional images of the plasma current profile from data collected at arbitrary positions and viewing angles. copyright 1997 American Institute of Physics

  20. TFR, the tokamak of Fontenay-aux-Roses

    International Nuclear Information System (INIS)

    1985-06-01

    Twelve years of the TFR tokamak operation are briefly reviewed. In the historical introduction, justifications of the experiment and main development stages are described. The importance of the choice of materials in contact with the plasma (impurities production) appears in the various paragraphs. The power lost by impurity radiation is compared with other losses in ohmic conditions (paragraph 2). Additional heating experiments: neutral beam heating paragraph 3 and ion cyclotron heating paragraph 4 are reported; their efficiency as well as deleterious induced effects are described. Important diagnostics development to measure impurities were made, giving experimental results used also in astrophysics and atomic physics. In the last paragraph magnetohydrodynamic phenomena are reported

  1. Interactive exploration of tokamak turbulence simulations in virtual reality

    International Nuclear Information System (INIS)

    Kerbel, G.D.; Pierce, T.; Milovich, J.L.; Shumaker, D.E.

    1996-01-01

    We have developed an immersive visualization system designed for interactive data exploration as an integral part of our computing environment for studying tokamak turbulence. This system of codes can reproduce the results of simulations visually for scrutiny in real time, interactively and with more realism than ever before. At peak performance, the VR system can present for view some 400 coordinated images per second. The long term vision this approach targets is a open-quote holodeck-like close-quote virtual-reality environment in which one can explore gyrofluid or gyrokinetic plasma simulations interactively and in real time, visually, with concurrent simulations of experimental diagnostic devices. In principle, such a open-quote virtual tokamak close-quote computed environment could be as all encompassing or as focussed as one likes, in terms of the physics involved. The computing framework in one within which a group of researchers can work together to produce a real and identifiable product with easy access to all contributions. This could be our version of NASA's next generation Numerical Wind Tunnel. The principal purpose of this VR capability for Numerical Tokamak simulation is to provide interactive visual experience to help create new ways of understanding aspects of the convective transport processes operating in tokamak fusion experiments. The effectiveness of the visualization method is strongly dependent on the density of frame-to-frame correlation. Below a threshold of this quantity, short term visual memory does not bridge the gap between frames well enough for there to exist a strong visual connection. Above the threshold, evolving structures appear clearly. The visualizations show the 3D structure of vortex evolution and the gyrofluid motion associated with it. We discovered that it was very helpful for visualizing the cross field flows to compress the virtual world in the toroidal angle

  2. Microwave Tokamak Experiment: Overview and status

    International Nuclear Information System (INIS)

    1990-05-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs

  3. Combined confinement system applied to tokamaks

    International Nuclear Information System (INIS)

    Ohkawa, Tihiro

    1986-01-01

    From particle orbit point of view, a tokamak is a combined confinement configuration where a closed toroidal volume is surrounded by an open confinement system like a magnetic mirror. By eliminating a cold halo plasma, the energy loss from the plasma becomes convective. The H-mode in diverted tokamaks is an example. Because of the favorable scaling of the energy confinement time with temperature, the performance of the tokamak may be significantly improved by taking advantage of this effect. (author)

  4. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

    LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

    1988-04-01

    The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

  5. First Wall and Operational Diagnostics

    International Nuclear Information System (INIS)

    Lasnier, C; Allen, S; Boedo, J; Groth, M; Brooks, N; McLean, A; LaBombard, B; Sharpe, J; Skinner, C; Whyte, D; Rudakov, D; West, W; Wong, C

    2006-01-01

    In this chapter we review numerous diagnostics capable of measurements at or near the first wall, many of which contribute information useful for safe operation of a tokamak. There are sections discussing infrared cameras, visible and VUV cameras, pressure gauges and RGAs, Langmuir probes, thermocouples, and erosion and deposition measurements by insertable probes and quartz microbalance. Also discussed are dust measurements by electrostatic detectors, laser scattering, visible and IR cameras, and manual collection of samples after machine opening. In each case the diagnostic is discussed with a view toward application to a burning plasma machine such as ITER

  6. Improvement of the tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, L

    1994-12-31

    Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.

  7. Advanced commercial Tokamak optimization studies

    International Nuclear Information System (INIS)

    Whitley, R.H.; Berwald, D.H.; Gordon, J.D.

    1985-01-01

    Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units

  8. Flux driven turbulence in tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Ghendrih, P.; Ottaviani, M.; Sarazin, Y.; Beyer, P.; Benkadda, S.; Waltz, R.E.

    1999-01-01

    This work deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts, are usually observed over a broad range of time and spatial scales. The existence of these fronts provide a way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyroBohm type in spite of these large scale transport events. Some departure from the gyroBohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux than at fixed temperature gradient, with the same time averaged profile. (author)

  9. Options for an ignited tokamak

    International Nuclear Information System (INIS)

    Sheffield, J.

    1984-02-01

    It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon β/sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed

  10. Starfire: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

    1979-01-01

    The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

  11. Recent Advancements in Microwave Imaging Plasma Diagnostics

    International Nuclear Information System (INIS)

    Park, H.; Chang, C.C.; Deng, B.H.; Domier, C.W.; Donni, A.J.H.; Kawahata, K.; Liang, C.; Liang, X.P.; Lu, H.J.; Luhmann, N.C. Jr.; Mase, A.; Matsuura, H.; Mazzucato, E.; Miura, A.; Mizuno, K.; Munsat, T.; Nagayama, K.; Nagayama, Y.; Pol, M.J. van de; Wang, J.; Xia, Z.G.; Zhang, W-K.

    2002-01-01

    Significant advances in microwave and millimeter wave technology over the past decade have enabled the development of a new generation of imaging diagnostics for current and envisioned magnetic fusion devices. Prominent among these are revolutionary microwave electron cyclotron emission imaging (ECEI), microwave phase imaging interferometers, imaging microwave scattering and microwave imaging reflectometer (MIR) systems for imaging electron temperature and electron density fluctuations (both turbulent and coherent) and profiles (including transport barriers) on toroidal devices such as tokamaks, spherical tori, and stellarators. The diagnostic technology is reviewed, and typical diagnostic systems are analyzed. Representative experimental results obtained with these novel diagnostic systems are also presented

  12. Magnetic island formation in tokamaks

    International Nuclear Information System (INIS)

    Yoshikawa, S.

    1989-04-01

    The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs

  13. Equilibrium Reconstruction in EAST Tokamak

    International Nuclear Information System (INIS)

    Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)

  14. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

    The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)

  15. Runaway electrons during tokamak startup

    International Nuclear Information System (INIS)

    Sharma, A.S.; Jayakumar, R.

    1988-01-01

    Runaway electrons significantly affect the plasma and impurity evolution during tokamak startup. During its rise, a runaway pulse stores magnetic flux inductively; this is then released during the decay phase of the runaway pulse. This process affects plasma formation, current initiation and current buildup. Because of their relativistic velocities the runaway electrons have higher ionization and excitation rates than the plasma electrons. This leads to a significant modification of the impurity behaviour and consequently the plasma evolution. (author). 20 refs, 8 figs

  16. Minimum scaling laws in tokamaks

    International Nuclear Information System (INIS)

    Zhang, Y.Z.; Mahajan, S.M.

    1986-10-01

    Scaling laws governing anomalous electron transport in tokamaks with ohmic and/or auxiliary heating are derived using renormalized Vlasov-Ampere equations for low frequency electromagnetic microturbulence. It is also shown that for pure auxiliary heating (or when auxiliary heating power far exceeds the ohmic power), the energy confinement time scales as tau/sub E/ ∼ P/sub inj//sup -1/3/, where P/sub inj/ is the injected power

  17. Gyrosheath near the tokamak edge

    International Nuclear Information System (INIS)

    Hazeltine, R.D.; Xiao, H.; Valanju, P.M.

    1993-03-01

    A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results

  18. Tokamak plasma boundary layer model

    International Nuclear Information System (INIS)

    Volkov, T.F.; Kirillov, V.D.

    1983-01-01

    A model has been developed for the limiter layer and for the boundary region of the plasma column in a tokamak to facilitate analytic calculations of the thickness of the limiter layers, the profiles and boundary values of the temperature and the density under various conditions, and the difference between the electron and ion temperatures. This model can also be used to analyze the recycling of neutrals, the energy and particle losses to the wall and the limiter, and other characteristics

  19. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Kieras, C.E.

    1982-12-01

    Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma

  20. Discharge cleaning for a tokamak

    International Nuclear Information System (INIS)

    Ishii, Shigeyuki

    1983-01-01

    Various methods of discharge cleaning for tokamaks are described. The material of the first walls of tokamaks is usually stainless steel, inconel, titanium and so on. Hydrogen is exclusively used as the discharge gas. Glow discharge cleaning (GDC), Taylor discharge cleaning (TDC), and electron cyclotron resonance discharge cleaning (ECR-DC) are discussed in this paper. The cleaning by GDC is made by moving a movable anode to the center of a tokamak vassel. Taylor found the good cleaning effect of induced discharge by high pressure and low power discharge. This is called TDC. When the frequency of high frequency discharge in a magnetic field is equal to that of the electron cyclotron resonance, the break down potential is lowered if the pressure is sufficiently low. The ECR-CD is made by using this effect. In TDC and ECR-DC, the electron temperature, which has a close relation to the production rate of H 0 , can be controlled by the pressure. In GDC, the operating pressure was improved by the radio frequency glow (RG) method. However, there is still the danger of arcing. In case of GDC and ECR-DC, the position of plasma can be controlled, but not in case of TDC. The TDC is accepted by most of takamak devices in the world. (Kato, T.)