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.

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

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)

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.

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.

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)

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.

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 �

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

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

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.

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.

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

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

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

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)

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.

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.

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)

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.

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)

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)

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.

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)

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

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)

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.

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

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)

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)

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

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.

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

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

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.

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)

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)

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

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.

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

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.

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.)

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.

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

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

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

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.

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

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)

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.

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.

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)

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.

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)

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)

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)

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)

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)

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

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

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)

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.

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)

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.

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

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)...

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

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

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.

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

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.

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.

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.

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

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

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

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)

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

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

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

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

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

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

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)

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

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.

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)

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.

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

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)

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)

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.

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)

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.

AVHRR GAC SST Reanalysis Version 1 (RAN1

Directory of Open Access Journals (Sweden)

Alexander Ignatov

2016-04-01

Full Text Available In response to its users’ needs, the National Oceanic and Atmospheric Administration (NOAA initiated reanalysis (RAN of the Advanced Very High Resolution Radiometer (AVHRR Global Area Coverage (GAC; 4 km sea surface temperature (SST data employing its Advanced Clear Sky Processor for Oceans (ACSPO retrieval system. Initially, AVHRR/3 data from five NOAA and two Metop satellites from 2002 to 2015 have been reprocessed. The derived SSTs have been matched up with two reference SSTs—the quality controlled in situ SSTs from the NOAA in situ Quality Monitor (iQuam and the Canadian Meteorological Centre (CMC L4 SST analysis—and analyzed in the NOAA SST Quality Monitor (SQUAM online system. The corresponding clear-sky ocean brightness temperatures (BT in AVHRR bands 3b, 4 and 5 (centered at 3.7, 11, and 12 µm, respectively have been compared with the Community Radiative Transfer Model simulations in another NOAA online system, Monitoring of Infrared Clear-sky Radiances over Ocean for SST (MICROS. For some AVHRRs, the time series of “AVHRR minus reference” SSTs and “observed minus model” BTs are unstable and inconsistent, with artifacts in the SSTs and BTs strongly correlated. In the official “Reanalysis version 1” (RAN1, data from only five platforms—two midmorning (NOAA-17 and Metop-A and three afternoon (NOAA-16, -18 and -19—were included during the most stable periods of their operations. The stability of the SST time series was further improved using variable regression SST coefficients, similarly to how it was done in the NOAA/NASA Pathfinder version 5.2 (PFV5.2 dataset. For data assimilation applications, especially those blending satellite and in situ SSTs, we recommend bias-correcting the RAN1 SSTs using the newly developed sensor-specific error statistics (SSES, which are reported in the product files. Relative performance of RAN1 and PFV5.2 SSTs is discussed. Work is underway to improve the calibration of AVHRR/3s and

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)

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.

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

[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

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)

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).

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

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)

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

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

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)

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

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.

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

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)

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)

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.)

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.

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.

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.

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

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)

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

Global 1-km Sea Surface Temperature (G1SST)

Data.gov (United States)

National Aeronautics and Space Administration — JPL OurOcean Portal: 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...

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#

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.

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)

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.

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.

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.

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.

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.

Spectroscopic system for impurity measurements in the TJ-1 Tokamak of JEN; Un sistema espectroscopico para medidas de impurezas en el Tokamak TJ-1 de la JEN

Energy Technology Data Exchange (ETDEWEB)

Navas, G; Zurro, B

1982-07-01

we describe a spectroscopic system with spatial resolution capability that has been configured for plasma diagnostic in the TJ-1 Tokamak of JEN. The experimental system, based on a one meter monochromator, has been absolutely calibrated using a tungsten-halogen lamp. The calibration procedures and the absolute spectral sensitivity are presented as well as its dependence with the polarization. A simplified spectroscopic model of the radiation emitted by the intrinsic plasma impurities (C, 0, . . . ) has been developed. A one dimensional model of the temporal evolution of various ionization stages in coronal equilibrium is used to predict the electron temperature and impurity concentration. This model has been applied to experimental data from several Tokamaks. (Author) 23 refs.

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.

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.)

Spectroscopic system for impurity measurements in the TJ-1 Tokamak of JEN

International Nuclear Information System (INIS)

Navas, G.; Zurro, B.

1982-01-01

we describe a spectroscopic system with spatial resolution capability that has been configured for plasma diagnostic in the TJ-1 Tokamak of JEN. The experimental system, based on a one meter monochromator, has been absolutely calibrated using a tungsten-halogen lamp. The calibration procedures and the absolute spectral sensitivity are presented as well as its dependence with the polarization. A simplified spectroscopic model of the radiation emitted by the intrinsic plasma impurities (C, 0, . . . ) has been developed. A one dimensional model of the temporal evolution of various ionization stages in coronal equilibrium is used to predict the electron temperature and impurity concentration. This model has been applied to experimental data from several Tokamaks. (Author) 23 refs

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...

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)

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.

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.

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

Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

International Nuclear Information System (INIS)

Lee, V.D.; Gohar, Y.

1985-01-01

A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of builtup, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

International Nuclear Information System (INIS)

Lee, V.D.; Gohar, Y.

1985-01-01

A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

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

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)

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

Using a 1-D model to reproduce diurnal SST signals

DEFF Research Database (Denmark)

Karagali, Ioanna; Høyer, Jacob L.

2014-01-01

The diurnal variability of SST has been extensively studied as it poses challenges for validating and calibrating satellite sensors, merging SST time series, oceanic and atmospheric modelling. As heat is significantly trapped close to the surface, the diurnal signal’s maximum amplitude is best...... captured by radiometers. The availability of infra-red retrievals from a geostationary orbit allows the hourly monitoring of the diurnal SST evolution. When infra-red SSTs are validated with in situ measurements a general mismatch is found, associated with the different reference depth of each type...... of measurement. A generally preferred approach to bridge the gap between in situ and remotely obtained measurements is through modelling of the upper ocean temperature. This ESA supported study focuses on the implementation of the 1 dimensional General Ocean Turbulence Model (GOTM), in order to resolve...

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.

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)

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.

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.

Influence of the helical resonant fields on the plasma potential in the TBR-1 Tokamak

International Nuclear Information System (INIS)

Ribeiro, C.; Silva, R.P. da; Caldas, I.L.; Fagundes, A.N.; Sanada, E.K.

1990-01-01

This work describes an experimental work that are in progress in TBR-1 tokamak about the influence of resonant helical fields on the plasma potential. TBR-1 is a small tokamak in operation in the Physics Institute of University of Sao Paulo and used for basic research, diagnostic development and personal formation. Its main parameters are: R(Major Radius) = 0.30 m; a v (Vessel Radius) = 0.11 m; a(Plasma Radius) = 0.08 m; R/a(Aspect Ratio) = 3.75; B φ (Toroidal Field) = 5 kG; n e0 (Central Electron Density) ≅ 7 x 10 18 m -3 ; T e0 (central electron temperature) ≅ 200 eV. (Author)

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

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

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

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

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.

Identification of the receptors for somatostatin (SST) and cortistatin (CST) in chickens and investigation of the roles of cSST28, cSST14, and cCST14 in inhibiting cGHRH1-27NH2-induced growth hormone secretion in cultured chicken pituitary cells.

Science.gov (United States)

Meng, Fengyan; Huang, Guian; Gao, Shunyu; Li, Juan; Yan, Zhenxin; Wang, Yajun

2014-03-25

Somatostatin receptors (SSTRs) are proposed to mediate the actions of somatostatin (SST) and its related peptide, cortistatin (CST), in vertebrates. However, the identity, functionality, and tissue expression of these receptors remain largely unknown in most non-mammalian vertebrates including birds. In this study, five SSTRs (named cSSTR1, cSSTR2, cSSTR3, cSSTR4, cSSTR5) were cloned from chicken brain by RT-PCR. Using a pGL3-CRE-luciferase reporter system, we demonstrated that activation of each cSSTR expressed in CHO cells by cSST28, cSST14 and cCST14 treatment could inhibit forskolin-induced luciferase activity of CHO cells, indicating the functional coupling of all cSSTRs to Gi protein(s). Interestingly, cSSTR1-4 expressed in CHO cells could be activated by cSST28, cSST14 and cCST14 with high potencies, suggesting that they may function as the receptors common for these peptides. In contrast, cSSTR5 could be potently activated by cSST28 only, indicating that it is a cSST28-specific receptor. Using RT-PCR, wide expression of cSSTRs was detected in chicken tissues including pituitary. In accordance with their expression in pituitary, cSST28, cSST14, and cCST14 were demonstrated to inhibit basal and novel cGHRH1-27NH2-induced GH secretion in cultured chicken pituitary cells dose-dependently (0-10nM) by Western blot analysis, suggesting the involvement of cSSTR(s) common for these peptides in mediating their inhibitory actions. Collectively, our study establishes a molecular basis to elucidate the roles of SST/CST in birds and provide insights into the roles of SST/CST in vertebrates, such as their conserved actions on pituitary. Copyright © 2014 Elsevier Ireland Ltd. All rights reserved.

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)

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

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

GHRSST Level 2P European Medspiration NAR16 SST:1

Data.gov (United States)

National Aeronautics and Space Administration — The EUMETSAT OSI-SAF NAR SST products are SST fields derived from NOAA/AVHRR data and available over 6 pre-defined zones, 4 times per day. After acquisition and...

GHRSST Level 2P European Medspiration NAR18 SST:1

Data.gov (United States)

National Aeronautics and Space Administration — The EUMETSAT OSI-SAF NAR SST products are SST fields derived from NOAA/AVHRR data and available over 6 pre-defined zones, 4 times per day. After acquisition and...

GHRSST Level 2P European Medspiration NAR17 SST:1

Data.gov (United States)

National Aeronautics and Space Administration — The EUMETSAT OSI-SAF NAR SST products are SST fields derived from NOAA/AVHRR data and available over 6 pre-defined zones, 4 times per day. After acquisition and...

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

International Nuclear Information System (INIS)

Kamleitner, J.

2015-01-01

the mode. Another mode investigated in this thesis is the electron fishbone instability, another m/n = 1/1 internal kink mode excited by resonant interaction with the drift reversal of precessing fast electrons. Significant differences in the observations on various tokamaks show the importance of a more systematic experimental study to advance the qualitative understanding and quantitative description of the instability. The presumed roles of barely trapped, barely passing and other specific regions of the electron distribution function in phase space were expected to be clarified by experiments on TCV using the HXRS as the main diagnostic. However, it proved more difficult than expected to destabilize this mode, and only preliminary though promising results were obtained

Sensitive study of the climatological SST by using ATSR global SST data sets

Science.gov (United States)

Xue, Yong; Lawrence, Sean P.; Llewellyn-Jones, David T.

1995-12-01

Climatological sea surface temperature (SST) is an initial step for global climate processing monitoring. A comparison has been made by using Oberhuber's SST data set and two years monthly averaged SST from ATSR thermal band data to force the OGCM. In the eastern Pacific Ocean, these make only a small difference to model SST. In the western Pacific Ocean, the use of Oberhuber's data set gives higher climatological SST than that using ATSR data. The SSTs were also simulated for 1992 using climatological SSTs from two years monthly averaged ATSR data and Oberhuber data. The forcing with SST from ATSR data was found to give better SST simulation than that from Oberhuber's data. Our study has confirmed that ATSR can provide accurate monthly averaged global SST for global climate processing monitoring.

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.

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

NCEP SST Analysis

Science.gov (United States)

Organization Search Go Search Polar Go MMAB SST Analysis Main page About MMAB Our Mission Our Personnel EMC Branches Global Climate & Weather Modeling Mesoscale Modeling Marine Modeling and Analysis Contact EMC (RTG_SST_HR) analysis For a regional map, click the desired area in the global SST analysis and anomaly maps

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)

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...

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.

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

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

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.

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)

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

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.

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

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)

On the HL-1M tokamak plasma confinement time

International Nuclear Information System (INIS)

Qin Yunwen

2001-01-01

Emphasizing that the tokamak plasma confinement time is the plasma particle or thermal energy loss characteristic time, the relevant physical concept and HL-1M tokamak experimental data analyses are reviewed

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

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

Texas Experimental Tokamak: A plasma research facility. Technical progress report, November 1, 1993--October 31, 1994

Energy Technology Data Exchange (ETDEWEB)

Wootton, A.J.

1994-07-01

The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics in order to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks and in particular to understand the role of turbulence. So that they can continue to study the physics that is most relevant to the fusion program, TEXT completed a significant device upgrade this year. The new capabilities of the device and new and innovative diagnostics were exploited in all main program areas including: (1) configuration studies; (2) electron cyclotron heating physics; (3) improved confinement modes; (4) edge physics/impurity studies; (5) central turbulence and transport; and (6) transient transport. Details of the progress in each of the research areas are described.

Texas Experimental Tokamak: A plasma research facility. Technical progress report, November 1, 1993--October 31, 1994

International Nuclear Information System (INIS)

Wootton, A.J.

1994-07-01

The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics in order to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks and in particular to understand the role of turbulence. So that they can continue to study the physics that is most relevant to the fusion program, TEXT completed a significant device upgrade this year. The new capabilities of the device and new and innovative diagnostics were exploited in all main program areas including: (1) configuration studies; (2) electron cyclotron heating physics; (3) improved confinement modes; (4) edge physics/impurity studies; (5) central turbulence and transport; and (6) transient transport. Details of the progress in each of the research areas are described

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

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

Observation of the m = 1 mode by microwave transmission measurements in the Tore Supra tokamak

International Nuclear Information System (INIS)

Giruzzi, G.; Segui, J.L.; Pecquet, A.L.; Gil, C.

1991-06-01

Microwave transmission measurements in the Tore Supra tokamak exhibit low-frequency oscillations of the transmitted power, associated to the presence of a saturated m = 1, n = 1 mode, as observed by soft X-ray diagnostics. It is shown that these oscillations are related to refraction effects, and specifically to modulations of the electron density profile due to a rotating magnetic island. An analytical solution of the ray equations in the presence of a rotating density perturbation is found, explaining the frequency spectrum of the oscillations

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.

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

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

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

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

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.

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.

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.

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

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

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

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.

GHRSST Level 4 Denmark DMI Global SST:1

Data.gov (United States)

National Aeronautics and Space Administration — The L4 analysis is based upon nighttime GHRSST L2P subskin SST observations from several satellites and instruments such as: AMSRE, ATS_NR_2P, AVHRR18_G,...

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

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)

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

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

Quantifying uncertainty in coral Sr/Ca-based SST estimates from Orbicella faveolata: A basis for multi-colony SST reconstructions

Science.gov (United States)

Richey, J. N.; Flannery, J. A.; Toth, L. T.; Kuffner, I. B.; Poore, R. Z.

2017-12-01

The Sr/Ca in massive corals can be used as a proxy for sea surface temperature (SST) in shallow tropical to sub-tropical regions; however, the relationship between Sr/Ca and SST varies throughout the ocean, between different species of coral, and often between different colonies of the same species. We aimed to quantify the uncertainty associated with the Sr/Ca-SST proxy due to sample handling (e.g., micro-drilling or analytical error), vital effects (e.g., among-colony differences in coral growth), and local-scale variability in microhabitat. We examine the intra- and inter-colony reproducibility of Sr/Ca records extracted from five modern Orbicella faveolata colonies growing in the Dry Tortugas, Florida, USA. The average intra-colony absolute difference (AD) in Sr/Ca of the five colonies during an overlapping interval (1997-2008) was 0.055 ± 0.044 mmol mol-1 (0.96 ºC) and the average inter-colony Sr/Ca AD was 0.039 ± 0.01 mmol mol-1 (0.51 ºC). All available Sr/Ca-SST data pairs from 1997-2008 were combined and regressed against the HadISST1 gridded SST data set (24 ºN and 82 ºW) to produce a calibration equation that could be applied to O. faveolata specimens from throughout the Gulf of Mexico/Caribbean/Atlantic region after accounting for the potential uncertainties in Sr/Ca-derived SSTs. We quantified a combined error term for O. faveolata using the root-sum-square (RMS) of the analytical, intra-, and inter-colony uncertainties and suggest that an overall uncertainty of 0.046 mmol mol-1 (0.81 ºC, 1σ), should be used to interpret Sr/Ca records from O. faveolata specimens of unknown age or origin to reconstruct SST. We also explored how uncertainty is affected by the number of corals used in a reconstruction by iteratively calculating the RMS error for composite coral time-series using two, three, four, and five overlapping coral colonies. Our results indicate that maximum RMS error at the 95% confidence interval on mean annual SST estimates is 1.4 º

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

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.)

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

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.

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.

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...

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#

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

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.)

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.

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

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

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)

Periodic disruptions in the MT-1 tokamak

International Nuclear Information System (INIS)

Zoletnik, S.

1988-11-01

Disruptive instabilities are common phenomena in toroidal devices, especially in tokamaks. Three types can be distinguished: internal, minor and major disruptions. Periodic minor disruptions in the MT-1 tokamak were measured systematically with values of the limiter safety factor between 4 and 10. The density limit as a function of plasma current and horizontal displacement was investigated. Precursor oscillations always appear before the instability with increasing amplitude but can be observed at the density limit with quasi-stationary amplitude. Phase correlation between precursor oscillations were measured with Mirnov coils and x-ray detectors, and they show good agreement with a simple magnetic island model. (R.P.) 11 refs.; 6 figs

GHRSST Level 2P European Medspiration SEVIRI SST:1

Data.gov (United States)

National Aeronautics and Space Administration — SST calculated from the IR channels of MSG at full resolution on a hourly basis. Remapping is made by space averaging, the production of the 3-hourly data is made by...

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)

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)

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.

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.

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.

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)

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.)

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.

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)

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

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

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.

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

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

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.

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.

Preliminary results of MHD stability in HL-1 tokamak

International Nuclear Information System (INIS)

Zheng Yongzhen; Ma Tengcai; Xiao Zhenggui Cai Renfang

1987-01-01

In this paper, MHD activities of HL-1 tokamak plasma are studied with Fourier transform and correlatio analysis. The poloidal modes m = 1, 2, 3,4 and toroidal modes n of MHD magnetic fluctuation signals are detected. Methods for suppressing MHD instabilities are suggested and tested, after MHD instabilities are studied in HL-1. The effects of MHD characteristics in the beginning stage of discharge on the whole process of discharge are analyzed. The disruption, in HL-1 device could be divided into three kinds: internal disruption, minor disruption and major disruption. The result shows that HL-1 will have a better operation condition if internal disruption appears. In is end, the stable operation region of HL-1 tokamak is also given

Soft x-ray measurements on the PLT Tokamak

International Nuclear Information System (INIS)

Von Goeler, S.; Sauthoff, N.; Bitter, M.

1977-10-01

Four experiments are described that currently run on the PLT tokamak and which utilize the soft x-ray emission of the plasma as a diagnostic: the pulse height analysis system for temperature and impurity measurements; the curved crystal Bragg spectrometer for the determination of ionization states of impurities; ''windowless'' surface barrier detectors for the investigation of the ultra soft x-ray radiation in the energy range 0.1 keV < hν < 1 keV and a silicon diode array for x-ray fluctuation measurements. For each diagnostic a short technical description and some recent results obtained with it on PLT are given in order to demonstrate its use

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

GHRSST Level 2P USA NAVOCEANO GOES11 SST:1

Data.gov (United States)

National Aeronautics and Space Administration — SST calculated from the IR channels of GOES-11 at full resolution on a half hourly basis. In raw satellite projection, vertically adjacent pixels averaged and read...

GHRSST Level 2P USA NAVOCEANO GOES12 SST:1

Data.gov (United States)

National Aeronautics and Space Administration — SST calculated from the IR channels of GOES-12 at full resolution on a half hourly basis. In raw satellite projection, vertically adjacent pixels averaged and read...

The Research Progress of the J-TEXT Tokamak

Science.gov (United States)

Zhuang, Ge; Wang, Zhijiang; Ding, Yonghua; Zhang, Ming; Yang, Zhoujun; Gao, Li; Zhang, Xiaoqing; Hu, Xiwei; Pan, Yuan

2010-11-01

In 2004, the TEXT-U tokamak was disassembled and shipped to China, and later on settle down in Huazhong University of Science and Technology. The machine was renamed as the Joint TEXT (J-TEXT) tokamak and obtained its first plasma in 2007. The typical J-TEXT Ohmic discharge was performed in the limiter configuration with the main parameters as follows: major radius R=1.05 m, minor radius a=0.27m, toroidal magnetic field BT=2.2T, plasma current Ip>200kA, line-averaged density ne˜ 2-3 . 1019/m^3, and electron temperature Te0˜ 700eV. Up till now, a few diagnostic systems used to facilitate routine operation and experimental studies were designed and developed. Benefiting from these diagnostic tools, the observation of MHD activities and the statistical analysis of disruption events were done. And measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the LCFS were also made as well. The preliminary results will be presented in detail in the meeting.

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)

Phase 1 RCRA Facility Investigation/Corrective Measures Study Work Plan for Single-Shell Tank (SST) Waste Management Areas

International Nuclear Information System (INIS)

MCCARTHY, M.M.

1999-01-01

This document is the master work plan for the Resource Conservation and Recovery Act of 1976 (RCRA) Corrective Action Program (RCAP) for single-shell tank (SST) farms at the US. Department of Energy's (DOE'S) Hanford Site. The DOE Office of River Protection (ORP) initiated the RCAP to address the impacts of past and potential future tank waste releases to the environment. This work plan defines RCAP activities for the four SST waste management areas (WMAs) at which releases have contaminated groundwater. Recognizing the potential need for future RCAP activities beyond those specified in this master work plan, DOE has designated the currently planned activities as ''Phase 1.'' If a second phase of activities is needed for the WMAs addressed in Phase 1, or if releases are detected at other SST WMAs, this master work plan will be updated accordingly

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

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

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

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

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

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.

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.

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.)

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

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

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)

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

SST algorithms in ACSPO reanalysis of AVHRR GAC data from 2002-2013

Science.gov (United States)

Petrenko, B.; Ignatov, A.; Kihai, Y.; Zhou, X.; Stroup, J.

2014-05-01

In response to a request from the NOAA Coral Reef Watch Program, NOAA SST Team initiated reprocessing of 4 km resolution GAC data from AVHRRs flown onboard NOAA and MetOp satellites. The objective is to create a longterm Level 2 Advanced Clear-Sky Processor for Oceans (ACSPO) SST product, consistent with NOAA operations. ACSPO-Reanalysis (RAN) is used as input in the NOAA geo-polar blended Level 4 SST and potentially other Level 4 SST products. In the first stage of reprocessing (reanalysis 1, or RAN1), data from NOAA-15, -16, -17, -18, -19, and Metop-A and -B, from 2002-present have been processed with ACSPO v2.20, and matched up with quality controlled in situ data from in situ Quality Monitor (iQuam) version 1. The ~12 years time series of matchups were used to develop and explore the SST retrieval algorithms, with emphasis on minimizing spatial biases in retrieved SSTs, close reproduction of the magnitudes of true SST variations, and maximizing temporal, spatial and inter-platform stability of retrieval metrics. Two types of SST algorithms were considered: conventional SST regressions, and recently developed incremental regressions. The conventional equations were adopted in the EUMETSAT OSI-SAF formulation, which, according to our previous analyses, provide relatively small regional biases and well-balanced combination of precision and sensitivity, in its class. Incremental regression equations were specifically elaborated to automatically correct for model minus observation biases, always present when RTM simulations are employed. Improved temporal stability was achieved by recalculation of SST coefficients from matchups on a daily basis, with a +/-45 day window around the current date. This presentation describes the candidate SST algorithms considered for the next round of ACSPO reanalysis, RAN2.

GHRSST Level 4 Denmark DMI North Sea Baltic SST:1

Data.gov (United States)

National Aeronautics and Space Administration — The L4 analysis is based upon nighttime GHRSST L2P subskin SST observations from several satellites and instruments such as: AMSRE, ATS_NR_2P, AVHRR18_G,...

High-beta tokamak research. Annual progress report, 1 August 1982-1 August 1983

International Nuclear Information System (INIS)

Navratil, G.A.

1983-08-01

The main research objectives during the past year fell into four areas: (1) detailed observations over a range of high-beta tokamak equilibria; (2) fabrication of an improved and more flexible high-beta tokamak based on our understanding of the present Torus II; (3) extension of the pulse length to 100 usec with power crowbar operation of the equilibrium field coil sets; and (4) comparison of our equilibrium and stability observations with computational models of MHD equilibrium and stability

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

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.

Electron density measurements in the TRIAM-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

Mitarai, O; Nakashima, H; Nakamura, K; Hiraki, N; Toi, K [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

1980-02-01

Electron density measurements in the TRIAM-1 tokamak are carried out by a 140 GHz microwave interferometer. To follow rapid density variations, a high-speed direct-reading type interferometer is constructed. The density of (1 - 20) x 10/sup 13/ cm/sup -3/ is measured.

Electron density measurements in the TRIAM-1 tokamak

International Nuclear Information System (INIS)

Mitarai, Osamu; Nakashima, Hisatoshi; Nakamura, Kazuo; Hiraki, Naoji; Toi, Kazuo

1980-01-01

Electron density measurements in the TRIAM-1 tokamak are carried out by a 140 GHz microwave interferometer. To follow rapid density variations, a high-speed direct-reading type interferometer is constructed. The density of (1 - 20) x 10 13 cm -3 is measured. (author)

Phase 1 RCRA Facility Investigation & Corrective Measures Study Work Plan for Single Shell Tank (SST) Waste Management Areas

Energy Technology Data Exchange (ETDEWEB)

MCCARTHY, M.M.

1999-08-01

This document is the master work plan for the Resource Conservation and Recovery Act of 1976 (RCRA) Corrective Action Program (RCAP) for single-shell tank (SST) farms at the US. Department of Energy's (DOE'S) Hanford Site. The DOE Office of River Protection (ORP) initiated the RCAP to address the impacts of past and potential future tank waste releases to the environment. This work plan defines RCAP activities for the four SST waste management areas (WMAs) at which releases have contaminated groundwater. Recognizing the potential need for future RCAP activities beyond those specified in this master work plan, DOE has designated the currently planned activities as ''Phase 1.'' If a second phase of activities is needed for the WMAs addressed in Phase 1, or if releases are detected at other SST WMAs, this master work plan will be updated accordingly.

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

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

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.

Runaway electrons in the TRIAM-1 tokamak

International Nuclear Information System (INIS)

Satoh, Takemichi; Nakamura, Kazuo; Toi, Kazuo; Nakamura, Yukio; Hiraki, Naoji

1981-01-01

Pulse height analysis of soft X-rays is carried out in the TRIAM-1 tokamak. The electron temperatures determined from the soft X-ray spectrum agree well with those from Thomson scattering. It is observed that low-energy runaway (slideaway) electrons appear in the high-current-density discharges. (author)

Runaway electrons in the TRIAM-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

Satoh, T; Nakamura, K; Toi, K; Nakamura, Y; Hiraki, N [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

1981-09-01

Pulse height analysis of soft X-rays is carried out in the TRIAM-1 tokamak. The electron temperatures determined from the soft X-ray spectrum agree well with those from Thomson scattering. It is observed that low-energy runaway (slideaway) electrons appear in the high-current-density discharges.

Decadal-scale teleconnection between South Atlantic SST and southeast Australia surface air temperature in austral summer

Science.gov (United States)

Xue, Jiaqing; Li, Jianping; Sun, Cheng; Zhao, Sen; Mao, Jiangyu; Dong, Di; Li, Yanjie; Feng, Juan

2018-04-01

Austral summer (December-February) surface air temperature over southeast Australia (SEA) is found to be remotely influenced by sea surface temperature (SST) in the South Atlantic at decadal time scales. In austral summer, warm SST anomalies in the southwest South Atlantic induce concurrent above-normal surface air temperature over SEA. This decadal-scale teleconnection occurs through the eastward propagating South Atlantic-Australia (SAA) wave train triggered by SST anomalies in the southwest South Atlantic. The excitation of the SAA wave train is verified by forcing experiments based on both linear barotropic and baroclinic models, propagation pathway and spatial scale of the observed SAA wave train are further explained by the Rossby wave ray tracing analysis in non-uniform basic flow. The SAA wave train forced by southwest South Atlantic warming is characterized by an anomalous anticyclone off the eastern coast of the Australia. Temperature diagnostic analyses based on the thermodynamic equation suggest anomalous northerly flows on western flank of this anticyclone can induce low-level warm advection anomaly over SEA, which thus lead to the warming of surface air temperature there. Finally, SST-forced atmospheric general circulation model ensemble experiments also demonstrate that SST forcing in the South Atlantic is associated with the SAA teleconnection wave train in austral summer, this wave train then modulate surface air temperature over SEA on decadal timescales. Hence, observations combined with numerical simulations consistently demonstrate the decadal-scale teleconnection between South Atlantic SST and summertime surface air temperature over SEA.

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)

Improved VIIRS and MODIS SST Imagery

Directory of Open Access Journals (Sweden)

Irina Gladkova

2016-01-01

Full Text Available Moderate Resolution Imaging Spectroradiometers (MODIS and Visible Infrared Imaging Radiometer Suite (VIIRS radiometers, flown onboard Terra/Aqua and Suomi National Polar-orbiting Partnership (S-NPP/Joint Polar Satellite System (JPSS satellites, are capable of providing superior sea surface temperature (SST imagery. However, the swath data of these multi-detector sensors are subject to several artifacts including bow-tie distortions and striping, and require special pre-processing steps. VIIRS additionally does two irreversible data reduction steps onboard: pixel aggregation (to reduce resolution changes across the swath and pixel deletion, which complicate both bow-tie correction and destriping. While destriping was addressed elsewhere, this paper describes an algorithm, adopted in the National Oceanic and Atmospheric Administration (NOAA Advanced Clear-Sky Processor for Oceans (ACSPO SST system, to minimize the bow-tie artifacts in the SST imagery and facilitate application of the pattern recognition algorithms for improved separation of ocean from cloud and mapping fine SST structure, especially in the dynamic, coastal and high-latitude regions of the ocean. The algorithm is based on a computationally fast re-sampling procedure that ensures a continuity of corresponding latitude and longitude arrays. Potentially, Level 1.5 products may be generated to benefit a wide range of MODIS and VIIRS users in land, ocean, cryosphere, and atmosphere remote sensing.

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

Challenges for Plasma Diagnostic in a Next Step Device (FIRE)

International Nuclear Information System (INIS)

Young, Kenneth M.

2002-01-01

The physics program of any next-step tokamak such as FIRE [Fusion Ignition Research Experiment] sets demands for plasma measurement which are at least as comprehensive as on present tokamaks, with the additional capabilities needed for control of the plasma and for understanding the effects of the alpha-particles. The diagnostic instrumentation must be able to provide the fine spatial and temporal resolution required for the advanced tokamak plasma scenarios. It must also be able to overcome the effects of neutron- and gamma-induced electrical noise in ceramic components or detectors, and fluorescence and absorption in optical components. There are practical engineering issues of minimizing radiation streaming while providing essential diagnostic access to the plasma. Many diagnostics will require components at or close to the first wall, e.g., ceramics and MI cable for magnetic diagnostics and mirrors for optical diagnostics; these components must be mounted to operate, and survive, i n fluxes which require special material selection. A better set of diagnostics of alpha-particles than that available for the TFTR [Tokamak Fusion Test Reactor] is essential; it must be qualified well before moving into D-T [deuterim-tritium] experiments. A start has been made to assessing the potential implementation of key diagnostics for the FIRE device. The present status is described

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.

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

Microwave tokamak experiment (MTX) first year of operation and future plans

International Nuclear Information System (INIS)

Jackson, M.C.

1989-01-01

The Microwave Tokamak Experiment (MTX) at Lawrence Livermore National Laboratory (LLNL) began plasma operations in November 1988, and our main goal is the study of electron-cyclotron heating (ECH) in plasma discharges. The MTX tokamak was relocated from the Massachusetts Institute of Technology (MIT), and we have re-created plasma parameters that are similar to those generated while the tokamak was at MIT. After stable ohmic operation was achieved, single-pulse FEL heating experiments began. During this phase, the FEL operated at low power levels on the way to its ultimate goal of 2 GW and 140 GHz with a 30-ns pulse length. We have developed a number of new diagnostics to measure these fast FEL pulses and the resulting plasma effects. In this paper, we present results that show the correlation of MTX data with MIT data, some of the operational modifications and procedures used, results to date from preliminary tokamak operations with the FEL, and our near-term operational plans. 7 refs., 8 figs., 1 tab

Stochasticity and the m = 1 mode in tokamaks

International Nuclear Information System (INIS)

Izzo, R.; Monticello, D.A.; Stodiek, W.; Park, W.

1986-05-01

It has recently been proposed that stochasticity resulting from toroidal coupling could lead to a saturation of the m = 1 internal mode in tokamaks. We present results from the nonlinear evolution of the m = 1 mode with toroidal coupling that show that stochasticity is not enough to cause saturation of the m = 1 mode

TBR-1 (Brazilian Tokamak) - Recent Results

International Nuclear Information System (INIS)

Fagundes, A.N.; Cruz Junior, D.F. da; Galvao, R.M.O.; Elizondo, J.I.; Nascimento, I.C. do; Sa, W.P. de; Sanada, E.K.; Silva, R.P.; Tuszel, A.G.; Vannucci, A.; Vuolo, J.H.

1987-08-01

The TBR-1 is a small Tokamak installed at the Physics Institute of the University of Sao Paulo. The machine was designed in 1977 and begun to be used in plasma scientific research in early 1980. its main characteristics are: Major radius, 0,30m; Minor radius (limiter), 0,08m; Toroidal field, 5 KG; Plasma current, 10KA (typical); Current duration, 6 ms (typical). In this paper we report the results of recent experimental research done in the TBR-1. (author) [pt

Development of high field superconducting Tokamak 'TRIAM-1M'

International Nuclear Information System (INIS)

Ito, Satoshi; Suzuki, Takao; Suzuki, Shohei; Nishi, Masatsugu; Kawasaki, Takahide.

1984-01-01

The tokamak nuclear fusion apparatus ''TRIAM-1M'' which is constructed in the Research Institute for Applied Mechanics, Kyushu University, has a number of distinctive features as compared with other tokamak projects, that is, the toroidal field coils are made of superconductors for the first time in Japan, and the apparatus is small and has strong magnetic field. Hitachi Ltd. designed and has forwarded the manufacture of the TRIAM-1M. In this paper, the total constitution of the apparatus and the design and manufacture of the plasma vacuum vessel, superconducting toroidal coils and others are reported. The objectives of research are the containment of strong field tokamak plasma and the establishment of the law of proportion, the development of turbulent flow heating method, the adoption of mixed wave current driving method and the practical use of Nb 3 Sn superconducting coils. The apparatus is composed of the vacuum vessel containing plasma, toroidal field coils, poloidal field coils, current transformer coils and turbulent flow heating coils for plasma heating, heat insulating vacuum vessel and supporting structures. The evacuating facility, helium liquefying refrigerator and cooling water facility are installed around the main body. (Kako, I.)

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)

GHRSST Level 2P USA NAVOCEANO AVHRR17 Global SST:1

Data.gov (United States)

National Aeronautics and Space Administration — Multi-Channel SST retreivals generated in real-time from NOAA-17 AVHRR Global Area Coverage data. Used operationally in Navy oceanographic analyses and forecasts.

GHRSST Level 2P USA NAVOCEANO AVHRR18 Global SST:1

Data.gov (United States)

National Aeronautics and Space Administration — Multi-Channel SST retreivals generated in real-time from NOAA-18 AVHRR Global Area Coverage data. Used operationally in Navy oceanographic analyses and forecasts.

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

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

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.)

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

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

Identifying role of subtropical southeast Pacific SST anomalies on precipitation dynamics in Central Chile

Science.gov (United States)

Bozkurt, D.; Garreaud, R.

2014-12-01

Central Chile (CC, western South America coasts, 28°S- 38°S) is the heartland of Chile with the highest population and important economic activities. The region is characterized by semiarid Mediterranean climate with a marked precipitation gradient along the coast from north to south, mostly due to the positioning of the South Pacific Subtropical Anticyclone and the midlatitude westerlies belt. Although there are several diagnostic studies that focus on the impact of tropical Pacific SST on CC precipitation variability via atmospheric teleconnections, less attention has been placed on impacts of subtropical southeast (SE) Pacific SST on precipitation. The later region is immediately adjacent to CC and it interferes with the overpassing atmospheric systems. In particular we want to assess the impact of a consistent cooling over the SE Pacific during the last 30 years. This study is being tackled by a combination of observational and reanalysis datasets together with numerical simulations. Observational dataset includes gridded dataset of CRU, TRMM and GPCP. Moreover, Reynolds SST data V2 based on AVHRR infrared satellite SST data is used for analyzing spatial and temporal changes in SST. Current modelling experiment includes a control simulation, used as reference, and sensitivity simulation that involves perturbations to SST over subtropical SE Pacific for a normal year austral winter (2001) season. A number of simulations with different initial conditions have been carried out by employing ICTP-RegCM4. The domain for simulations was centered at 82oW and 32oW with 288x288 grid cells on 20 km spatial resolution. Preliminary results indicate that the response of precipitation in CC to SST anomalies in the subtropical SE Pacific exhibits more or less linear behavior. In the colder SST experiments, drier conditions dominate over CC, which is possibly related with the intensification of South Pacific Subtropical Anticyclone (SPSA) or a reduction in the available

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)

Compact tokamak reactors. Part 1 (analytic results)

International Nuclear Information System (INIS)

Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

1996-01-01

We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. 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 tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper 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 recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

Characteristics of the positive ion source at reduced gas feed

Energy Technology Data Exchange (ETDEWEB)

Sharma, S. K., E-mail: sksharma@ipr.res.in; Bharathi, P.; Prahlad, V.; Patel, P. J.; Choksi, B.; Jana, M. R.; Bansal, L. K.; Qureshi, K.; Sumod, C. B.; Vadher, V.; Thakkar, D.; Gupta, L. N.; Rambabu, S.; Parmar, S.; Contractor, N.; Sahu, A. K.; Pandya, B.; Sridhar, B.; Pandya, S.; Baruah, U. K. [Institute for Plasma Research, Bhat, Gandhinagar (India)

2014-11-15

The neutral beam injector of steady state superconducting tokamak (SST1-NBI) at IPR is designed for injecting upto 1.7 MW of neutral beam (Hº, 30–55 keV) power to the tokamak plasma for heating and current drive. Operations of the positive ion source (PINI or Plug-In-Neutral-Injector) of SST1-NBI were carried out on the NBI test stand. The PINI was operated at reduced gas feed rate of 2–3 Torr l/s, without using the high speed cryo pumps. Experiments were conducted to achieve a stable beam extraction by optimizing operational parameters namely, the arc current (120–300 A), acceleration voltage (16–40 kV), and a suitable control sequence. The beam divergence, power density profiles, and species fractions (H{sup +}:H{sub 2}{sup +}:H{sub 3}{sup +}) were measured by using the diagnostics such as thermal calorimetry, infrared thermography, and Doppler shift spectroscopy. The maximum extracted beam current was about 18 A. A further increase of beam current was found to be limited by the amount of gas feed rate to the ion source.

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)

Dynamical diagnostics of the SST annual cycle in the eastern equatorial Pacific: Part II analysis of CMIP5 simulations

Science.gov (United States)

Chen, Ying-Ying; Jin, Fei-Fei

2017-12-01

In this study, a simple coupled framework established in Part I is utilized to investigate inter-model diversity in simulating the equatorial Pacific SST annual cycle (SSTAC). It demonstrates that the simulated amplitude and phase characteristics of SSTAC in models are controlled by two internal dynamical factors (the damping rate and phase speed) and two external forcing factors (the strength of the annual and semi-annual harmonic forcing). These four diagnostic factors are further condensed into a dynamical response factor and a forcing factor to derive theoretical solutions of amplitude and phase of SSTAC. The theoretical solutions are in remarkable agreement with observations and CMIP5 simulations. The great diversity in the simulated SSTACs is related to the spreads in these dynamic and forcing factors. Most models tend to simulate a weak SSTAC, due to their weak damping rate and annual harmonic forcing. The latter is due to bias in the meridional asymmetry of the annual mean state of the tropical Pacific, represented by the weak cross-equatorial winds in the cold tongue region.

The role of SST variability in the simulation of the MJO

Science.gov (United States)

Stan, Cristiana

2017-12-01

The sensitivity of the Madden-Julian Oscillation to high-frequency variability (period 1-5 days) of sea surface temperature (SST) is investigated using numerical experiments with the super-parameterized Community Climate System Model. The findings of this study emphasize the importance of air-sea interactions in the simulation of the MJO, and stress the necessity of an accurate representation of ocean variability on short time scales. Eliminating 1-5-day variability of surface boundary forcing reduces the intraseasonal variability (ISV) of the tropics during the boreal winter. The ISV spectrum becomes close to the red noise background spectrum. The variability of atmospheric circulation shifts to longer time scales. In the absence of high-frequency variability of SST the MJO power gets confined to wavenumbers 1-2 and the magnitude of westward power associated with Rossby waves increases. The MJO convective activity propagating eastward from the Indian Ocean does not cross the Maritime Continent, and convection in the western Pacific Ocean is locally generated. In the Indian Ocean convection tends to follow the meridional propagation of SST anomalies. The response of the MJO to 1-5-day variability in the SST is through the charging and discharging mechanisms contributing to the atmospheric column moist static energy before and after peak MJO convection. Horizontal advection and surface fluxes show the largest sensitivity to SST perturbations.

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

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...

Spectroscopic diagnostics of high temperature plasmas, January 1, 1991--December 31, 1991

International Nuclear Information System (INIS)

1991-01-01

During 1991, the activities of the John Hopkins University Plasma Spectroscopy Group have covered several areas of research, in the domain of XUV spectroscopy of magnetically confined fusion plasmas. While the main effort concentrated on the development of novel diagnostics which utilize Layered Synthetic Microstructures (LSMs) as the dispersive/filtering elements, work has been done in the area of detector development and the physics of the tokamak edge plasma. An XUV monochromator for the 20--200 Angstrom range, which uses flat LSMs, has been built and is currently operated on the D3-D tokamak at General Atomics in San Diego. A design for a follow-up experiment at D3-D is now in progress. As a preliminary step toward tokamak plasma imaging in the XUV wavelength range using curved LSM coated substrates, a prototype XUV camera was built and operated in our laboratory in image the A1 3 emission at λ-175 Angstrom from a Penning Ionization Discharge plasma. Based on these laboratory results, the design of the XUV camera, which will image plasma in the Phaedrus T tokamak O VI emission (150 Angstrom), has been completed. This instrument is presently under construction. Also a detailed design of a system composed of four LSM based imaging devices for N e (0) and T e (0) fluctuation measurements on TEXT has been completed. The accuracy and the uniformity of the LSM coatings on flat and small curved surfaces used in the above-mentioned instruments have been evaluated in our laboratory using an in-house built calibration facility and at the SURF II synchrotron at the National Institute of Standards and Technology

Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

2009-01-01

Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

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

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

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

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)

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.

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...

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)

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

GHRSST Level 2P USA NAVOCEANO AVHRR METOP-A Global SST:1

Data.gov (United States)

National Aeronautics and Space Administration — Multi-Channel SST retreivals generated in real-time from METOP-A AVHRR Global Area Coverage data. Used operationally in Navy oceanographic analyses and forecasts.

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.

Addendum to 'Half coalescence of the m=1, n=1 magnetic island in tokamaks'

International Nuclear Information System (INIS)

Bussac, M.N.; Pellat, R.

1985-01-01

As an addendum to our previous work concerning the half-coalescence instability of an m=1, n=1 magnetic island in tokamaks, the potential energy is given for an arbitrary shape of the separatrix. (orig.)

SST, NOAA POES AVHRR, GAC, 0.1 degrees, Global, Day and Night

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — NOAA CoastWatch provides sea surface temperature (SST) products derived from NOAA's Polar Operational Environmental Satellites (POES). This data provides global area...

NOAA Optimum Interpolation (OI) SST V2

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The optimum interpolation (OI) sea surface temperature (SST) analysis is produced weekly on a one-degree grid. The analysis uses in situ and satellite SST's plus...

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)

A synchronization system to digitalize TJ-1 Tokamak data

International Nuclear Information System (INIS)

Guasp, J.; Perez-Navarro, A.; Pacios, L.

1983-01-01

At TJ-1 Tokamak signals are stored on a 60-channel magnetic memory. In this report, a system to address those channels and synchronize readout is presented. Digitalized signals are stored in structured files on PDP-11/34 magnetic disks. (author)

EAST-AIA deployment under vacuum: Calibration of laser diagnostic system using computer vision

Energy Technology Data Exchange (ETDEWEB)

Yang, Yang, E-mail: yangyang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, 350 Shushanhu Rd, Hefei, Anhui (China); Song, Yuntao; Cheng, Yong; Feng, Hansheng; Wu, Zhenwei; Li, Yingying; Sun, Yongjun; Zheng, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, 350 Shushanhu Rd, Hefei, Anhui (China); Bruno, Vincent; Eric, Villedieu [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France)

2016-11-15

Highlights: • The first deployment of the EAST articulated inspection arm robot under vacuum is presented. • A computer vision based approach to measure the laser spot displacement is proposed. • An experiment on the real EAST tokamak is performed to validate the proposed measure approach, and the results shows that the measurement accuracy satisfies the requirement. - Abstract: For the operation of EAST tokamak, it is crucial to ensure that all the diagnostic systems are in the good condition in order to reflect the plasma status properly. However, most of the diagnostic systems are mounted inside the tokamak vacuum vessel, which makes them extremely difficult to maintain under high vacuum condition during the tokamak operation. Thanks to a system called EAST articulated inspection arm robot (EAST-AIA), the examination of these in-vessel diagnostic systems can be performed by an embedded camera carried by the robot. In this paper, a computer vision algorithm has been developed to calibrate a laser diagnostic system with the help of a monocular camera at the robot end. In order to estimate the displacement of the laser diagnostic system with respect to the vacuum vessel, several visual markers were attached to the inner wall. This experiment was conducted both on the EAST vacuum vessel mock-up and the real EAST tokamak under vacuum condition. As a result, the accuracy of the displacement measurement was within 3 mm under the current camera resolution, which satisfied the laser diagnostic system calibration.

Synthetic Microwave Imaging Reflectometry diagnostic using 3D FDTD Simulations

Science.gov (United States)

Kruger, Scott; Jenkins, Thomas; Smithe, David; King, Jacob; Nimrod Team Team

2017-10-01

Microwave Imaging Reflectometry (MIR) has become a standard diagnostic for understanding tokamak edge perturbations, including the edge harmonic oscillations in QH mode operation. These long-wavelength perturbations are larger than the normal turbulent fluctuation levels and thus normal analysis of synthetic signals become more difficult. To investigate, we construct a synthetic MIR diagnostic for exploring density fluctuation amplitudes in the tokamak plasma edge by using the three-dimensional, full-wave FDTD code Vorpal. The source microwave beam for the diagnostic is generated and refelected at the cutoff surface that is distorted by 2D density fluctuations in the edge plasma. Synthetic imaging optics at the detector can be used to understand the fluctuation and background density profiles. We apply the diagnostic to understand the fluctuations in edge plasma density during QH-mode activity in the DIII-D tokamak, as modeled by the NIMROD code. This work was funded under DOE Grant Number DE-FC02-08ER54972.

Active stabilization of n=0 and n=1 modes in the TCV tokamak

International Nuclear Information System (INIS)

Hofmann, F.; Tonetti, G.; Ward, D.J.; Gribov, Yu.

1991-04-01

The limits of operation of an elongated tokamak are generally defined by axisymmetric (n=0), free-boundary n=1, and ballooning modes. While it has become common practice to stabilize n=0 modes by a combination of passive and active systems, very few experiments have been done so far to investigate active stabilization of n=1 modes. In this paper, we discuss the possibilities for acitive stabilization of n=0 and n=1 in the TCV tokamak. (author) 2 figs., 2 tabs., 10 refs

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

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

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

On the influence of simulated SST warming on rainfall projections in the Indo-Pacific domain: an AGCM study

Science.gov (United States)

Zhang, Huqiang; Zhao, Y.; Moise, A.; Ye, H.; Colman, R.; Roff, G.; Zhao, M.

2018-02-01

Significant uncertainty exists in regional climate change projections, particularly for rainfall and other hydro-climate variables. In this study, we conduct a series of Atmospheric General Circulation Model (AGCM) experiments with different future sea surface temperature (SST) warming simulated by a range of coupled climate models. They allow us to assess the extent to which uncertainty from current coupled climate model rainfall projections can be attributed to their simulated SST warming. Nine CMIP5 model-simulated global SST warming anomalies have been super-imposed onto the current SSTs simulated by the Australian climate model ACCESS1.3. The ACCESS1.3 SST-forced experiments closely reproduce rainfall means and interannual variations as in its own fully coupled experiments. Although different global SST warming intensities explain well the inter-model difference in global mean precipitation changes, at regional scales the SST influence vary significantly. SST warming explains about 20-25% of the patterns of precipitation changes in each of the four/five models in its rainfall projections over the oceans in the Indo-Pacific domain, but there are also a couple of models in which different SST warming explains little of their precipitation pattern changes. The influence is weaker again for rainfall changes over land. Roughly similar levels of contribution can be attributed to different atmospheric responses to SST warming in these models. The weak SST influence in our study could be due to the experimental setup applied: superimposing different SST warming anomalies onto the same SSTs simulated for current climate by ACCESS1.3 rather than directly using model-simulated past and future SSTs. Similar modelling and analysis from other modelling groups with more carefully designed experiments are needed to tease out uncertainties caused by different SST warming patterns, different SST mean biases and different model physical/dynamical responses to the same underlying

COBE-SST2 Sea Surface Temperature and Ice

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — A new sea surface temperature (SST) analysis on a centennial time scale is presented. The dataset starts in 1850 with monthly 1x1 means and is periodically updated....

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

Scaling law of runaway electrons in the HL-1M tokamak

International Nuclear Information System (INIS)

Zheng Yongzhen

2005-01-01

Runaway confinement time in ohmic and additionally heated tokamak plasmas presents an anomalous behavior in comparison with theoretical predictions based on neoclassical models. A one-dimensional numerical including generation and loss effects for runaway electrons is used to deduce the dependence of runaway energy ε τ on runaway confinement time. The simulation results are presented in the form of a scaling law for ε τ on plasma parameters. The scaling of ε τ and therefore the runaway confinement time and runaway electron diffusivity has been studied in the HL-1M tokamak, by measuring hard X-ray spectra under different experimental conditions. (authors)

Characterization and distribution of GHRH, PACAP, TRH, SST and IGF1 mRNAs in the green iguana.

Science.gov (United States)

Ávila-Mendoza, José; Pérez-Rueda, Ernesto; Urban-Sosa, Valeria; Carranza, Martha; Martínez-Moreno, Carlos G; Luna, Maricela; Arámburo, Carlos

2018-01-01

The somatotropic axis (SA) regulates numerous aspects of vertebrate physiology such as development, growth, and metabolism and has influence on several tissues including neural, immune, reproductive and gastric tract. Growth hormone (GH) is a key component of SA, it is synthesized and released mainly by pituitary somatotrophs, although now it is known that virtually all tissues can express GH, which, in addition to its well-described endocrine roles, also has autocrine/paracrine/intracrine actions. In the pituitary, GH expression is regulated by several hypothalamic neuropeptides including GHRH, PACAP, TRH and SST. GH, in turn, regulates IGF1 synthesis in several target tissues, adding complexity to the system since GH effects can be exerted either directly or mediated by IGF1. In reptiles, little is known about the SA components and their functional interactions. The aim of this work was to characterize the mRNAs of the principal SA components in the green iguana and to develop the tools that allow the study of the structural and functional evolution of this system in reptiles. By employing RT-PCR and RACE, the cDNAs encoding for GHRH, PACAP, TRH, SST and IGF1 were amplified and sequenced. Results showed that these cDNAs coded for the corresponding protein precursors of 154, 170, 243, 113, and 131 amino acids, respectively. Of these, GHRH, PACAP, SST and IGF1 precursors exhibited a high structural conservation with respect to its counterparts in other vertebrates. On the other hand, iguana's TRH precursor showed 7 functional copies of mature TRH (pyr-QHP-NH 2 ), as compared to 4 and 6 copies of TRH in avian and mammalian proTRH sequences, respectively. It was found that in addition to its primary production site (brain for GHRH, PACAP, TRH and SST, and liver for IGF1), they were also expressed in other peripheral tissues, i.e. testes and ovaries expressed all the studied mRNAs, whereas TRH and IGF1 mRNAs were observed ubiquitously in all tissues considered. These

Development of a measurement technique to characterize erosion and redeposition in a tokamak by speckle interferometry

International Nuclear Information System (INIS)

Dore, P.

2006-11-01

This work aims at proving the feasibility of temporal phase shifting speckle interferometry to make erosion/redeposition measurements on plasma facing components in situ on a tokamak. Results show clearly that the interferometric technique can be implemented on a tokamak to provide erosion/redeposition measurements. The optical setup and the interferograms acquisition and processing have been developed and tested in laboratory before being suited to the complex tokamak environment. We finally have an optical technique able to characterize erosion/redeposition mechanisms (amount of eroded/redeposited material, location) on optically rough plasma facing components (carbon fibre composite, tungsten). These components, suffering from random displacements (as vibrations) during acquisition, are relatively large (∼ 50 x 50 cm 2 ) and could be situated far away from the CCD camera (∼ 3 m). Now, we need to define the regions of plasma facing components where we want to make erosion and redeposition measurements. After that, we propose a diagnostic to validate the optical technique in situ on a tokamak, allowing us to develop a diagnostic for ITER. (author)

Experiment of laser thomson scattering at HL-1 tokamak device

International Nuclear Information System (INIS)

Zuo Henian; Chen Jiafu; Yan Derong; Liu Aiping; Shi Peilan; Wang Wei; Liu Xiaomei

1989-05-01

The structure and performance of the Ruby Laser Thomson Scattering apparatus for HL-1 tokamak device is described. The method of acquisition and calibration of multichannel scattered signals are presented. Examples of measured electron temperature T. with experimental error are given

Ion and electron parameters in the alcator C tokamak scrape-off region

International Nuclear Information System (INIS)

Wan, A.S.H.

1986-05-01

Janus is a bi-directional, multi-functional edge probe used to diagnose the ion and electron parameters in the Alcator C tokamak scrape-off region. Two mirror image sets of diagnostics are aligned to face the electron and ion sides along magnetic field lines. Each set of diagnostics consists of a retarding-field energy analyzer (RFEA), a Langmuir probe, and a calorimeter. The RFEA can alternatively sample both the ion and electron parallel energy distribution functions during a tokamak discharge. From the Langmuir probe, one can infer electron temperature, density, and the plasma floating potential. Simple Langmuir probe theory is found to yield the best agreement between the measured Langmuir probe characteristics and the RFEA-inferred T/sub e/. The calorimeter independently detects the total parallel heat flux incident to an electrically floating plate. The measured sheath transmission coefficient, however, is typically lower than the theoretically predicted value by a factor of approx.3. Together these diagnostics enable detailed, localized edge plasma characterization on Alcator C

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

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)

An in situ accelerator-based diagnostic for plasma-material interactions science on magnetic fusion devices.

Science.gov (United States)

Hartwig, Zachary S; Barnard, Harold S; Lanza, Richard C; Sorbom, Brandon N; Stahle, Peter W; Whyte, Dennis G

2013-12-01

This paper presents a novel particle accelerator-based diagnostic that nondestructively measures the evolution of material surface compositions inside magnetic fusion devices. The diagnostic's purpose is to contribute to an integrated understanding of plasma-material interactions in magnetic fusion, which is severely hindered by a dearth of in situ material surface diagnosis. The diagnostic aims to remotely generate isotopic concentration maps on a plasma shot-to-shot timescale that cover a large fraction of the plasma-facing surface inside of a magnetic fusion device without the need for vacuum breaks or physical access to the material surfaces. Our instrument uses a compact (~1 m), high-current (~1 milliamp) radio-frequency quadrupole accelerator to inject 0.9 MeV deuterons into the Alcator C-Mod tokamak at MIT. We control the tokamak magnetic fields--in between plasma shots--to steer the deuterons to material surfaces where the deuterons cause high-Q nuclear reactions with low-Z isotopes ~5 μm into the material. The induced neutrons and gamma rays are measured with scintillation detectors; energy spectra analysis provides quantitative reconstruction of surface compositions. An overview of the diagnostic technique, known as accelerator-based in situ materials surveillance (AIMS), and the first AIMS diagnostic on the Alcator C-Mod tokamak is given. Experimental validation is shown to demonstrate that an optimized deuteron beam is injected into the tokamak, that low-Z isotopes such as deuterium and boron can be quantified on the material surfaces, and that magnetic steering provides access to different measurement locations. The first AIMS analysis, which measures the relative change in deuterium at a single surface location at the end of the Alcator C-Mod FY2012 plasma campaign, is also presented.

High-resolution Thomson scattering system on the COMPASS tokamak: Evaluation of plasma parameters and error analysis

Czech Academy of Sciences Publication Activity Database

Aftanas, Milan; Böhm, Petr; Bílková, Petra; Weinzettl, Vladimír; Zajac, Jaromír; Žáček, František; Stöckel, Jan; Hron, Martin; Pánek, Radomír; Scannell, R.; Walsh, M.

2012-01-01

Roč. 83, č. 10 (2012), 10E350-10E350 ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] R&D Projects: GA ČR GA202/09/1467; GA MŠk 7G10072 Institutional research plan: CEZ:AV0Z20430508 Keywords : error analysis * Monte Carlo methods * plasma density * plasma diagnostics * plasma temperature * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://dx.doi.org/10.1063/1.4743956

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

Control system of neoclassical tearing modes in real time on HL-2A tokamak

Science.gov (United States)

Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

2017-11-01

The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

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)

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.

Soft x-ray measurements in the TRIAM-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

Satoh, T; Toi, K; Nakamura, K; Nakamura, Y; Hiraki, N [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

1981-07-01

Soft X-ray pulse height analysis system has been designed and constructed for measurements of electron distribution function and impurity with high spatial resolution (0.5 cm) and temporal resolution (2 msec) in the TRIAM-1 tokamak. The experimental results about electron temperature, enhancement factor, Z sub(eff) and runaway electrons are presented and discussed.

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.

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

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

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)

Pulse discharge cleaning of the vacuum vessel of HL-1 tokamak

International Nuclear Information System (INIS)

Li Guodong; Zhu Yukun; Xiao Zhenggui; Sun Shouqi; Ze Mingrui

1986-01-01

The HL-1 Tokamak was test-operated on September 21, 1984. During the period of vacuum conditioning, including 60 hours of baking up to 200 deg C and 7 x 10 4 shots of pulse discharge cleaning, the calculated quantities of carbon and oxygen removed are equivalent to 24 and 6 monolayers, respectively. Then, 124 shots of tokamak discharge were performed with low level plasma parameters. The plasma current and pulse length achieved were 60 kA and 85 ms at the toroidal magnetic field of 15 kG. This paper described the techniques used and the effect on discharge characteristics of bakeout and pulse discharge cleaning of the vacuum vessel

M.H.D. activity associated with the q=1 surface in the Tore-Supra tokamak; Activite M.H.D. associee a la surface q=1 dans le tokamak Tore-Supra

Energy Technology Data Exchange (ETDEWEB)

Cristofani, P.

1996-02-12

In order to increase the temperature, density and confinement time of the plasma energy inside tokamak devices, several heating and fuel injection techniques have been used. However, the increase of the energy content of the central part of the plasma leads to instabilities in the confinement magnetic structure which can degrade the confinement properties and the temperature performances. Inside the plasma, the ``q=1`` surface plays an important role in the confinement process. The aim of this thesis is to study the experimental physics related to this surface with the analysis of the ``saw-tooth`` periodical internal relaxations and of the ``snake`` structure. The first chapter gives a general introduction about thermonuclear fusion and a description of the plasma and of its equilibrium. Chapter 2 is devoted to the description of the soft X-ray tomography, the diagnostic technique used in this work. In chapter 3, a theoretical presentation of plasma stability and a comparison with experimental results obtained in the Tore-Supra tokamak are given. The observations of saw-tooth instabilities are presented with the principal theoretical models which are used to explain this phenomenon. The snake density instability localized in the central part of the plasma is described in chapter 4 with an attempt of interpretation. The equation of the size evolution of a magnetic island was modified to test different models which can explain the snake stability. One model is based on the modification of the bootstrap current induced by the presence of the snake, and on the local modification of the current induced by the accumulation of impurities inside the snake. (J.S.). 107 refs.

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

SST Anomaly, NOAA POES AVHRR, Casey and Cornillon Climatology, 0.1 degrees, Global

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — NOAA CoastWatch distributes SST anomaly data using a combination of the POES AVHRR Global Area Coverage data, and data from a climatological database by Casey and...

Design of charge exchange recombination spectroscopy for the joint Texas experimental tokamak

Energy Technology Data Exchange (ETDEWEB)

Chi, Y.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Cheng, Z. F.; Hou, S. Y.; Cheng, C.; Li, Z.; Wang, J. R.; Wang, Z. J. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

2014-11-15

The old diagnostic neutral beam injector first operated at the University of Texas at Austin is ready for rejoining the joint Texas experimental tokamak (J-TEXT). A new set of high voltage power supplies has been equipped and there is no limitation for beam modulation or beam pulse duration henceforth. Based on the spectra of fully striped impurity ions induced by the diagnostic beam the design work for toroidal charge exchange recombination spectroscopy (CXRS) system is presented. The 529 nm carbon VI (n = 8 − 7 transition) line seems to be the best choice for ion temperature and plasma rotation measurements and the considered hardware is listed. The design work of the toroidal CXRS system is guided by essential simulation of expected spectral results under the J-TEXT tokamak operation conditions.

GRASS-SST, Fission Products Gas Release and Fuel Swelling in Steady-State and Transients

International Nuclear Information System (INIS)

Zawadzki, S.

2001-01-01

1 - Description of program or function: GRASS-SST is a comprehensive, mechanistic model for the prediction of fission-gas behaviour in UO 2 -base fuels during steady-state and transient conditions. GRASS-SST treats fission-gas release and fuel swelling on an equal basis and simultaneously treats all major mechanisms that influence fission-gas behaviour. Models are included for intra- and inter-granular fission-gas bubble behaviour as well as a mechanistic description of the role of grain-edge inter-linked porosity on fission-gas release and swelling. GRASS-SST calculations include the effects of gas production from fissioning uranium atoms, bubble nucleation, a realistic equation of state for xenon, lattice bubble diffusivities based on experimental observations, bubble migration, bubble coalescence, re-solution, temperature and temperature gradients, inter-linked porosity, and fission-gas interaction with structural defects (dislocations and grain boundaries) on both the distribution of fission-gas within the fuel and on the amount of fission-gas released from the fuel. GRASS-SST includes the effects of the degree of nonequilibrium in the UO 2 lattice on fission-gas bubble mobility and bubble coalescence and also accounts for the observed formation of grain-surface channels. GRASS-SST also includes mechanistic models for grain-growth/grain boundary sweeping and for the behaviour of fission gas during liquefaction/dissolution and fuel melting conditions. 2 - Method of solution: A system of coupled equations for the evolution of the fission-gas bubble-size distributions in the lattice, on dislocations, on grain faces, and grain edges is derived based on the GRASS-SST models. Given a set of operating conditions, GRASS-SST calculates the bubble radii for the size classes of bubbles under consideration using a realistic equation of state for xenon as well as a generalised capillary relation. 3 - Restrictions on the complexity of the problem: Maxima of : 1 axial section

Diagnostic development for current density profile control at KSTAR

Energy Technology Data Exchange (ETDEWEB)

Ko, J., E-mail: jinseok@nfri.re.kr [National Fusion Research Institute, Daejeon 34133 (Korea, Republic of); University of Science and Technology, Daejeon 34113 (Korea, Republic of); Chung, J. [National Fusion Research Institute, Daejeon 34133 (Korea, Republic of); Messmer, M.C.C. [Department of Applied Physics, Eindhoven University of Technology, Eindhoven (Netherlands)

2016-11-01

Highlights: • The motional Stark effect (MSE) diagnostic installed at KSTAR. • Engineering challenges and solutions on the design and fabrication of the front optics housing and filter modules. • Characterization of the bandpass filters and the responses to polarized light. - Abstract: The current density profile diagnostics are critical for the control of the steady-state burning plasma operations. A multi-channel motional Stark effect (MSE) diagnostic system has been implemented for the measurements of the internal magnetic field structures that constrain the magnetic equilibrium reconstruction to accurately produce the tokamak safety factor and current density profiles for the Korea Superconducting Tokamak Advanced Research (KSTAR). This work presents the design and fabrication of the front optics and the filter modules and the calibration activities for the MSE diagnostic at KSTAR.

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

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.

Investigation of advanced materials for fusion alpha particle diagnostics

Energy Technology Data Exchange (ETDEWEB)

Bonheure, G., E-mail: g.bonheure@fz-juelich.de [Laboratory for Plasma Physics, Association “Euratom-Belgian State”, Royal Military Academy, Avenue de la Renaissance, 30 Kunstherlevinglaan, B-1000 Brussels (Belgium); Van Wassenhove, G. [Laboratory for Plasma Physics, Association “Euratom-Belgian State”, Royal Military Academy, Avenue de la Renaissance, 30 Kunstherlevinglaan, B-1000 Brussels (Belgium); Hult, M.; González de Orduña, R. [Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Strivay, D. [Centre Européen d’Archéométrie, Institut de Physique Nucléaire, Atomique et de Spectroscopie, Université de Liège (Belgium); Vermaercke, P. [SCK-CEN, Boeretang, B-2400 Mol (Belgium); Delvigne, T. [DSI SPRL, 3 rue Mont d’Orcq, Froyennes B-7503 (Belgium); Chene, G.; Delhalle, R. [Centre Européen d’Archéométrie, Institut de Physique Nucléaire, Atomique et de Spectroscopie, Université de Liège (Belgium); Huber, A.; Schweer, B.; Esser, G.; Biel, W.; Neubauer, O. [Forschungszentrum Jülich GmbH, Institut für Plasmaphysik, EURATOM-Assoziation, Trilateral Euregio Cluster, D-52425 Jülich (Germany)

2013-10-15

Highlights: ► We examine the feasibility of alpha particle measurements in ITER. ► We test advanced material detectors borrowed from the GERDA neutrino experiment. ► We compare experimental results on TEXTOR tokamak with our detector response model. ► We investigate the detector response in ITER full power D–T plasmas. ► Advanced materials show good signal to noise ratio and alpha particle selectivity. -- Abstract: Fusion alpha particle diagnostics for ITER remain a challenging task. Standard escaping alpha particle detectors in present tokamaks are not applicable to ITER and techniques suitable for fusion reactor conditions need further research and development [1,2]. The activation technique is widely used for the characterization of high fluence rates inside neutron reactors. Tokamak applications of the neutron activation technique are already well developed [3] whereas measuring escaping ions using this technique is a novel fusion plasma diagnostic development. Despite low alpha particle fluence levels in present tokamaks, promising results using activation technique combined with ultra-low level gamma-ray spectrometry [4] were achieved before in JET [5,6]. In this research work, we use new advanced detector materials. The material properties beneficial for alpha induced activation are (i) moderate neutron cross-sections (ii) ultra-high purity which reduces neutron-induced background activation and (iii) isotopic tailoring which increases the activation yield of the measured activation product. Two samples were obtained from GERDA[7], an experiment aimed at measuring the neutrinoless double beta decay in {sup 76}Ge. These samples, made of highly pure (9 N) germanium highly enriched to 87% in isotope Ge-76, were irradiated in real D–D fusion plasma conditions inside the TEXTOR tokamak. Comparison of the calculated and the experimentally measured activity shows good agreement. Compared to previously investigated high temperature ceramic material [8

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)

The Relationships between Tropical Pacific and Atlantic SST and Northeast Brazil Monthly Precipitation.

Science.gov (United States)

Bertacchi Uvo, Cintia; Repelli, Carlos A.; Zebiak, Stephen E.; Kushnir, Yochanan

1998-04-01

The monthly patterns of northeast Brazil (NEB) precipitation are analyzed in relation to sea surface temperature (SST) in the tropical Pacific and Atlantic Oceans, using singular value decomposition. It is found that the relationships between precipitation and SST in both basins vary considerably throughout the rainy season (February-May). In January, equatorial Pacific SST is weakly correlated with precipitation in small areas of southern NEB, but Atlantic SST shows no significant correlation with regional precipitation. In February, Pacific SST is not well related to precipitation, but south equatorial Atlantic SST is positively correlated with precipitation over the northern Nordeste, the latter most likely reflecting an anomalously early (or late) southward migration of the ITCZ precipitation zone. During March, equatorial Pacific SST is negatively correlated with Nordeste precipitation, but no consistent relationship between precipitation and Atlantic SST is found. Atlantic SST-precipitation correlations for April and May are the strongest found among all months or either ocean. Precipitation in the Nordeste is positively correlated with SST in the south tropical Atlantic and negatively correlated with SST in the north tropical Atlantic. These relationships are strong enough to determine the structure of the seasonal mean SST-precipitation correlations, even though the corresponding patterns for the earlier months of the season are quite different. Pacific SST-precipitation correlations for April and May are similar to those for March. Extreme wet (dry) years for the Nordeste occur when both Pacific and Atlantic SST patterns for April and May occur simultaneously. A separate analysis reinforces previous findings in showing that SST in the tropical Pacific and the northern tropical Atlantic are positively correlated and that tropical Pacific-south Atlantic correlations are negligible.Time-lagged analyses show the potential for forecasting either seasonal mean

Electrical testing and performance evaluation of 1:1 prototype JET ELM control coils

Energy Technology Data Exchange (ETDEWEB)

Kanabar, Deven, E-mail: deven@ipr.res.in; Roy, Swati; Ghate, Mahesh; Raj, Piyush; Kundu, Ananya; Kumar, Nitish; Bhavsar, Dhaval; Panchal, Arun; Pradhan, Subrata, E-mail: pradhan@ipr.res.in

2016-12-15

Highlights: • An appropriate 1:1 prototype Small and Large ELM coils for JET have been successfully manufactured using indigenous technologies. • ELM coils have been tested extensively to verify and validated their performance with respect to operational requirement of JET. • The test results confirm and validate the electrical performance of ELM coils. - Abstract: Magnet Technology Development Division at Institute for Plasma research is engaged in extensive R & D for appropriate technologies towards manufacturing of Edge Localized Mode (ELM) magnets for Large Tokamak such as Joint European Torus (JET) as well as for Steady State Superconducting Tokamak (SST-1). Under this project, manufacturing of 1:1 prototype of Large and Small JET ELM control coils (CC) is completed incorporating indigenously developed manufacturing and insulation technologies. Performance evaluation of both the types of coil has been completed for its current carrying capability and insulation resistance as required by various operational scenarios of JET. Experimental setups, test procedures and measurements for electrical characterization of both type of ELM control coil has been discussed in this paper.

Electrical testing and performance evaluation of 1:1 prototype JET ELM control coils

International Nuclear Information System (INIS)

Kanabar, Deven; Roy, Swati; Ghate, Mahesh; Raj, Piyush; Kundu, Ananya; Kumar, Nitish; Bhavsar, Dhaval; Panchal, Arun; Pradhan, Subrata

2016-01-01

Highlights: • An appropriate 1:1 prototype Small and Large ELM coils for JET have been successfully manufactured using indigenous technologies. • ELM coils have been tested extensively to verify and validated their performance with respect to operational requirement of JET. • The test results confirm and validate the electrical performance of ELM coils. - Abstract: Magnet Technology Development Division at Institute for Plasma research is engaged in extensive R & D for appropriate technologies towards manufacturing of Edge Localized Mode (ELM) magnets for Large Tokamak such as Joint European Torus (JET) as well as for Steady State Superconducting Tokamak (SST-1). Under this project, manufacturing of 1:1 prototype of Large and Small JET ELM control coils (CC) is completed incorporating indigenously developed manufacturing and insulation technologies. Performance evaluation of both the types of coil has been completed for its current carrying capability and insulation resistance as required by various operational scenarios of JET. Experimental setups, test procedures and measurements for electrical characterization of both type of ELM control coil has been discussed in this paper.

MHD precursor to disruption in Iran tokamak 1

International Nuclear Information System (INIS)

Alireza, Hojabri; Fatemeh, Hajakbari; Alireza, Hojabri; Mahmmod, Ghoranneviss; Fatemeh, Hajakbari

2004-01-01

The purpose of this paper is to investigate the major disruptions occurring in low-q(a) discharges in Iran Tokamak 1, and to compare the theoretical and experimental results for the rate of island growth. The study of precursor phase of disruption can be predicted and avoided using suitable control systems. In this paper are described the stability analysis and the observed growth rates indicating that the rotating modes are tearing modes. (authors)

International Workshop on Diagnostics for ITER

CERN Document Server

Gorini, Giuseppe; Sindoni, Elio

1996-01-01

This book of proceedings collects the papers presented at the Workshop on Diagnostics for ITER, held at Villa Monastero, Varenna (Italy), from August 28 to September 1, 1995. The Workshop was organised by the International School of Plasma Physics "Piero Caldirola. " Established in 1971, the ISPP has organised over fifty advanced courses and workshops on topics mainly related to plasma physics. In particular, courses and workshops on plasma diagnostics (previously held in 1975, 1978, 1982, 1986, and 1991) can be considered milestones in the history of this institution. Looking back at the proceedings of the previous meetings in Varenna, one can appreciate the rapid progress in the field of plasma diagnostics over the past 20 years. The 1995 workshop was co-organised by the Istituto di Fisica del Plasma of the National Research Council (CNR). In contrast to previous Varenna meetings on diagnostics, which have covered diagnostics in present-day tokamaks and which have had a substantial tutorial component, the 1...

Results of an integration study of a diagnostics port plug in ITER

International Nuclear Information System (INIS)

Salasca, Sophie; Cantone, Bruno; Grosman, André; Esposito, Basilio; Moro, Fabio; Morocco, Daniele; Villari, Rosaria; Angelone, Maurizio; Rincon, Esther; Hidalgo, Carlos; Nagy, Daniel; Kocsis, Gabor; Varela, Paulo; Porempovics, Gabor; Perrollaz, Guillaume; Patel, Kunal; Krivchenkov, Yuri; Walsh, Michael

2013-01-01

Highlights: ► An extensive study on the integration of diagnostics in a port plug of ITER has been performed. ► It has shown that the diagnostic performances could not be reached if their number was not decreased. ► A design of Diagnostic Shield Modules has been validated through mechanical and thermal analyses. ► These analyses have confirmed that the highest loads are concentrated in the vicinity of the plasma. -- Abstract: Diagnostics in ITER are mandatory to characterize the parameters of plasma and study its interactions with plasma-facing components. Diagnostics components in the vicinity of the plasma are supported by metallic structures called port plugs. At the tokamak mid-plane, these components are installed in port plugs through intermediate structures called drawers. Apart from hosting the diagnostics, the port plugs act as shielding against neutrons and gammas, in order to limit the nuclear loads in crucial components (such as diagnostics and superconducting coils) as well as the dose levels in the controlled zones of the tokamak. The radiation shielding function of the port plugs is ensured through an optimized mixture of heavy metallic materials and water, forming shielding blocks surrounding the diagnostics and called Diagnostic Shield Modules (DSMs). These DSMs constitute the rear part of the drawers (the front part being composed of the Diagnostic First Wall). This paper presents the main results of a study performed in Europe on the integration of a particular diagnostics port plug, the Equatorial Port Plug 1 (EPP1). The paper first provides the results of the EPP1 diagnostics integration analysis. In a second step it focuses on the design of the EPP1 DSMs and summarizes the major results of a thorough set of analyses aiming at studying the DSMs behaviour under different loads, suggesting recommendations to improve their current design

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)

Determination of plasma column transverse section in the TBR-1 tokamak

International Nuclear Information System (INIS)

Conde, M.E.

1986-01-01

The temporal evolution of plasma column transverse section in the TBR-1 tokamak is determined. The experimental melhod is based on the simulation of toroidal current distribution in plasma by a set of toroidal filaments. The currents in these filaments are determined by minimization of square error between the magnetic field produced by filaments and the field measured into the tokamak vacuum vessel. For the measurement of magnetic field, twenty small magnetic coils were constructed and installated in the region protected by current limiters. The plasma column transverse cross section is determined by poloidal field produced by the currents in filaments. The multipole moments of plasma current distribution and the Λ Shafranov parameter were obtained. (M.C.K.) [pt

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.

Plasma behavior with molecular beam injection in the HL-1m tokamak

International Nuclear Information System (INIS)

Yao Lianghua; Tang Nianyi; Cui Zhengying; Xu Deming; Deng Zhongchao; Ding Xuantong; Luo Junlin; Dong Jiafu; Guo Gancheng; Yang Shikun; Cui Chenghe; Xiao Zhenggui; Liu Dequan; Chen Xiaoping; Yan Longwen; Yan Donghai; Wang Enyao; Deng Xiwen

1999-01-01

The authors report effect of the new fueling method of high speed molecular beam injection on Tokamak confinement improvement. The present method is an improvement of conventional gas puffing, with performance comparable to the small pellet injection in HL-1M and also to the slow pellet in ASDEX. The fact that a shallower fueling can lead to similar confinement improvement as a deep one suggests that there may exist a critical position in a Tokamak plasma such that any kind of fueling will have a better confinement as long as it can give rise to density peaking at the critical position

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

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

Radiation (UV-IR) measurements in the TRIAM-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

Nakamura, K; Nakamura, Y; Hiraki, N; Itoh, S [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

1981-07-01

Scaling law is presented that the ratio P sub(R)/P sub(I) (P sub(R): radiation loss, P sub(I): input power) is proportional to the parameter a sup(3/2)/R (a: minor radius, R: major radius). The discussion is presented about the design of the apparatus for measuring the radiation (ultraviolet to infrared) from the TRIAM-1 tokamak plasma.

Diagnostics Plan for the National Compact Stellarator Experiment

International Nuclear Information System (INIS)

D. Johnson; T. Brown; H. Neilson; G. Schilling; H. Takahashi; M. Zarnstorff; M. Cole; E. Lazarus; and M. Fenstermacher

2002-01-01

The National Compact Stellarator Experiment (NCSX) is a stellarator-tokamak hybrid seeking to combine the good confinement, high beta and moderate aspect ratio of the tokamak with the quasi-steady-state operation and good stability properties of the stellarator. A preliminary list of measurement requirements, intended to satisfy the needs of the phased research plan, provides the basis for a full complement of plasma diagnostics. It is important to consider this full set, even at this early stage, to assess the adequacy of the stellarator design for diagnostic port access. The 3-D nature of the plasma is a measurement challenge, as is the necessity for high spatial resolution to assess the quality of magnetic surfaces. Other diagnostic requirements include the need for re-entrant views that penetrate the cryostat, for a convenient e-beam probe for field line mapping, and for a diagnostic neutral beam for active spectroscopy

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

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

Progress on the MSE diagnostic for ITER

International Nuclear Information System (INIS)

Lotte, Ph.; Giannella, R.; Von Hellermann, M.; Kuldkepp, M.; Rachlew, E.; Malaquias, A.; Costley, A.; Walker, C.

2004-01-01

The Motional Stark Effect (MSE) diagnostic is now considered as an essential diagnostic for an accurate determination of current profiles in tokamak discharges. It mainly allows a measurement of the direction of the total magnetic field, a very powerful constraint for the determination of the safety factor profile. The realisation of such a diagnostic on ITER implies to face new challenges, because of the bigger size of the machine and of its hard environment. Now, most of the foreseen difficulties have been examined, solutions envisaged, and we propose to review them in this paper. This article is divided into 3 parts: 1) principle of the MSE diagnostic and its feasibility at higher Lorentz electric fields, 2) spatial and time resolution of the diagnostic, and 3) the light collection system

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

The accuracy of SST retrievals from AATSR: An initial assessment through geophysical validation against in situ radiometers, buoys and other SST data sets

Science.gov (United States)

Corlett, G. K.; Barton, I. J.; Donlon, C. J.; Edwards, M. C.; Good, S. A.; Horrocks, L. A.; Llewellyn-Jones, D. T.; Merchant, C. J.; Minnett, P. J.; Nightingale, T. J.; Noyes, E. J.; O'Carroll, A. G.; Remedios, J. J.; Robinson, I. S.; Saunders, R. W.; Watts, J. G.

The Advanced Along-Track Scanning Radiometer (AATSR) was launched on Envisat in March 2002. The AATSR instrument is designed to retrieve precise and accurate global sea surface temperature (SST) that, combined with the large data set collected from its predecessors, ATSR and ATSR-2, will provide a long term record of SST data that is greater than 15 years. This record can be used for independent monitoring and detection of climate change. The AATSR validation programme has successfully completed its initial phase. The programme involves validation of the AATSR derived SST values using in situ radiometers, in situ buoys and global SST fields from other data sets. The results of the initial programme presented here will demonstrate that the AATSR instrument is currently close to meeting its scientific objectives of determining global SST to an accuracy of 0.3 K (one sigma). For night time data, the analysis gives a warm bias of between +0.04 K (0.28 K) for buoys to +0.06 K (0.20 K) for radiometers, with slightly higher errors observed for day time data, showing warm biases of between +0.02 (0.39 K) for buoys to +0.11 K (0.33 K) for radiometers. They show that the ATSR series of instruments continues to be the world leader in delivering accurate space-based observations of SST, which is a key climate parameter.

Half- coalescence of the m/n = 1 magnetic island in Tokamaks

International Nuclear Information System (INIS)

Bussac, M.N.; Pellat, R.

1986-01-01

We show that a configuration containing an m/n = 1 magnetic island is unstable to an ideal MHD mode. The expected nonlinear implications of this instability could explain the disruptive phase of the classical sawtooth behaviour of Tokamak plasmas

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

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

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

Time-resolving electron temperature diagnostic for ALCATOR C

International Nuclear Information System (INIS)

Fairfax, S.A.

1984-05-01

A diagnostic that provides time-resolved central electron temperatures has been designed, built, and tested on the ALCATOR C Tokamak. The diagnostic uses an array of fixed-wavelength x-ray crystal monochromators to sample the x-ray continuum and determine the absolute electron temperature. The resolution and central energy of each channel were chosen to exclude any contributions from impurity line radiation. This document describes the need for such a diagnostic, the design methodology, and the results with typical ALCATOR C plasmas. Sawtooth (m = 1) temperature oscillations were observed after pellet fueling of the plasma. This is the first time that such oscillations have been observed with an x-ray temperature diagnostic

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

Development of a remotely maintainable radio-frequency module for the Compact Ignition Tokamak

International Nuclear Information System (INIS)

Snider, J.D.

1988-01-01

The Compact Ignition Tokamak (CIT) will require reliable remote handling (RH) systems to overcome failures in diagnostic and operational equipment. Oak Ridge National laboratory (ORNL) is responsible for the ex-vessel remote maintenance systems for the CIT. Part of this effort is performing remote maintenance demonstrations on replicas of various CIT equipment. To ensure successful RH, the machine must be designed with proven remote maintenance features. In the demonstrations, critical remote maintenance features are tested before actual CIT equipment designs are finalized. Designs and procedures required to remotely remove and install a radio-frequency (rf) module from a modplane port on the tokamak were recently demonstrated at ORNL. This testing identified both successful design features for remote maintenance of the rf module and areas that require further development. 1 ref., 11 figs

Energy balance in TM-1-MH Tokamak (ohmical heating)

Science.gov (United States)

Stoeckel, J.; Koerbel, S.; Kryska, L.; Kopecky, V.; Dadalec, V.; Datlov, J.; Jakubka, K.; Magula, P.; Zacek, F.; Pereverzev, G. V.

1981-10-01

Plasma in the TM-1-MH Tokamak was experimentally studied in the parameter range: tor. mg. field B = 1,3 T, plasma current I sub p = 14 kA, electron density N sub E 3.10 to the 19th power cubic meters. The two numerical codes are available for the comparison with experimental data. TOKATA-code solves simplified energy balance equations for electron and ion components. TOKSAS-code solves the detailed energy balance of the ion component.

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)

The TaSST: Tactile Sleeve for Social Touch

NARCIS (Netherlands)

Huisman, G.; Darriba Frederiks, A.; van Dijk, B.; Heylen, D.; Kröse, B.

2013-01-01

In this paper we outline the design process of the TaSST (Tactile Sleeve for Social Touch), a touch-sensitive vibrotactile arm sleeve. The TaSST was designed to enable two people to communicate different types of touch over a distance. The touch-sensitive surface of the sleeve consists of a grid of

The TaSST: Tactile sleeve for social touch

NARCIS (Netherlands)

Huisman, Gijs; Darriba Frederiks, Aduén; van Dijk, Elisabeth M.A.G.; Heylen, Dirk K.J.; Krose, Ben

In this paper we outline the design process of the TaSST (Tactile Sleeve for Social Touch), a touch-sensitive vibrotactile arm sleeve. The TaSST was designed to enable two people to communicate different types of touch over a distance. The touch-sensitive surface of the sleeve consists of a grid of

The TaSST - Tactile Sleeve for Social Touch

NARCIS (Netherlands)

Huisman, Gijs; Darriba Frederiks, Aduén; Van Dijk, Betsy; Heylen, Dirk

2013-01-01

In this paper we outline the design process of TaSST (Tactile Sleeve for Social Touch), a touch-sensitive vibrotactile arm sleeve. The TaSST was designed to enable two people to communicate different types of touches over a distance. The touch-sensitive surface of the sleeve consists of a grid of

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)

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)

Report on the high magnetic field tokamak TRIAM-1

Energy Technology Data Exchange (ETDEWEB)

Ito, T; Kawai, Y; Toi, K; Hiraki, N; Nakamure, K [Kyushu Univ., Fukuoke (Japan). Research Inst. for Applied Mechanics

1981-02-01

A high magnetic field tokamak has been constructed at Kyushu University to study the confinement of high magnetic field tokamak plasma and turbulent heating. The tokamak device consists of toroidal field coils, vertical field coils, horizontal field coils, primary windings, a transformer iron core, turbulent heating coils, and a vacuum chamber. For the observation of plasma, plasma monitors, a micro-wave interferometer, a laser scattering system, a neutral particle energy analyzer, a soft X-ray detector, and a visible spectrometer were installed on the vacuum chamber. The experimental results showed that the central electron temperature was about 640 eV, the central ion temperature 280 eV and mean electron density 2.2 x 10/sup 14//cm/sup 3/. It was found that the proportionality law of electron density and confinement time was valid for this small plasma system. By the turbulent heating, the central ion temperature increased from 170 eV to 580 eV.

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

Novel sst2-selective somatostatin agonists. Three-dimensional consensus structure by NMR

Science.gov (United States)

Grace, Christy Rani R.; Erchegyi, Judit; Koerber, Steven C.; Reubi, Jean Claude; Rivier, Jean; Riek, Roland

2008-01-01

The three-dimensional NMR structures of six octapeptide agonist analogues of somatostatin (SRIF) in the free form are described. These analogues, with the basic sequence H-DPhe/Phe2-c[Cys3-Xxx7-DTrp8-Lys9-Thr10-Cys14]-Thr-NH2 (the numbering refers to the position in native SRIF), with Xxx7 being Ala/Aph, exhibit potent and highly selective binding to human SRIF type 2 (sst2) receptors. The backbone of these sst2-selective analogues have the usual type-II’ β-turn reported in the literature for sst2/3/5-subtype-selective analogues. Correlating biological results and NMR studies led to the identification of the side chains of DPhe2, DTrp8 and Lys9 as the necessary components of the sst2 pharmacophore. This is the first study to show that the aromatic ring at position 7 (Phe7) is not critical for sst2 binding and that it plays an important role in sst3 and sst5 binding. This pharmacophore is therefore different from that proposed by others for sst2/3/5 analogues. PMID:16854054

Li-BES detection system for plasma turbulence measurements on the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Berta, Miklós; Anda, G.; Bencze, A.; Dunai, D.; Háček, Pavel; Hron, Martin; Kovácsik, A.; Krbec, Jaroslav; Pánek, Radomír; Réfy, D.; Véres, G.; Weinzettl, Vladimír; Zoletnik, S.

96-97, October (2015), s. 795-798 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : BES * plasma diagnostics * COMPASS tokamak * density fluctuations * plasma density profile Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://www.sciencedirect.com/science/article/pii/S0920379615300442

SST and OLR relationship during Indian summer monsoon: a coupled climate modelling perspective

Science.gov (United States)

Chaudhari, Hemantkumar S.; Hazra, Anupam; Pokhrel, Samir; Chakrabarty, Chandrima; Saha, Subodh Kumar; Sreenivas, P.

2018-04-01

The study mainly investigates sea surface temperature (SST) and outgoing longwave radiation (OLR) relationships in coupled climate model. To support the analysis, high-level cloud and OLR relationship is also investigated. High-level cloud and OLR relationship depicts significant negative correlation over the entire monsoon regime. Coupled climate model is able to produce the same. SST and OLR relationship in observation also depicts significant negative relationship, in particular, over the Equatorial Eastern Indian Ocean (EIO) region. Climate Forecast System version 2 (CFSv2) is able to portray the negative relationship over EIO region; however, it is underestimated as compared to observation. Significant negative correlations elucidate that local SSTs regulate the convection and further it initiates Bjerknes feedback in the central Indian Ocean. It connotes that SST anomalies during monsoon period tend to be determined by oceanic forcing. The heat content of the coastal Bay of Bengal shows highest response to EIO SST by a lag of 1 month. It suggests that the coastal region of the Bay of Bengal is marked by coastally trapped Kelvin waves, which might have come from EIO at a time lag of 1 month. Sea surface height anomalies, depth at 20 °C isotherms and depth at 26 isotherms also supports the above hypothesis. Composite analysis based on EIO index and coupled climate model sensitivity experiments also suggest that the coastal Bay of Bengal region is marked by coastally trapped Kelvin waves, which are propagated from EIO at a time lag of 1 month. Thus, SST and OLR relationship pinpoints that the Bay of Bengal OLR (convection) is governed by local ocean-atmospheric coupling, which is influenced by the delayed response from EIO brought forward through oceanic planetary waves at a lag of 1 month. These results have utmost predictive value for seasonal and extended range forecasting. Thus, OLR and SST relationship can constitute a pivotal role in investigating the

DIII-D Thomson Scattering Diagnostic Data Acquisition, Processing and Analysis Software

International Nuclear Information System (INIS)

Middaugh, K.R.; Bray, B.D.; Hsieh, C.L.; McHarg, B.B.Jr.; Penaflor, B.G.

1999-01-01

One of the diagnostic systems critical to the success of the DIII-D tokamak experiment is the Thomson scattering diagnostic. This diagnostic is unique in that it measures local electron temperature and density: (1) at multiple locations within the tokamak plasma; and (2) at different times throughout the plasma duration. Thomson ''raw'' data are digitized signals of scattered light, measured at different times and locations, from the laser beam paths fired into the plasma. Real-time acquisition of this data is performed by specialized hardware. Once obtained, the raw data are processed into meaningful temperature and density values which can be analyzed for measurement quality. This paper will provide an overview of the entire Thomson scattering diagnostic software and will focus on the data acquisition, processing, and analysis software implementation. The software falls into three general categories: (1) Set-up and Control: Initializes and controls all Thomson hardware and software, synchronizes with other DIII-D computers, and invokes other Thomson software as appropriate. (2) Data Acquisition and Processing: Obtains raw measured data from memory and processes it into temperature and density values. (3) Analysis: Provides a graphical user interface in which to perform analysis and sophisticated plotting of analysis parameters

q=1 advanced tokamak experiments in JET and comparison with ASDEX Upgrade

International Nuclear Information System (INIS)

Joffrin, E.; Wolf, R.; Alper, B.

2002-01-01

The ASDEX Upgrade advanced tokamak scenario with central q close to 1 has been reproduced on JET. For almost identical q profiles, the comparative analysis does show similar features like the fishbone activity and the current profile evolution. In JET, transport analyses indicates that an internal transport barrier (ITB) has been produced. Gradient length criterions based on the ion temperature gradient turbulence stabilization are used to characterize the ITBs in both devices. The trigger of ITBs is associated with rational surfaces in both devices although the underlying physics for this triggering seems different. This experiment has the prospect to get closer to identity experiments between the two tokamaks. (author)

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.

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.

Novel diagnostics for dust in space, Laboratory and fusion plasmas

International Nuclear Information System (INIS)

Castaldo, C.

2011-01-01

In situ diagnostics for mobile dust, based on dust impact ionization phenomena, as well as silica aerogel dust collectors are discussed for applications to space and fusion plasmas. The feasibility of an electro-optical probe to detect hypervelocity (>1 km/s) dust particles in tokamaks is evaluated. For quiescent plasmas, a diagnostic of submicron dust based on measurements of plasma fluctuation spectra can be used (copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

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

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.

A Tropical View of Atlantic Multidecadal SST Variability over the Last Two Millennia

Science.gov (United States)

Wurtzel, J. B.; Black, D. E.; Thunell, R.; Peterson, L. C.; Tappa, E. J.; Rahman, S.

2011-12-01

Instrumental and proxy-reconstructions show the existence of a 60-80 year periodicity in Atlantic sea surface temperature (SST), known as the Atlantic Multidecadal Oscillation (AMO). The AMO is correlated with circum-tropical Atlantic climate phenomena such as Sahel and Nordeste rainfall, as well as Atlantic hurricane patterns. Though it has been suggested that the AMO is controlled by thermohaline circulation, much debate exists as to whether the SST fluctuations are a result of anthropogenic forcing or natural climate variability. Our ability to address this issue has been limited by instrumental SST records that rarely extend back more than 50-100 years and proxy reconstructions that are largely terrestrial-based. Here we present a high-resolution marine sediment-derived reconstruction of seasonal tropical Atlantic SSTs from the Cariaco Basin spanning the past two millennia that is correlated with instrumental SSTs and the AMO for the period of overlap. The full record demonstrates that seasonality is largely controlled by variations in winter/spring SST. Wavelet analysis of the proxy data suggest that variability in the 60-80 year band evolved 250 years ago, while 40-60 year periodicities dominate earlier parts of the record. At least over the last millennia, multidecadal- and centennial- scale SST variability in the tropical Atlantic appears related to Atlantic meridional overturning circulation (AMOC) fluctuations and its associated northward heat transport that in turn may be driven by solar variability. An inverse correlation between the tropical proxy annual average SST record and Δ14C indicates that the tropics experienced positive SST anomalies during times of reduced solar activity, possibly as a result of decreased AMOC strength (Figure 1).

Development of in situ cleaning techniques for diagnostic mirrors in ITER

International Nuclear Information System (INIS)

Litnovsky, A.; Laengner, M.; Matveeva, M.; Schulz, Ch.; Marot, L.; Voitsenya, V.S.; Philipps, V.; Biel, W.; Samm, U.

2011-01-01

Mirrors will be used in all optical and laser-based diagnostic systems of ITER. In the severe environment, the optical characteristics of mirrors will be degraded, hampering the entire performance of the respective diagnostics. A minute impurity deposition of 20 nm of carbon on the mirror is sufficient to decrease the mirror reflectivity by tens of percent outlining the necessity of the mirror cleaning in ITER. The results of R and D on plasma cleaning of molybdenum diagnostic mirrors are reported. The mirrors contaminated with amorphous carbon films in the laboratory conditions and in the tokamaks were cleaned in steady-state hydrogenic plasmas. The maximum cleaning efficiency of 4.2 nm/min was reached for the laboratory and soft tokamak hydrocarbon films, whereas for the hard tokamak films the carbidization of mirrors drastically decreased the cleaning efficiency down to 0.016 nm/min. This implies the necessity of sputtering cleaning of contaminated mirrors as the only reliable tool to remove the deposits by plasma cleaning. An overview of R and D program on mirror cleaning is provided along with plans for further studies and the recommendations for ITER mirror-based diagnostics.

Daily MUR SST, Interim near-real-time (nrt) product

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 (MUR, or Multi-scale ultra-high resolution Temperature) by the JPL sciengists Drs. Mike...

Impacts of SST Patterns on Rapid Intensification of Typhoon Megi (2010)

Science.gov (United States)

Kanada, Sachie; Tsujino, Satoki; Aiki, Hidenori; Yoshioka, Mayumi K.; Miyazawa, Yasumasa; Tsuboki, Kazuhisa; Takayabu, Izuru

2017-12-01

Typhoon Megi (2010), a very intense tropical cyclone with a minimum central pressure of 885 hPa, was characterized by especially rapid intensification. We investigated this intensification process by a simulation experiment using a high-resolution (0.02° × 0.02°) three-dimensional atmosphere-ocean coupled regional model. We also performed a sensitivity experiment with a time-fixed sea surface temperature (SST). The coupled model successfully simulated the minimum central pressure of Typhoon Megi, whereas the fixed SST experiment simulated an excessively low minimum central pressure of 839 hPa. The simulation results also showed a close relationship between the radial SST profiles and the rapid intensification process. Because the warm sea increased near-surface water vapor and hence the convective available potential energy, the high SST in the eye region facilitated tall and intense updrafts inside the radius of maximum wind speed and led to the start of rapid intensification. In contrast, high SST outside this radius induced local secondary updrafts that inhibited rapid intensification even if the mean SST in the core region exceeded 29.0°C. These secondary updrafts moved inward and eventually merged with the primary eyewall updrafts. Then the storm intensified rapidly when the high SST appeared in the eye region. Thus, the changes in the local SST pattern around the storm center strongly affected the rapid intensification process by modulating the radial structure of core convection. Our results also show that the use of a high-resolution three-dimensional atmosphere-ocean coupled model offers promise for improving intensity forecasts of tropical cyclones.

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.

Genome-Scale Analysis Reveals Sst2 as the Principal Regulator of Mating Pheromone Signaling in the Yeast Saccharomyces cerevisiae†

Science.gov (United States)

Chasse, Scott A.; Flanary, Paul; Parnell, Stephen C.; Hao, Nan; Cha, Jiyoung Y.; Siderovski, David P.; Dohlman, Henrik G.

2006-01-01

A common property of G protein-coupled receptors is that they become less responsive with prolonged stimulation. Regulators of G protein signaling (RGS proteins) are well known to accelerate G protein GTPase activity and do so by stabilizing the transition state conformation of the G protein α subunit. In the yeast Saccharomyces cerevisiae there are four RGS-homologous proteins (Sst2, Rgs2, Rax1, and Mdm1) and two Gα proteins (Gpa1 and Gpa2). We show that Sst2 is the only RGS protein that binds selectively to the transition state conformation of Gpa1. The other RGS proteins also bind Gpa1 and modulate pheromone signaling, but to a lesser extent and in a manner clearly distinct from Sst2. To identify other candidate pathway regulators, we compared pheromone responses in 4,349 gene deletion mutants representing nearly all nonessential genes in yeast. A number of mutants produced an increase (sst2, bar1, asc1, and ygl024w) or decrease (cla4) in pheromone sensitivity or resulted in pheromone-independent signaling (sst2, pbs2, gas1, and ygl024w). These findings suggest that Sst2 is the principal regulator of Gpa1-mediated signaling in vivo but that other proteins also contribute in distinct ways to pathway regulation. PMID:16467474