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

International Nuclear Information System (INIS)

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

2006-01-01

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

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)

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

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.

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.

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)

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 �

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

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)

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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.

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)

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

Science.gov (United States)

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

2017-12-01

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

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

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

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.

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.

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)

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)

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

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.

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.

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

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.

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.

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)

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

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)

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

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

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.

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.

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

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)

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.

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

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.

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.

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)

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

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 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

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

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.

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)

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)...

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.

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.

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

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)

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)

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

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)

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

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

Overview of data acquisition system for SST-1 diagnostics

International Nuclear Information System (INIS)

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

2016-01-01

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

Overview of data acquisition system for SST-1 diagnostics

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

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.

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.

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.

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)

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.)

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

Steady State Advanced Tokamak (SSAT): The mission and the machine

International Nuclear Information System (INIS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

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

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.

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)

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.

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)

Realizing steady-state tokamak operation for fusion energy

International Nuclear Information System (INIS)

Luce, T. C.

2011-01-01

Continuous operation of a tokamak for fusion energy has clear engineering advantages but requires conditions beyond those sufficient for a burning plasma. The fusion reactions and external sources must support both the pressure and the current equilibrium without inductive current drive, leading to demands on stability, confinement, current drive, and plasma-wall interactions that exceed those for pulsed tokamaks. These conditions have been met individually, and significant progress has been made in the past decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are operated routinely without disruptions near pressure limits, as needed for steady-state operation. Fully noninductive sustainment with more than half of the current from intrinsic currents has been obtained for a resistive time with normalized pressure and confinement approaching those needed for steady-state conditions. One remaining challenge is handling the heat and particle fluxes expected in a steady-state tokamak without compromising the core plasma performance.

Accessibility of high β tokamak states

International Nuclear Information System (INIS)

Hogan, J.T.

1978-05-01

Encouraging results with neutral beam heating and adiabatic compression of tokamak plasmas have prompted new experiments which will study the approach to high β states. As projected tokamak β values become nonnegligible (average β of 4% is the goal), the models previously used for transport calculations will become inadequate. These models will be required to account for the evolution of the magnetic geometry, along with the change in plasma parameters. We present an axisymmetric transport model which should be useful for studying the approach to higher β values in tokamak experiments. Results from transport calculations with this model allow us to draw a parallel between observed behavior in seemingly unrelated experiments: electron heating by neutral injection in the ORMAK device and adiabatic compression in the ATC experiment. Finally, we find that the nature of cross-field transport may be expected to change as significant β values are reached. Enhanced transport from ballooning instabilities is likely to play a role as important as that now played by sawtooth (m = 1) and saturated (m = 2) instabilities. New techniques for describing this transport are required

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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)

Relaxed states of tokamak plasmas

International Nuclear Information System (INIS)

Kucinski, M.Y.; Okano, V.

1993-01-01

The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)

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)

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...

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.

Technology and physics in the Tokamak Program: The need for an integrated, steady-state RandD tokamak experiment

International Nuclear Information System (INIS)

1988-05-01

The Steady-state Tokamak (STE) Experiment is a proposed superconducting-coil, hydrogen-plasma tokamak device intended to address the integrated non-nuclear issues of steady state, high-power tokamak physics and technology. Such a facility has been called for in the US program plan for the mid 1990's, and will play a unique role in the world-wide fusion effort. Information from STE on steady-state current drive, plasma control, and high power technology will contribute significantly to the operating capabilities of future steady-state devices. This paper reviews preliminary designs and expected technological contributions to the US and world fusion reactor research from each of the above mentioned reactor systems. This document is intended as a proposal and feasibility discussion and does not include exhaustive technical reviews. 12 figs., 3 tabs

A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

Directory of Open Access Journals (Sweden)

Mitsuru Kikuchi

2010-11-01

Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

Impact of the Gulf of California SST on simulating precipitation and crop productivity in the Southwestern United States

Science.gov (United States)

Kim, S.; Kim, J.; Prasad, A. K.; Stack, D. H.; El-Askary, H. M.; Kafatos, M.

2012-12-01

Like other ecosystems, agricultural productivity is substantially affected by climate factors. Therefore, accurate climatic data (i.e. precipitation, temperature, and radiation) is crucial to simulating crop yields. In order to understand and anticipate climate change and its impacts on agricultural productivity in the Southwestern United States, the WRF regional climate model (RCM) and the Agricultural Production Systems sIMulator (APSIM) were employed for simulating crop production. 19 years of WRF RCM output show that there is a strong dry bias during the warm season, especially in Arizona. Consequently, the APSIM crop model indicates very low crop yields in this region. We suspect that the coarse resolution of reanalysis data could not resolve the relatively warm Sea Surface Temperature (SST) in the Gulf of California (GC), causing the SST to be up to 10 degrees lower than the climatology. In the Southwestern United States, a significant amount of precipitation is associated with North American Monsoon (NAM). During the monsoon season, the low-level moisture is advected to the Southwestern United States via the GC, which is known to be the dominant moisture source. Thus, high-resolution SST data in the GC is required for RCM simulations to accurately represent a reasonable amount of precipitation in the region, allowing reliable evaluation of the impacts on regional ecosystems.and evaluate impacts on regional ecosystems. To evaluate the influence of SST on agriculture in the Southwestern U.S., two sets of numerical simulations were constructed: a control, using unresolved SST of GC, and daily updated SST data from the MODIS satellite sensor. The meteorological drivers from each of the 6 year RCM runs were provided as input to the APSIM model to determine the crop yield. Analyses of the simulated crop production, and the interannual variation of the meteorological drivers, demonstrate the influence of SST on crop yields in the Southwestern United States.

System studies for quasi-steady-state advanced physics tokamak

International Nuclear Information System (INIS)

Reid, R.L.; Peng, Y.K.M.

1983-11-01

Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated

The steady-state tokamak program

International Nuclear Information System (INIS)

Politzer, D.A.; Nevins, W.M.

1992-01-01

This paper reports on a steady-state tokamak experiment (STE) needed to develop the technology and physics data base required for construction of a steady-state fusion power demonstration reactor in the early 21st century. The STE will provide an integrated facility for the development and demonstration of steady-state and particle handling, low-activation high-heat-flux components and materials, efficient current drive, and continuous plasma performance in steady-state, with reactor-like plasma conditions under severe conditions of heat and particle bombardment of the wall. The STE facility will also be used to develop operation and control scenarios for ITER

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.

Magnetic sensor for steady state tokamak

Energy Technology Data Exchange (ETDEWEB)

Neyatani, Yuzuru; Mori, Katsuharu; Oguri, Shigeru; Kikuchi, Mitsuru [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1996-06-01

A new type of magnetic sensor has been developed for the measurement of steady state magnetic fields without DC-drift such as integration circuit. The electromagnetic force induced to the current which leads to the sensor was used for the measurement. For the high frequency component which exceeds higher than the vibration frequency of sensor, pick-up coil was used through the high pass filter. From the results using tokamak discharges, this sensor can measure the magnetic field in the tokamak discharge. During {approx}2 hours measurement, no DC drift was observed. The sensor can respond {approx}10ms of fast change of magnetic field during disruptions. We confirm the extension of measured range to control the current which leads to the sensor. (author).

Review of tokamak power reactor and blanket designs in the United States

International Nuclear Information System (INIS)

Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

1986-01-01

The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

High-β steady-state advanced tokamak regimes for ITER and FIRE

International Nuclear Information System (INIS)

Meade, D.M.; Sauthoff, N.R.; Kessel, C.E.; Budny, R.V.; Gorelenkov, N.; Jardin, S.C.; Schmidt, J.A.; Navratil, G.A.; Bialek, J.; Ulrickson, M.A.; Rognlein, T.; Mandrekas, J.

2005-01-01

An attractive tokamak-based fusion power plant will require the development of high-β steady-state advanced tokamak regimes to produce a high-gain burning plasma with a large fraction of self-driven current and high fusion-power density. Both ITER and FIRE are being designed with the objective to address these issues by exploring and understanding burning plasma physics both in the conventional H-mode regime, and in advanced tokamak regimes with β N ∼ 3 - 4, and f bs ∼50-80%. ITER has employed conservative scenarios, as appropriate for its nuclear technology mission, while FIRE has employed more aggressive assumptions aimed at exploring the scenarios envisioned in the ARIES power-plant studies. The main characteristics of the advanced scenarios presently under study for ITER and FIRE are compared with advanced tokamak regimes envisioned for the European Power Plant Conceptual Study (PPCS-C), the US ARIES-RS Power Plant Study and the Japanese Advanced Steady-State Tokamak Reactor (ASSTR). The goal of the present work is to develop advanced tokamak scenarios that would fully exploit the capability of ITER and FIRE. This paper will summarize the status of the work and indicate critical areas where further R and D is needed. (author)

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.

Burn cycle requirements comparison of pulsed and steady-state tokamak reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Ehst, D.A.

1983-12-01

Burn cycle parameters and energy transfer system requirements were analyzed for an 8-m commercial tokamak reactor using four types of cycles: conventional, hybrid, internal transformer, and steady state. Not surprisingly, steady state is the best burn mode if it can be achieved. The hybrid cycle is a promising alternative to the conventional. In contrast, the internal transformer cycle does not appear attractive for the size tokamak in question

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

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...

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)

Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

International Nuclear Information System (INIS)

Baker, C.C.

1980-01-01

STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described

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

Recent progresses on high performance steady-state plasmas in the superconducting tokamak TRIAM-1M

International Nuclear Information System (INIS)

Itoh, Satoshi; Sato, Kohnosuke; Nakamura, Kazuo

1999-01-01

The overview of TRIAM-1M experiments is described. The up-to-date issues for steady-state operation are presented through the experience of the achievement of super ultra long tokamak discharges (SULD) sustained by lower hybrid current drive (LHCD) over 2 hours. The importance of the control of an initial phase of plasma, the avoidance of the concentration of huge heat load, the wall conditioning, and abrupt stop of the long discharges are proposed as the indispensable issues for the achievement of the steady-state operation of tokamak. A high ion temperature (HIT) discharge fully sustained by 2.45 GHz LHCD with both high ion temperature and steep temperature gradient is successfully demonstrated for longer than 1 min in the limiter configuration. The HIT discharges can be obtained in the narrow window of density and position. Moreover, the avoidance of the concentration of heat load on a limiter is the key point for the achievement and its long sustainment. As the effective thermal insulation between the wall and the plasma is improved on the single null configuration, HIT discharges with peak ion temperature > 5keV and steeper gradient up to 85 keV/m can be achieved by the exquisite control of density and position. The plasmas with high κ ∼1.5 can be also demonstrated for longer than 1 min. The current profile is also well-controlled for about 2 orders in magnitude longer than the current diffusion time using combined LHCD. The serious damage to the material of the first wall caused by energetic neutral particles produced via charge exchange process is also described. As the neutral particles cannot be affected by magnetic field, this damage by neutral particles must be avoided by the new technique. (author)

Vulcan: A steady-state tokamak for reactor-relevant plasma–material interaction science

International Nuclear Information System (INIS)

Olynyk, G.M.; Hartwig, Z.S.; Whyte, D.G.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Mumgaard, R.T.; Podpaly, Y.A.

2012-01-01

Highlights: ► A new scaling for obtaining reactor similarity in the divertor of scaled tokamaks. ► Conceptual design for a tokamak (“Vulcan”) to implement this new scaling. ► Demountable superconducting coils and compact neutron shielding. ► Helium-cooled high-temperature vacuum vessel and first wall. ► High-field-side lower hybrid current drive for non-inductive operation. - Abstract: An economically viable magnetic-confinement fusion reactor will require steady-state operation and high areal power density for sufficient energy output, and elevated wall/blanket temperatures for efficient energy conversion. These three requirements frame, and couple to, the challenge of plasma–material interaction (PMI) for fusion energy sciences. Present and planned tokamaks are not designed to simultaneously meet these criteria. A new and expanded set of dimensionless figures of merit for PMI have been developed. The key feature of the scaling is that the power flux across the last closed flux surface P/S ≃ 1 MW m −2 is to be held constant, while scaling the core volume-averaged density weakly with major radius, n ∼ R −2/7 . While complete similarity is not possible, this new “P/S” or “PMI” scaling provides similarity for the most critical reactor PMI issues, compatible with sufficient current drive efficiency for non-inductive steady-state core scenarios. A conceptual design is developed for Vulcan, a compact steady-state deuterium main-ion tokamak which implements the P/S scaling rules. A zero-dimensional core analysis is used to determine R = 1.2 m, with a conventional reactor aspect ratio R/a = 4.0, as the minimum feasible size for Vulcan. Scoping studies of innovative fusion technologies to support the Vulcan PMI mission were carried out for three critical areas: a high-temperature, helium-cooled vacuum vessel and divertor design; a demountable superconducting toroidal field magnet system; and a steady-state lower hybrid current drive system

Implications of rf current drive theory for next step steady-state tokamak design

International Nuclear Information System (INIS)

Schultz, J.H.

1985-06-01

Two missions have been identified for a next-step tokamak experiment in the United States. The more ambitious Mission II device would be a superconducting tokamak, capable of doing long-pulse ignition demonstrations, and hopefully capable of also being able to achieve steady-state burn. A few interesting lines of approach have been identified, using a combination of logical design criteria and parametric system scans [SC85]. These include: (1) TIBER: A point-design suggested by Lawrence Livermore, that proposes a machine with the capability of demonstrating ignition, high beta (10%) and high Q (=10), using high frequency, fast-wave current drive. The TIBER topology uses moderate aspect ratio and high triangularity to achieve high beta. (2) JET Scale-up. (3) Magic5: It is argued here that an aspect ratio of 5 is a magic number for a good steady-state current drive experiment. A moderately-sized machine that achieves ignition and is capable of high Q, using either fast wave or slow wave current drive is described. (4) ET-II: The concept of a highly elongated tokamak (ET) was first proposed as a low-cost approach to Mission I, because of the possibility of achieving ohmic ignition with low-stress copper magnets. We propose that its best application is really for commercial tokamaks, using fast-wave current drive, and suggest a Mission II experiment that would be prototypical of such a reactor

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.

Conceptual design of the steady state tokamak reactor (SSTR)

International Nuclear Information System (INIS)

Oikawa, A.; Kikuchi, M.; Seki, Y.; Nishio, S.; Ando, T.; Ohara, Y.; Takizuka, Tani, K.; Ozeki, T.; Koizumi, K.; Ikeda, B.; Suzuki, Y.; Ueda, N.; Kageyama, T.; Yamada, M.; Mizoguchi, T.; Iida, F.; Ozawa, Y.; Mori, S.; Yamazaki, S.; Kobayashi, T.; Adachi, H.J.; Shinya, K.; Ozaki, A.; Asahara, M.; Konishi, K.; Yokogawa, N.

1992-01-01

This paper reports that on the basis of a high bootstrap current fraction observation with JT-60, the concept of steady state tokamak reactor , the SSTR, was conceived and was evolved with the design activity of the SSTR at JAERI. Also results of ITER/FER design activities has enhanced the SSTR design. Moreover the remarkable progress of R and D for fusion reactor engineering, especially in the development of superconducting coils and negative ion based NBI at JAERI have promoted the SSTR conceptual design as a realistic power reactor. Although present fusion power reactor designs are currently considered to be too large and costly, results of the SSTR conceptual design suggest that an efficient and promising tokamak reactor will be feasible. The conceptual design of the SSTR provides a realistic reference for a demo tokamak reactor

Concept study of the Steady State Tokamak Reactor (SSTR)

International Nuclear Information System (INIS)

1991-06-01

The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)

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...

Bifurcated states of a rotating tokamak plasma in the presence of a static error-field

International Nuclear Information System (INIS)

Fitzpatrick, R.

1998-01-01

The bifurcated states of a rotating tokamak plasma in the presence of a static, resonant, error-field are strongly analogous to the bifurcated states of a conventional induction motor. The two plasma states are the open-quotes unreconnectedclose quotes state, in which the plasma rotates and error-field-driven magnetic reconnection is suppressed, and the open-quotes fully reconnectedclose quotes state, in which the plasma rotation at the rational surface is arrested and driven magnetic reconnection proceeds without hindrance. The response regime of a rotating tokamak plasma in the vicinity of the rational surface to a static, resonant, error-field is determined by three parameters: the normalized plasma viscosity, P, the normalized plasma rotation, Q 0 , and the normalized plasma resistivity, R. There are 11 distinguishable response regimes. The extents of these regimes are calculated in P endash Q 0 endash R space. In addition, an expression for the critical error-field amplitude required to trigger a bifurcation from the open-quotes unreconnectedclose quotes to the open-quotes fully reconnectedclose quotes state is obtained in each regime. The appropriate response regime for low-density, ohmically heated, tokamak plasmas is found to be the nonlinear constant-ψ regime for small tokamaks, and the linear constant-ψ regime for large tokamaks. The critical error-field amplitude required to trigger error-field-driven magnetic reconnection in such plasmas is a rapidly decreasing function of machine size, indicating that particular care may be needed to be taken to reduce resonant error-fields in a reactor-sized tokamak. copyright 1998 American Institute of Physics

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.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

International Nuclear Information System (INIS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-01-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

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.

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

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

Adaptive optimal stochastic state feedback control of resistive wall modes in tokamaks

International Nuclear Information System (INIS)

Sun, Z.; Sen, A.K.; Longman, R.W.

2006-01-01

An adaptive optimal stochastic state feedback control is developed to stabilize the resistive wall mode (RWM) instability in tokamaks. The extended least-square method with exponential forgetting factor and covariance resetting is used to identify (experimentally determine) the time-varying stochastic system model. A Kalman filter is used to estimate the system states. The estimated system states are passed on to an optimal state feedback controller to construct control inputs. The Kalman filter and the optimal state feedback controller are periodically redesigned online based on the identified system model. This adaptive controller can stabilize the time-dependent RWM in a slowly evolving tokamak discharge. This is accomplished within a time delay of roughly four times the inverse of the growth rate for the time-invariant model used

Contour analysis of steady state tokamak reactor performance

International Nuclear Information System (INIS)

Devoto, R.S.; Fenstermacher, M.E.

1990-01-01

A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

Wave-driver options for low-aspect-ratio steady-state tokamak reactors

International Nuclear Information System (INIS)

Ehst, D.A.

1981-02-01

Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A = 2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value

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

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

SST: Single-Stream Temporal Action Proposals

KAUST Repository

Buch, Shyamal; Escorcia, Victor; Shen, Chuanqi; Ghanem, Bernard; Niebles, Juan Carlos

2017-01-01

Our paper presents a new approach for temporal detection of human actions in long, untrimmed video sequences. We introduce Single-Stream Temporal Action Proposals (SST), a new effective and efficient deep architecture for the generation of temporal action proposals. Our network can run continuously in a single stream over very long input video sequences, without the need to divide input into short overlapping clips or temporal windows for batch processing. We demonstrate empirically that our model outperforms the state-of-the-art on the task of temporal action proposal generation, while achieving some of the fastest processing speeds in the literature. Finally, we demonstrate that using SST proposals in conjunction with existing action classifiers results in improved state-of-the-art temporal action detection performance.

SST: Single-Stream Temporal Action Proposals

KAUST Repository

Buch, Shyamal

2017-11-09

Our paper presents a new approach for temporal detection of human actions in long, untrimmed video sequences. We introduce Single-Stream Temporal Action Proposals (SST), a new effective and efficient deep architecture for the generation of temporal action proposals. Our network can run continuously in a single stream over very long input video sequences, without the need to divide input into short overlapping clips or temporal windows for batch processing. We demonstrate empirically that our model outperforms the state-of-the-art on the task of temporal action proposal generation, while achieving some of the fastest processing speeds in the literature. Finally, we demonstrate that using SST proposals in conjunction with existing action classifiers results in improved state-of-the-art temporal action detection performance.

State diagrams of tokamaks and state transitions

International Nuclear Information System (INIS)

Minardi, E.

1992-01-01

In a simple one-fluid cylindrical model of transport and of dissipative effects, the family of the magnetic states of the Tokamak which correspond to a vanishing entropy production in the confinement region is characterized by a define relation or ''state equation'' involving the relevant parameters of the discharge. An investigation is made as to how the entropy production changes when the current density profile is rearranged by a perturbation which conserves the poloidal magnetic flux. It is shown that for a sufficiently short time interval, that is to say t 2 E τ s where τ E is the energy confinement time and τ s is the resistive time, neighbouring bifurcating equilibria exist which can be reached with a flux-conserving transition and with increase of the magnetic entropy. The family of these new states can also be characterized by a state equation involving the relevant discharge parameters. When the state equations of the two families are simultaneously satisfied by the same set of parameter values, a flux-conserving, entropy-increasing transition may take place between states of the two families. The modifications of the current density and of the temperature profiles involved in the transition and the conditions that the discharge parameters should satisfy in order that the transition could occur are investigated. (author)

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

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

A dynamic state observer for real-time reconstruction of the tokamak plasma profile state and disturbances

NARCIS (Netherlands)

Felici, F.; De Baar, M.; Steinbuch, M.

2014-01-01

A dynamic observer is presented which can reconstruct the internal state of a tokamak fusion plasma, consisting of the spatial distribution of current and temperature, from measurements. Today, the internal plasma state is usually reconstructed by solving an ill-conditioned inversion problem using a

Steady-state operation requirements of tokamak fusion reactor concepts

International Nuclear Information System (INIS)

Knobloch, A.F.

1991-06-01

In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

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.

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.

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,...

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.

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

Operating tokamaks with steady-state toroidal current

International Nuclear Information System (INIS)

Fisch, N.J.

1981-04-01

Continuous operation of a tokamak requires, among other things, a means of continuously providing the toroidal current. Various methods have been proposed to provide this current including methods which utilize radio-frequency waves in any of several frequency regimes. Here we elaborate on the prospects of incorporating these current-drive techniques in tokamak reactors, concentrating on the theoretical minimization of the power requirements

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...

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

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

Time evolution of tokamak states with flow

International Nuclear Information System (INIS)

Kerner, W.; Weitzner, H.

1985-12-01

The general dissipative Braginskii single-fluid model is applied to simulate tokamak transport. An expansion with respect to epsilon = (ω/sub i/tau/sub i/) -1 , the factor by which perpendicular and parallel transport coefficients differ, yields a numerically tractable scheme. The resulting 1-1/2 D procedure requires computation of 2D toroidal equilibria with flow together with the solution of a system of ordinary 1D flux-averaged equations for the time evolution of the profiles. 13 refs

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...

Implementing a finite-state off-normal and fault response system for disruption avoidance in tokamaks

Science.gov (United States)

Eidietis, N. W.; Choi, W.; Hahn, S. H.; Humphreys, D. A.; Sammuli, B. S.; Walker, M. L.

2018-05-01

A finite-state off-normal and fault response (ONFR) system is presented that provides the supervisory logic for comprehensive disruption avoidance and machine protection in tokamaks. Robust event handling is critical for ITER and future large tokamaks, where plasma parameters will necessarily approach stability limits and many systems will operate near their engineering limits. Events can be classified as off-normal plasmas events, e.g. neoclassical tearing modes or vertical displacements events, or faults, e.g. coil power supply failures. The ONFR system presented provides four critical features of a robust event handling system: sequential responses to cascading events, event recovery, simultaneous handling of multiple events and actuator prioritization. The finite-state logic is implemented in Matlab®/Stateflow® to allow rapid development and testing in an easily understood graphical format before automated export to the real-time plasma control system code. Experimental demonstrations of the ONFR algorithm on the DIII-D and KSTAR tokamaks are presented. In the most complex demonstration, the ONFR algorithm asynchronously applies ‘catch and subdue’ electron cyclotron current drive (ECCD) injection scheme to suppress a virulent 2/1 neoclassical tearing mode, subsequently shuts down ECCD for machine protection when the plasma becomes over-dense, and enables rotating 3D field entrainment of the ensuing locked mode to allow a safe rampdown, all in the same discharge without user intervention. When multiple ONFR states are active simultaneously and requesting the same actuator (e.g. neutral beam injection or gyrotrons), actuator prioritization is accomplished by sorting the pre-assigned priority values of each active ONFR state and giving complete control of the actuator to the state with highest priority. This early experience makes evident that additional research is required to develop an improved actuator sharing protocol, as well as a methodology to

Physics design of an ultra-long pulsed tokamak reactor

International Nuclear Information System (INIS)

Ogawa, Y.; Inoue, N.; Wang, J.; Yamamoto, T.; Okano, K.

1993-01-01

A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

Overview of steady-state tokamak operation and current drive experiments in TRIAM-1M

International Nuclear Information System (INIS)

Zushi, H.; Nakamura, K.; Hanada, K.

2005-01-01

Experiments aiming at 'day long operation at high performance' have been carried out. The record value of the discharge duration was updated to 5 h and 16 min. Steady-state tokamak operation (SSTO) is studied under the localized PWI conditions. The distributions of the heat load, the particle recycling flux and impurity source are investigated to understand the co-deposition and wall pumping. Formation and sustainment of an internal transport barrier ITB in enhanced current drive mode (ECD) has been investigated by controlling the lower hybrid driven current profile by changing the phase spectrum. An ITER relevant remote steering antenna for electron cyclotron wave ECW injection was installed and a relativistic Doppler resonance of the oblique propagating extraordinary wave with energetic electrons driven by lower hybrid waves was studied. (author)

Internal transport barrier physics for steady state operation in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Wakatani, Masahiro [Kyoto Univ., Graduate School of Engineering, Uji, Kyoto (Japan); Fukuda, Takeshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Connor, Jack W. [Culham Science Centre, EURATOM/UKAEA Association (United Kingdom); Garbet, Xavier [Culham Science Centre, EFDA-JET CSU (United Kingdom); Gormezano, Claude [Associazone EURATOM-ENEA sulla Fusione C.R. Frascati (Italy); Mukhovatov, Vladimir [ITER Naka Joint Work Site, ITER Physics Unit, Naka, Ibaraki (Japan)

2003-07-01

Experimental results for the ITB (Internal Transport Barrier) formation and sustainment are compiled in a unified manner to find common features of ITBs in tokamaks. Global scaling laws for threshold power to obtain the ITBs are discussed. Theoretical models for plasmas with ITBs are summarized from stability and transport point of view. Finally possibility to obtain steady-state ITBs will be discussed in addition to extrapolation to ITER. (author)

Tokamaks. 2. ed.

International Nuclear Information System (INIS)

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

1997-01-01

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

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

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.

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...

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

Magnetic mapping of the TBR-1 tokamak

International Nuclear Information System (INIS)

Matta, Jose Antonio Sevidanes da.

1994-01-01

Axisymmetric hydromagnetic equilibria for an ideal conducting current fluid are described by means of an asymptotic expansion in powers of the inverse aspect ratio (ε α/R 0 ) that satisfies the Grad-Shafranov equation (equilibrium condition). The main profiles for these equilibria were computer. The effect of non-axisymmetric perturbations on the magnetic surfaces if the tokamak TBR-1 is investigated. The used method is able to show how magnetic field from external helical currents split the rational magnetic surface giving rise to the chain of magnetic islands. The Poincare map of the field line perturbed by resonant helical windings has been obtained numerically for the typical TBR-1 parameters. For increasing parameters of perturbation secondary resonances were observed, that transform the bound-state-like contours of a given island into similar structures of secondary magnetic islands. These results were used to determine the spectrum of the perturbation created by resonant helical windings. (author). 87 refs., 30 figs

Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime

Science.gov (United States)

Ren, Q.

2015-11-01

Recent DIII-D experiments have advanced the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Fully noninductive plasmas with extremely high values of the poloidal beta, βp >= 4 , have been sustained at βT >= 2 % for long durations with excellent energy confinement quality (H98y,2 >= 1 . 5) and internal transport barriers (ITBs) generated at large minor radius (>= 0 . 6) in all channels (Te, Ti, ne, VTf). Large bootstrap fraction (fBS ~ 80 %) has been obtained with high βp. ITBs have been shown to be compatible with steady state operation. Because of the unusually large ITB radius, normalized pressure is not limited to low βN values by internal ITB-driven modes. βN up to ~4.3 has been obtained by optimizing the plasma-wall distance. The scenario is robust against several variations, including replacing some on-axis with off-axis neutral beam injection (NBI), adding electron cyclotron (EC) heating, and reducing the NBI torque by a factor of 2. This latter observation is particularly promising for extension of the scenario to EAST, where maximum power is obtained with balanced NBI injection, and to a reactor, expected to have low rotation. However, modeling of this regime has provided new challenges to state-of-the-art modeling capabilities: quasilinear models can dramatically underpredict the electron transport, and the Sauter bootstrap current can be insufficient. The analysis shows first-principle NEO is in good agreement with experiments for the bootstrap current calculation and ETG modes with a larger saturated amplitude or EM modes may provide the missing electron transport. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA

Investigation of plasma equilibrium in HL-1 tokamak

International Nuclear Information System (INIS)

Lu Zhihong; Jiang Yunxia; Yang Jinwei; Zhang Baozhu; Qiu Wei; Qin Yunwen

1987-01-01

In this paper, the plasma equilibrium in HL-1 tokamak has been discussed. The horizontal and vertical displacement of plasma is measured using a symmetical magneic probe system, and the temporal evolution of displacements is given by a data acquisition system with micro-computer. The influence of various stray fields on plasma equilibrium has been analysed. The direction and value of horizontal stray field induced by the totoidal field coils and primary windings are determined using vertical displacement data. By adjusting parameters of internal and external vertical field, in the flat part of discharge current, the plasma can be kept in its equilibuium state at the place where is 3 cm outer of chamber certre, i.e nearby the centre of limiter

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)

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

this problem, we have validated and scaled our design calculations with performance parameters of the Neutral Beam Injector at IPP, Julich, Germany. The performance test of the SST-1 PINI ion source was done at MARION Test Stand at IPP, Julich. Analyses of these results indicate that the measured power profile and the optical parameters of the beam are in good agreement with the simulation results. These parameters are stable over the beam pulse of 14s with extracted beam energy of 31 MJ at 41 kV. This paper presents these results and details out future work need to be done in order to assess the steady state stability of the beam parameters.

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.

Recent SST trends and Flood Disasters in Brazil

Science.gov (United States)

Yamashiki, Y.; Behera, S. K.; Inoue, S.; Netrananda, S.; Silva, R. D.; Takara, K. T.; Yamagata, T.

2010-12-01

We analyzed recent variations in the sea surface temperature (SST) anomalies of Pacific and Atlantic Oceans to understand their roles in extreme discharge of Amazon River Basin. In general, higher than monthly average discharge appears when La Niña condition forms and lower than monthly average discharge appears when El Niño condition forms. We also investigated the relationship between SST anomalies and recent floods in Brazil during the period of 1980-2010. Most severe floods (e.g. 2003 and 2010 Rio de Janeiro-São Paulo Flood) in austral summer occurred when El Niño Modoki appears in the Pacific Ocean. In addition, warm waters in tropical South Atlantic Ocean between American and African Coast also helped the moisture convergence to the affected region. Floods in some other locations (for example, Itaipava flood occurred in Maranhao State in 2008) occurred when a La Niña Modoki appeared in Pacific Ocean. These flood disasters in Brazil associated with climate phenomena may increase due to warmer SST trend under the global warming stress.

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

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.

On the minimum circulating power of steady state tokamaks

International Nuclear Information System (INIS)

Itoh, K.; Itoh, S.; Fukuyama, A.; Yagi, M.

1995-07-01

Circulating power for the sustenance and profile control of the steady state tokamak plasmas is discussed. The simultaneous fulfillment of the MHD stability at high beta value, the improved confinement and the stationary equilibrium requires the rotation drive as well as the current drive. In addition to the current drive efficiency, the efficiency for the rotation drive is investigated. The direct rotation drive by the external torque, such as the case of beam injection, is not efficient enough. The mechanism and the magnitude of the spontaneous plasma rotation are studied. (author)

Ultra-long pulse operation using lower hybrid waves on the superconducting high field tokamak TRIAM-1M

International Nuclear Information System (INIS)

Moriyama, S.; Nakamura, Y.; Nagao, A.; Jotaki, E.; Nakamura, K.; Hiraki, N.; Itoh, S.

1990-01-01

Ultra-long pulse operation (>3 min) was achieved on the superconducting high field tokamak TRIAM-1M. In this operation, the plasma current was maintained with a relatively peaked current distribution by the 2.45 GHz radiofrequency power (P RF ≤ 35 kW) alone. A stationary plasma with a driven current of up to 35 kA and a line averaged electron density of up to 3x10 12 cm -3 was produced by precise plasma position and gas feed control. The extremely long discharge showed the interesting characteristics that the high temperatures of about 1 keV for the electrons and about 0.5 keV for the ions were kept almost constant during steady state current drive and that there was no impurity accumulation which could have a fatally adverse effect on steady state tokamak operation. (author). 16 refs, 17 figs

Tokamak and RFP ignition requirements

International Nuclear Information System (INIS)

Werley, K.A.

1991-01-01

A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

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

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.

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.

Demonstration of SST value as EBVs descriptor in the Mediterranean Sea

Science.gov (United States)

Valentini, E.; Filipponi, F.; Nguyen Xuan, A.; Taramelli, A.

2017-12-01

Sea Surface Temperature is an Essential Climate and Ocean Variable (ECV - EOV) able to capture critical scales in the seascape warming patterns and to highlight the exceeding of thresholds. This presentation addresses the changes of the SST in the last three decades over the Mediterranean Sea, a "Large Marine Ecosystem (LME)", in order to speculate the value of such powerful variable, as proxy for the assessment of ecosystem state in terms of ecosystem structures, functions and composition key descriptor. Time series of daily SST for the period 1982-2016, estimated from multi-sensor satellite data and provided by Copernicus Marine Environment Monitoring Service (CMEMS-EU) are used to perform different statistical analysis on common fish species. Results highlight the critical conditions, the general trends as well as the spatial and temporal patterns, in terms of thermal growth, vitality and stress influence on selected fish species. Results confirm a constant increasing trend in SST with an average rise of 1.4° C in the past thirty years. The variance associated to the average trend is not constant across the entire Mediterranean Sea opening the way to multiple scenarios for fish growth and vitality in the diverse sub-basins. A major effort is oriented in addressing the cross-scale ecological interactions to assess the feasibility of using SST as descriptor for Essential Biodiversity Variables, able to prioritize areas and to feed operational tools for planning and management in the Mediterranean LME.

Three novel tokamak plasma regimes in TFTR

International Nuclear Information System (INIS)

Furth, H.P.

1985-10-01

Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

A comparison of steady-state ARIES and pulsed PULSAR tokamak power plants

International Nuclear Information System (INIS)

Bathke, C.G.

1994-01-01

The multi-institutional ARIES study has completed a series of three steady-state and two pulsed cost-optimized conceptual designs of commercial tokamak fusion power plants that vary the level of assumed advances in technology and physics. The cost benefits of various design options are compared quantitatively. Possible means to improve the economic competitiveness of fusion are suggested

[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

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

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.)

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.

TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Henning, C.D.; Gilleland, J.R.

1988-01-01

The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

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

Mass transport and the bootstrap current from Ohm's law in steady-state tokamaks

International Nuclear Information System (INIS)

Kim, J.-S.; Greene, J.M.

1989-01-01

The consequences of mass conservation and Ohm's law are examined for steady state Tokamaks. In a Tokamak, magnetofluid-dynamic waves rapidly equilibrate pressure and toroidal field along magnetic surfaces. As a result, the detailed current distribution is determined by the flux surface averaged poloidal and toroidal currents. The electrons that carry the plasma current are impeded in their motion by interactions with ions, which is resistivity and its generalizations, and by interactions with electrons, which is viscosity and its generalizations. The important viscous terms arise from the interaction between trapped and untrapped electrons, and so viscosity acts by impeding poloidal current. properly chosen, the results of neoclassical theory are The neoclassical viscous coefficient is here regarded as less likely than Spitzer conductivity to be experimentally relevant in a turbulent Tokamak. Thus, the toroidal Ohm's law is regarded as being more reliable than the poloidal Ohm's law. A combination of toroidal and poloidal Ohm's law, namely the component parallel to the magnetic field, eliminates the influence of plasma fueling, and directly relates the bootstrap current and the pressure gradient. The latter is the usual relation, but, since i

Vaporization Mode and State of the Ablatant of a Deuterium Pellet in Tokamak Discharges

DEFF Research Database (Denmark)

Chang, C. T.

1983-01-01

The ablation of a deuterium pellet under prevailing tokamak discharge conditions is shown to be a dynamic phase transition process. An alternative boundary condition at the pellet surface is formulated. Computational results based on the new boundary condition showed that the state of the ablatant...

Tokamak Physics Experiment (TPX) design

International Nuclear Information System (INIS)

Schmidt, J.A.

1995-01-01

TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

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.

A steady state tokamak operation by use of magnetic monopoles

International Nuclear Information System (INIS)

Narihara, K.

1991-12-01

A steady state tokamak operation based on a magnetic monopole circuit is considered. Circulation of a chain of iron cubes which trap magnetic monopoles generates the needed loop voltage. The monopole circuit is enclosed by a series of solenoid coils in which magnetic field is feedback controlled so that the force on the circuit balance against the mechanical friction. The driving power is supplied through the current sources of poloidal, ohmic and solenoid coils. The current drive efficiency is same as that of the ohmic current drive. (author)

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)

The ARIES tokamak fusion reactor study

International Nuclear Information System (INIS)

Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

1989-01-01

The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

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...

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.

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

Magnetic field structure of experimental high beta tokamak equilibria

International Nuclear Information System (INIS)

Deniz, A.V.

1986-01-01

The magnetic field structure of several low and high β tokamaks in the Columbia High Beta Tokamak (HBT) was determined by high-impedance internal magnetic probes. From the measurement of the magnetic field, the poloidal flux, toroidal flux, toroidal current, and safety factor are calculated. In addition, the plasma position and cross-sectional shape are determined. The extent of the perturbation of the plasma by the probe was investigated and was found to be acceptably small. The tokamaks have major radii of approx.0.24 m, minor radii of approx.0.05 m, toroidal plasma current densities of approx.10 6 A/m 2 , and line-integrated electron densities of approx.10 20 m -2 . The major difference between the low and high β tokamaks is that the high β tokamak was observed to have an outward shift in major radius of both the magnetic center and peak of the toroidal current density. The magnetic center moves inward in major radius after 20 to 30 μsec, presumably because the plasma maintains major radial equilibrium as its pressure decreases from radiation due to impurity atoms. Both the equilibrium and the production of these tokamaks from a toroidal field stabilized z-pinch are modeled computationally. One tokamak evolves from a state with low β features, through a possibly unstable state, to a state with high β features

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.

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...

Advanced tokamak burning plasma experiment

International Nuclear Information System (INIS)

Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.

2001-01-01

A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

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...

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)

Presheath profiles in simulated tokamak edge plasmas

International Nuclear Information System (INIS)

LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

1988-04-01

The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

Design of a 3.7 GHz oscillator for the solid state drive of the LHCD system

International Nuclear Information System (INIS)

Sainkar, Sandeep; Dixit, Harish; Cheeran, Alice; Sharma, P.K.

2017-01-01

The LHCD system is commissioned on the SST-1 tokamak for the current drive. It has a capability to generate power of 2 MW CW at 3.7 GHz and deliver the power to the tokamak via a grill antenna through a phased array of wave guides. The system relies on 4 Klystrons (TH-2103D) each generating 500 kW CW power. The klystrons act as an amplifier providing a gain of 40 dB with a bandwidth of 10 MHz and amplify the input power provided by a solid state driver. The klystron requires a supply of 65 kV and 20A for its operation and has to be extensively conditioned before it can be operated even for obtaining lower power levels. This paper describes the design of oscillator for this system. The oscillator is based on bipolar junction transistor BFR360F. Linear and non-linear analysis has been performed on the design to ascertain its performance. The oscillator delivered a power of 20 mW at 3.7 GHz

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

Saturated ideal kink/peeling formations described as three-dimensional magnetohydrodynamic tokamak equilibrium states

Energy Technology Data Exchange (ETDEWEB)

Cooper, W. A.; Brunetti, D.; Duval, B. P.; Faustin, J. M.; Graves, J. P.; Kleiner, A.; Patten, H.; Pfefferlé, D.; Porte, L.; Raghunathan, M.; Reimerdes, H.; Sauter, O.; Tran, T. M., E-mail: wilfred.cooper@epfl.ch [Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne (Switzerland)

2016-04-15

Free boundary magnetohydrodynamic equilibrium states with spontaneous three dimensional deformations of the plasma-vacuum interface are computed for the first time. The structures obtained have the appearance of saturated ideal external kink/peeling modes. High edge pressure gradients yield toroidal mode number n = 1 corrugations for a high edge bootstrap current and larger n distortions when this current is small. Deformations in the plasma boundary region induce a nonaxisymmetric Pfirsch-Schlüter current driving a field-aligned current ribbon consistent with reported experimental observations. A variation in the 3D equilibrium confirms that the n = 1 mode is a kink/peeling structure. We surmise that our calculated equilibrium structures constitute a viable model for the edge harmonic oscillations and outer modes associated with a quiescent H-mode operation in shaped tokamak plasmas.

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....

HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

International Nuclear Information System (INIS)

Schultz, J.H.; Montgomery, D.B.

1983-04-01

HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

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)

Tokamak poloidal-field systems. Progress report, January 1-December 31, 1981

International Nuclear Information System (INIS)

Rogers, J.D.

1982-03-01

Work on the superconducting tokamak poloidal field system (TPFS) program is being redirected. The development of the 20 MJ, 50 kA, 7.5 T superconducting programmed energy storage coil is being terminated. The superconductor for the 20 MJ coil is being processed only to an intermediate state, and manufacture of the epoxy fiberglass dewar is being stopped. Further, development of the TPFS test facility is in abeyance. Change in program emphasis arises from prospective rf plasma current driven or beam heated tokamaks with programmed coil characteristics for the poloidal field being different from those to have been simulated by the 20 MJ coil and from budgetary constraints. Work is reported on the development of the coil, conductor, nonconducting dewar, and test facility to the recent time when the program change was instigated. Work in support of the Large Coil Test Facility (LCTF) and the Fusion Engineering Design (FED) Center is given. Analysis of the experiments on the 400 kJ METS coil test was completed

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.)

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

Study of a compact reversed shear Tokamak reactor

International Nuclear Information System (INIS)

Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.

1998-01-01

A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)

Magnetic confinement experiment. I: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1995-08-01

Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

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

Compact tokamak reactors

International Nuclear Information System (INIS)

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

1997-01-01

The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

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

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 strategy for plasma equilibrium in a tokamak

International Nuclear Information System (INIS)

Miskell, R.V.

1975-08-01

Dynamic control of the plasma position within the torus of a TOKAMAK fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. (auth)

Preliminary design study of a steady state tokamak device

International Nuclear Information System (INIS)

Miya, Naoyuki; Nakajima, Shinji; Ushigusa, Kenkichi; and athors)

1992-09-01

Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)

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.

Physics design requirements for the Tokamak Physics Experiment (TPX)

International Nuclear Information System (INIS)

Neilson, G.H.; Goldston, R.J.; Jardin, S.C.; Reiersen, W.T.; Porkolab, M.; Ulrickson, M.

1993-01-01

The design of TPX is driven by physics requirements that follow from its mission. The tokamak and heating systems provide the performance and profile controls needed to study advanced steady state tokamak operating modes. The magnetic control systems provide substantial flexibility for the study of regimes with high beta and bootstrap current. The divertor is designed for high steady state power and particle exhaust

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

First observation of a new zonal-flow cycle state in the H-mode transport barrier of the experimental advanced superconducting Tokamak

DEFF Research Database (Denmark)

Xu, G.S.; Wang, H. Q.; Wan, B. N.

2012-01-01

A new turbulence-flow cycle state has been discovered after the formation of a transport barrier in the H-mode plasma edge during a quiescent phase on the EAST superconducting tokamak. Zonal-flow modulation of high-frequency-broadband (0.05-1MHz) turbulence was observed in the steep-gradient region...... leading to intermittent transport events across the edge transport barrier. Good confinement (H-98y,H-2 similar to 1) has been achieved in this state, even with input heating power near the L-H transition threshold. A novel model based on predator-prey interaction between turbulence and zonal flows...... reproduced this state well. © 2012 American Institute of Physics. [http://dx.doi.org/10.1063/1.4769852]...

Numerical simulation for HT-6M tokamak electrical transient behaviours

International Nuclear Information System (INIS)

Yu Yuanqi; Liu Baohua; Pan Yuan

1991-02-01

The following main points are concerned: (1) State equations used for dynamic analysis of all electrical parameters of the tokamak are derived. (2) In order to increase plasma volt-seconds and to get plasma current with longer sustainment phase, a power supply scheme for HT-6M and its numerical simulation are studied. (3) The distribution of energy flow in coupling loops of the tokamak is discussed, and the energy transfer ratio from the OH loop and vertical field loop to the plasma is also analyzed

On residual gas analysis during high temperature baking of graphite tiles

International Nuclear Information System (INIS)

Prakash, A A; Chaudhuri, P; Khirwadkar, S; Reddy, D Chenna; Saxena, Y C; Chauhan, N; Raole, P M

2008-01-01

Steady-state Super-conducting Tokamak-1 (SST-1) is a medium size tokamak with major radius of 1.1 m and minor radius of 0.20 m. It is designed for plasma discharge duration of 1000 seconds to obtain fully steady-state plasma operation. Plasma Facing Components (PFC), consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be UHV compatible for steady state operation. All PFC are made up of graphite tiles mechanically attached to the copper alloy substrate. Graphite is one of the preferred first wall armour material in present day tokamaks. High thermal shock resistance and low atomic number of carbon are the most important properties of graphite for this application. High temperature vacuum baking of graphite tiles is the standard process to remove the impurities. Residual Gas Analyzer (RGA) has been used for qualitative and quantitative measurements of released gases from graphite tiles during baking. Surface Analysis of graphite tiles has also been done before and after baking. This paper describes the residual gas analysis during baking and surface analysis of graphite tiles

On residual gas analysis during high temperature baking of graphite tiles

Energy Technology Data Exchange (ETDEWEB)

Prakash, A A; Chaudhuri, P; Khirwadkar, S; Reddy, D Chenna; Saxena, Y C [Institute for Plasma Research, Bhat, Gandhinagar - 382 428 (India); Chauhan, N; Raole, P M [Facilitation Center for Industrial Plasma Technologies, IPR, Gandhinagar (India)], E-mail: arun@ipr.res.in

2008-05-01

Steady-state Super-conducting Tokamak-1 (SST-1) is a medium size tokamak with major radius of 1.1 m and minor radius of 0.20 m. It is designed for plasma discharge duration of 1000 seconds to obtain fully steady-state plasma operation. Plasma Facing Components (PFC), consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be UHV compatible for steady state operation. All PFC are made up of graphite tiles mechanically attached to the copper alloy substrate. Graphite is one of the preferred first wall armour material in present day tokamaks. High thermal shock resistance and low atomic number of carbon are the most important properties of graphite for this application. High temperature vacuum baking of graphite tiles is the standard process to remove the impurities. Residual Gas Analyzer (RGA) has been used for qualitative and quantitative measurements of released gases from graphite tiles during baking. Surface Analysis of graphite tiles has also been done before and after baking. This paper describes the residual gas analysis during baking and surface analysis of graphite tiles.

Liquid tin limiter for FTU tokamak

Energy Technology Data Exchange (ETDEWEB)

Vertkov, A., E-mail: avertkov@yandex.ru [JSC “Red Star”, Moscow (Russian Federation); Lyublinski, I. [JSC “Red Star”, Moscow (Russian Federation); NRNU MEPhI, Moscow (Russian Federation); Zharkov, M. [JSC “Red Star”, Moscow (Russian Federation); Mazzitelli, G.; Apicella, M.L.; Iafrati, M. [Associazione EURATOM-ENEA sulla Fusione, C. R. Frascati, Frascati, Rome, Italy, (Italy)

2017-04-15

Highlights: • First steady state operating liquid tin limiter TLL is under study on FTU tokamak. • The cooling system with water spray coolant for TLL has been developed and tested. • High corrosion resistance of W and Mo in molten Sn confirmed up to 1000 °C. • Wetting process with Sn has been developed for Mo and W. - Abstract: The liquid Sn in a matrix of Capillary Porous System (CPS) has a high potential as plasma facing material in steady state operating fusion reactor owing to its physicochemical properties. However, up to now it has no experimental confirmation in tokamak conditions. First steady state operating limiter based on the CPS with liquid Sn installed on FTU tokamak and its experimental study is in progress. Several aspects of the design, structural materials and operation parameters of limiter based on tungsten CPS with liquid Sn are considered. Results of investigation of corrosion resistance of Mo and W in Sn and their wetting process are presented. The heat removal for limiter steady state operation is provided by evaporation of flowing gaswater spray. The effectiveness of such heat removal system is confirmed in modelling tests with power flux up to 5 MW/m2.

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...

Demonstration tokamak power plant

International Nuclear Information System (INIS)

Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

1983-01-01

A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System

Physical design of MW-class steady-state spherical tokamak, QUEST

International Nuclear Information System (INIS)

Hanada, K.; Sato, K.N.; Zushi, H.; Nakamura, K.; Sakamoto, M.; Idei, H.; Hasegawa, M.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Higashizono, Y.; Yoshida, N.; Takase, Y.; Ejiri, A.; Ogawa, Y.; Ono, Y.; Yoshida, Z.; Mitarai, O.; Maekawa, T.; Kishimoto, Y.; Ishiguro, M.; Yoshinaga, T.; Igami, H.; Hirooka, Y.; Komori, A.; Motojima, O.; Sudo, S.; Yamada, H.; Ando, A.; Asakura, Nobuyuki; Matsukawa, Makoto; Ishida, A.; Ohno, N.; Peng, M.

2008-10-01

QUEST (R=0.68 m, a=0.4 m) focuses on the steady state operation of the spherical tokamak (ST) by controlled PWI and electron Bernstain wave (EBW) current drive (CD). The QUEST project will be developed along two phases, phase I: steady state operation with plasma current, I p =20-30 kA on open divertor configuration and phase II: steady state operation with I p = 100 kA and β of 10% in short pulse on closed divertor configuration. Feasibility of the missions on QUEST was investigated and the suitable machine size of QUEST was decided based on the physical view of plasma parameters. Electron Bernstein wave (EBW) current drive are planned to establish the maintenance of plasma current in steady state. Mode conversion efficiency to EBW was calculated and the conversion of 95% will be expected. A new type antenna for QUEST has been fabricated to excite EBW effectively. The situation of heat and particle handling is challenging, and W and high temperature wall is adopted. The start-up scenario of plasma current was investigated based on the driven current by energetic electron and the most favorable magnetic configuration for start-up is proposed. (author)

A simulation study on burning profile tailoring of steady state, high bootstrap current tokamaks

International Nuclear Information System (INIS)

Nakamura, Y.; Takei, N.; Tobita, K.; Sakamoto, Y.; Fujita, T.; Fukuyama, A.; Jardin, S.C.

2007-01-01

From the aspect of fusion burn control in steady state DEMO plant, the significant challenges are to maintain its high power burning state of ∝3-5 GW without burning instability, hitherto well-known as ''thermal stability'', and also to keep its desired burning profile relevant with internal transport barrier (ITB) that generates high bootstrap current. The paper presents a simulation modeling of the burning stability coupled with the self-ignited fusion burn and the structure-formation of the ITB. A self-consistent simulation, including a model for improved core energy confinement, has pointed out that in the high power fusion DEMO plant there is a close, nonlinear interplay between the fusion burnup and the current source of non-inductive, ITB-generated bootstrap current. Consequently, as much distinct from usual plasma controls under simulated burning conditions with lower power (<<1 GW), the selfignited fusion burn at a high power burning state of ∝3-5 GW becomes so strongly selforganized that any of external means except fuelling can not provide the effective control of the stable fusion burn.It is also demonstrated that externally applied, inductive current perturbations can be used to control both the location and strength of ITB in a fully noninductive tokamak discharge. We find that ITB structures formed with broad noninductive current sources such as LHCD are more readily controlled than those formed by localized sources such as ECCD. The physics of the inductive current is well known. Consequently, we believe that the controllability of the ITB is generic, and does not depend on the details of the transport model (as long as they can form an ITB for sufficiently reversed magnetic shear q-profile). Through this external control of the magnetic shear profile, we can maintain the ITB strength that is otherwise prone to deteriorate when the bootstrap current increases. These distinguishing capabilities of inductive current perturbation provide steady

Linear and nonlinear kinetic-stability studies in tokamaks

International Nuclear Information System (INIS)

Tang, W.M.; Chance, M.S.; Chen, L.; Krommes, J.A.; Lee, W.W.; Rewoldt, G.

1982-09-01

This paper presents results of theoretical investigations on important linear kinetic properties of low frequency instabilities in toroidal systems and on nonlinear processes which could significantly influence their impact on anomalous transport. Analytical and numerical methods and also particle simulations have been employed to carry out these studies. In particular, the following subjects are considered: (1) linear stability analysis of kinetic instabilities for realistic tokamak equilibria and the application of such calculations to the PDX and PLT tokamak experiments including the influence of a hot beam-ion component; (2) determination of nonlinearly saturated, statistically steady states of three interacting drift modes; and (3) gyrokinetic particle simulation of drift instabilities

Economic analyses of alpha channeling in tokamak power plants

International Nuclear Information System (INIS)

Ehst, D.A.

1998-01-01

The hot-ion-mode of operation [1] has long been thought to offer optimized performance for long-pulse or steady-state magnetic fusion power plants. This concept was revived in recent years when theoretical considerations suggested that nonthermal fusion alpha particles could be made to channel their power density preferentially to the fuel ions [2,3]. This so-called anomalous alpha particle slowing down can create plasmas with fuel ion temperate T i somewhat larger than the electron temperature T e , which puts more of the beta-limited plasma pressure into the useful fuel species (rather than non-reacting electrons). As we show here, this perceived benefit may be negligible or nonexistent for tokamaks with steady state current drive. It has likewise been argued [2,3] that alpha channeling could be arranged such that little or no external power would be needed to generate the steady state toroidal current. Under optimistic assumptions we show that such alpha-channeling current drive would moderately improve the economic performance of a first stability tokamak like ARIES-I [4], however a reversed-shear (advanced equilibrium) tokamak would likely not benefit since traditional radio-wave (rf) electron-heating current drive power would already be quite small

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

Economic comparison of MHD equilibrium options for advanced steady state tokamak power plants

International Nuclear Information System (INIS)

Ehst, D.A.; Kessel, C.E.; Jardin, S.C.; Krakowski, R.A.; Bathke, C.G.; Mau, T.K.; Najmabadi, F.

1998-01-01

Progress in theory and in tokamak experiments leads to questions of the optimal development path for commercial tokamak power plants. The economic prospects of future designs are compared for several tokamak operating modes: (high poloidal beta) first stability, second stability and reverse shear. Using a simplified economic model and selecting uniform engineering performance parameters, this comparison emphasizes the different physics characteristics - stability and non- inductive current drive - of the various equilibria. The reverse shear mode of operation is shown to offer the lowest cost of electricity for future power plants. (author)

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)

Advanced tokamak physics in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)

2000-12-01

Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)

HT-7U superconducting tokamak: Physics design, engineering progress and schedule

International Nuclear Information System (INIS)

Wan Yuanxi

2002-01-01

The superconducting tokamak research program begun in China in ASIPP since 1994. The program is included in existent superconducting tokamak HT-7 and the next new superconducting tokamak HT-7U which is one of national key research projects in China. With the elongation cross-section, divertor and higher plasma parameter the main objectives of HT-7U are widely investigation both of the physics and technology for steady state advanced tokamak as well as the investigation of power and particle handle under steady-state operation condition. The physics and engineering design have been completed and significant progresses on R and D and fabrication have been achieved. HT-7U will begin assembly at 2003 and possible to get first plasma around 2004. (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

Steady-state resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1979-12-01

If spatially-averaged values of the beta ratio can reach 5 to 10% in tokamaks, as now seems likely, resistive toroidal-field coils may be advantageous for use in reactors intended for fusion-neutron applications. The present investigation has parameterized the design of steady-state water-cooled copper coils of rectangular cross section in order to maximize figures of merit such as the ratio of fusion neutron wall loading to coil power dissipation. Four design variations distinguished by different ohmic-heating coil configurations have been examined. For a wall loading of 0.5 MW/m 2 , minimum TF-coil lifetime costs (including capital and electricity costs) are found to occur with coil masses in the range 2400 to 4400 tons, giving 200 to 250 MW of resistive dissipation, which is comparable with the total power drain of the other reactor subsystems

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

Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

International Nuclear Information System (INIS)

Kawashima, H.; Sakurai, S.; Shimizu, K.; Takizuka, T.; Tamai, H.; Matsukawa, M.; Fujita, T.; Miura, Y.

2006-01-01

The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2 . Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design

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.

Experimental and calculating study on the stressed state of superconducting coils of toroidal field in the T-15 tokamak

International Nuclear Information System (INIS)

Vaulina, I.G.; Gusev, S.V.; Sivkova, G.N.

1987-01-01

Results of calculational and experimental atudy of stress-deformed state of superconducting coils of the T-15 tokamak toroidal field are presented. The calculations are made using the method of finite elements and refined theory of cores. Experimental studies were carried out using elastic tensometric model of polymer materials. Test results are compared with the calculational results. Divergence between calculational and experimental values of displacement of characteristic points in the unit does not exceed 20 %. Results of model studies confirm the expediency of the calculational model used for designing SOTP unit for the T-15 tokamak

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

Loss less real-time data compression based on LZO for steady-state Tokamak DAS

International Nuclear Information System (INIS)

Pujara, H.D.; Sharma, Manika

2008-01-01

The evolution of data acquisition system (DAS) for steady-state operation of Tokamak has been technology driven. Steady-state Tokamak demands a data acquisition system which is capable enough to acquire data losslessly from diagnostics. The needs of loss less continuous acquisition have a significant effect on data storage and takes up a greater portion of any data acquisition systems. Another basic need of steady state of nature of operation demands online viewing of data which loads the LAN significantly. So there is strong demand for something that would control the expansion of both these portion by a way of employing compression technique in real time. This paper presents a data acquisition systems employing real-time data compression technique based on LZO. It is a data compression library which is suitable for data compression and decompression in real time. The algorithm used favours speed over compression ratio. The system has been rigged up based on PXI bus and dual buffer mode architecture is implemented for loss less acquisition. The acquired buffer is compressed in real time and streamed to network and hard disk for storage. Observed performance of measure on various data type like binary, integer float, types of different type of wave form as well as compression timing overheads has been presented in the paper. Various software modules for real-time acquiring, online viewing of data on network nodes have been developed in LabWindows/CVI based on client server architecture

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.

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

INITIAL SINGLE-SHELL TANK (SST) SYSTEM PERFORMANCE ASSESSMENT OF THE HANFORD SITE

International Nuclear Information System (INIS)

JARAYSI, M.N.

2007-01-01

The ''Initial Single-Shell Tank System Performance Assessment for the Hanford Site [1] (SST PA) presents the analysis of the long-term impacts of residual wastes assumed to remain after retrieval of tank waste and closure of the SST farms at the US Department of Energy (DOE) Hanford Site. The SST PA supports key elements of the closure process agreed upon in 2004 by DOE, the Washington State Department of Ecology (Ecology), and the US Environmental Protection Agency (EPA). The SST PA element is defined in Appendix I of the ''Hanford Federal Facility Agreement and Consent Order'' (HFFACO) (Ecology et al. 1989) [2], the document that establishes the overall closure process for the SST and double-shell tank (DST) systems. The approach incorporated in the SST PA integrates substantive features of both hazardous and radioactive waste management regulations into a single analysis. The defense-in-depth approach used in this analysis defined two major engineering barriers (a surface barrier and the grouted tank structure) and one natural barrier (the vadose zone) that will be relied on to control waste release into the accessible environment and attain expected performance metrics. The analysis evaluates specific barrier characteristics and other site features that influence contaminant migration by the various pathways. A ''reference'' case and a suite of sensitivity/uncertainty cases are considered. The ''reference case'' evaluates environmental impacts assuming central tendency estimates of site conditions. ''Reference'' case analysis results show residual tank waste impacts on nearby groundwater, air resources; or inadvertent intruders to be well below most important performance objectives. Conversely, past releases to the soil, from previous tank farm operations, are shown to have groundwater impacts that re significantly above most performance objectives. Sensitivity/uncertainty cases examine single and multiple parameter variability along with plausible alternatives

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

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 component failure rates for pulsed versus steady state tokamak operation

International Nuclear Information System (INIS)

Cadwallader, L.C.

1992-07-01

This report presents component failure rate data sources applicable to magnetic fusion systems, and defines multiplicative factors to adjust these data for specific use on magnetic fusion experiment designs. The multipliers address both long pulse and steady state tokamak operation. Thermal fatigue and radiation damage are among the leading reasons for large multiplier values in pulsed operation applications. Field failure rate values for graphite protective tiles are presented, and beryllium tile failure rates in laboratory testing are also given. All of these data can be used for reliability studies, safety analyses, design tradeoff studies, and risk assessments

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

Recent progress on the Compact Ignition Tokamak (CIT)

International Nuclear Information System (INIS)

Ignat, D.W.

1987-01-01

This report describes work done on the Compact Ignition Tokamak (CIT), both at the Princeton Plasma Physics Laboratory (PPPL) and at other fusion laboratories in the United States. The goal of CIT is to reach ignition in a tokamak fusion device in the mid-1990's. Scientific and engineering features of the design are described, as well as projected cost and schedule

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

ACHIEVING AND SUSTAINING STEADY-STATE ADVANCED TOKAMAK CONDITIONS ON DIII-D

International Nuclear Information System (INIS)

WADE, MR; MURAKAMI, M; BRENNAN, DP; CASPER, TA; FERRON, JR; GAROFALO, AM; GREENFIELD, CM; HYATT, AW; JAYAKUMAR, R; KINSEY, JE; LAHAYE, RJ; LAO, LL; LAZARUS, EA; LOHR, J; LUCE, TC; PETTY, CC; POLITZER, PA; PRATER, R; STRAIT, EJ; TURNBULL, AD; WATKINS, JG; WEST, WP

2002-01-01

Recent experiments on the DIII-D tokamak have demonstrated the feasibility of sustaining advanced tokamak conditions that combine high fusion power density (β > 4%), high bootstrap current fraction (f BS ∼ 65%), and high non-inductive current fractions (f NI ∼ 85%) for several energy confinement times. The duration of such conditions is limited only by resistive relaxation of the current density profile. Modeling studies indicate that the application of off-axis ECCD will be able to maintain a favorable current density profile for several seconds

Achieving and sustaining steady-state advanced tokamak conditions on DIII-D

International Nuclear Information System (INIS)

Wade, M.R.; Murakami, M.; Brennan, D.P.

2003-01-01

Recent experiments on the DIII-D tokamak have demonstrated the feasibility of sustaining advanced tokamak conditions that combine high fusion power density (β > 4%), high bootstrap current fraction (f BS ∼ 65%), and high non-inductive current fractions (f NI ∼85%) for several energy confinement times. The duration of such conditions is limited only by resistive relaxation of the current density profile. Modeling studies indicate that the application of off-axis ECCD will be able to maintain a favorable current density profile for several seconds. (author)

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)

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).

Tokamak Physics Experiment (TPX) power supply design and development

International Nuclear Information System (INIS)

Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.

1995-01-01

The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes

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...

Control strategy for plasma equilibrium in a tokamak

International Nuclear Information System (INIS)

Miskell, R.V.

1975-01-01

The dynamic control of the plasma position within the torus of a Tokamak fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. The model considers eddy currents in the conducting shell surrounding the torus and the classical Shafranov equilibrium equation. The equations necessary to characterize the operating conditions of a TOKAMAK are cast in state variable form. Two control variables are selected, the vertical field current and the plasma temperature. The figure of merit chosen minimizes the shift of the plasma within the torus and considers position perturbations necessary to maintain the dense and hotter portions of the plasma profile in the center of the torus, i.e., overcome uneven poloidal fields due to the toroidal geometry. The model uses a Kalman filter to estimate unmeasured state variables, and uses the second variation of the calculus of variations to maintain an optimal control path. (Diss. Abstr. Int., B)

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.

The prognostic value of sST2 and galectin-3 considering different aetiologies in non-ischaemic heart failure.

Science.gov (United States)

Binas, David; Daniel, Hanna; Richter, Anette; Ruppert, Volker; Schlüter, Klaus-Dieter; Schieffer, Bernhard; Pankuweit, Sabine

2018-01-01

Several studies indicate a prognostic value of sST2 and galectin-3 in heart failure (HF). While previous studies focused on ischaemic cause of HF, we investigated the role of sST2 and galectin-3 in patients with non-ischaemic dilated cardiomyopathy (DCM). sST2 and galectin-3 serum concentrations were measured in 262 subjects with DCM. Survival rates were determined for all-cause mortality (ACM) and cardiac mortality (CM). In a univariate model, sST2 as a continuous variable was a predictor of ACM (HR 1.05; 95% CI 1.03 to 1.07, PACM (HR 1.10; 95% CI 1.05 to 1.17, PACM (HR 1.04; 95% CI 1.01 to 1.07, P=0.019) was significant. In a multivariate model, the prognostic value of the sST2 main group remained intact for ACM (HR 1.04; 95% CI 1.02 to 1.07, P=0.003).Univariate and multivariate analysis of galectin-3 as continuous variable did not show any significant result. However, in a quartile model, intermediate values of galectin-3 were significantly associated with a lower event rate of ACM and CM. The study revealed that sST2 predicts ACM and CM in patients with non-ischaemic HF and could be useful especially in patients with inflammatory background. Our findings that intermediate levels of galectin-3 allow for better prognosis were new and different to other investigations. NCT03090425; Results.

Ballooning stable high beta tokamak equilibria

International Nuclear Information System (INIS)

Tuda, Takashi; Azumi, Masafumi; Kurita, Gen-ichi; Takizuka, Tomonori; Takeda, Tatsuoki

1981-04-01

The second stable regime of ballooning modes is numerically studied by using the two-dimensional tokamak transport code with the ballooning stability code. Using the simple FCT heating scheme, we find that the plasma can locally enter this second stable regime. And we obtained equilibria with fairly high beta (β -- 23%) stable against ballooning modes in a whole plasma region, by taking into account of finite thermal diffusion due to unstable ballooning modes. These results show that a tokamak fusion reactor can operate in a high beta state, which is economically favourable. (author)

Role of Transient Receptor Potential Ankyrin 1 Ion Channel and Somatostatin sst4 Receptor in the Antinociceptive and Anti-inflammatory Effects of Sodium Polysulfide and Dimethyl Trisulfide

Directory of Open Access Journals (Sweden)

István Z. Bátai

2018-02-01

Full Text Available Transient receptor potential ankyrin 1 (TRPA1 non-selective ligand-gated cation channels are mostly expressed in primary sensory neurons. Polysulfides (POLYs are Janus-faced substances interacting with numerous target proteins and associated with both protective and detrimental processes. Activation of TRPA1 in sensory neurons, consequent somatostatin (SOM liberation and action on sst4 receptors have recently emerged as mediators of the antinociceptive effect of organic trisulfide dimethyl trisulfide (DMTS. In the frame of the present study, we set out to compare the participation of this mechanism in antinociceptive and anti-inflammatory effects of inorganic sodium POLY and DMTS in carrageenan-evoked hind-paw inflammation. Inflammation of murine hind paws was induced by intraplantar injection of carrageenan (3% in 30 µL saline. Animals were treated intraperitoneally with POLY (17 µmol/kg or DMTS (250 µmol/kg or their respective vehicles 30 min prior paw challenge and six times afterward every 60 min. Mechanical pain threshold and swelling of the paws were measured by dynamic plantar aesthesiometry and plethysmometry at 2, 4, and 6 h after initiation of inflammation. Myeloperoxidase (MPO activity in the hind paws were detected 6 h after challenge by luminescent imaging. Mice genetically lacking TRPA1 ion channels, sst4 receptors and their wild-type counterparts were used to examine the participation of these proteins in POLY and DMTS effects. POLY counteracted carrageenan-evoked mechanical hyperalgesia in a TRPA1 and sst4 receptor-dependent manner. POLY did not influence paw swelling and MPO activity. DMTS ameliorated all examined inflammatory parameters. Mitigation of mechanical hyperalgesia and paw swelling by DMTS were mediated through sst4 receptors. These effects were present in TRPA1 knockout animals, too. DMTS inhibited MPO activity with no participation of the sensory neuron–SOM axis. While antinociceptive effects of

Effect of recycling in the HL-1M tokamak

International Nuclear Information System (INIS)

Zheng Yongzhen

2004-01-01

Tokamak plasma discharge disruption at high density is investigated. The instability analysis on model indicates that the disruption is resulted from the energy loss arising from hydrogen recycling on the edge of the plasma. This energy loss could lead to a contraction of the current channel and the production of a disruptively unstable configuration. Using a simple model we shall investigate the implications of recycling for disruptions. The critical high-density n≤1.6 x 10 20 m -3 is reached in HL-1M. (author)

Plasma density determination by microwave interferometry. The 2 mm interferometer of the TJ-1 tokamak

International Nuclear Information System (INIS)

Manero, F.; Martin, R.

1984-01-01

In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density profiles from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (author)

Plasma density determination by microwave interferometry .- The 2 mm interferometer of the TJ-1 Tokamak

International Nuclear Information System (INIS)

Martin, R.; Manero, F.

1984-01-01

In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs

Tokamak engineering mechanics

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Weiyue; Du, Shijun

2014-01-01

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

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

Electronic density measurement in the TB R-1 tokamak using Faraday rotation

International Nuclear Information System (INIS)

Elizondo, Juan Iraburu

1996-01-01

In this work, the experimental results of electronic density measurements in the TBR-1 tokamak, obtained by Faraday rotation of a microwave beam, are presented, The beam (65 GHz, 500 MW) is generated by a Klystron and crosses the plasma in the horizontal plane. The density values obtained are in agreement with the measurements of a conventional microwave interferometer. As a result of numerical simulations and measurements, it can be concluded that it would be advisable the use of lower wavelengths, to minimize the beam refraction when it crosses the plasma. The results show the feasibility of the Faraday rotation method for density measurement, in the first experiment performed in a tokamak, for the geometry considered. (author)

Plasma density determination by microwave interferometry .- The 2 mm interferometer of the TJ-1 Tokamak; Determinacion de la densidad de un plasma por interferometria de microondas. El interferometro de 2 mm del Tokamak TJ-1

Energy Technology Data Exchange (ETDEWEB)

Martin, R; Manero, F

1984-07-01

In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs.

Radial profiles of hard X-ray emission during steady state current drive in the TRIAM-1M tokamak

International Nuclear Information System (INIS)

Nakamura, Y.; Takabatake, Y.; Jotaki, E.; Moriyama, S.; Nagao, A.; Nakamura, K.; Hiraki, N.; Itoh, S.

1990-01-01

The hard X-ray emission from the TRIAM-1M tokamak plasma during steady state lower hybrid current drive with a discharge duration of a few minutes was measured with sodium iodide scintillation spectrometers. The radial profiles of the X-ray emission were also measured and indicate that, in the low density regime (n e =(1-3)x10 12 cm -3 ), the current carrying high energy electrons are mainly in the inner region of the plasma column and their radial profile remains unchanged during current drive. On the other hand, high density discharges (n e =(3-6)x10 12 cm -3 ) are always accompanied by an abrupt drop of the plasma current, and the X-ray emission profile changes from peaked to broad. This change can be attributed to the conditions of wave accessibility. As the electron density increases, the accessibility of the plasma to lower hybrid waves with low values of the parallel wave number n parallel is significantly reduced and high energy electrons resonating with the waves are produced at the plasma periphery. Interaction of these electrons with the limiters causes an increase of the electron density in this region; waves with low n parallel then become completely excluded from the inner part of the plasma column. This interpretation is supported by measurements of the density profile and impurity radiation, and has been confirmed in an investigation of discharges with additional gas puffing. (author). 17 refs, 21 figs

Tokamak concept innovations

International Nuclear Information System (INIS)

1986-04-01

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

Who will save the tokamak - Harry Potter, Arnold Schwarzenegger, or Shaquille O'Neil?

Science.gov (United States)

Freidberg, J.; Mangiarotti, F.; Minervini, J.

2014-10-01

The tokamak is the current leading contender for a fusion power reactor. The reason for the preeminence of the tokamak is its high quality plasma physics performance relative to other concepts. Even so, it is well known that the tokamak must still overcome two basic physics challenges before becoming viable as a DEMO and ultimately a reactor: (1) the achievement of non-inductive steady state operation, and (2) the achievement of robust disruption free operation. These are in addition to the PMI problems faced by all concepts. The work presented here demonstrates by means of a simple but highly credible analytic calculation that a ``standard'' tokamak cannot lead to a reactor - it is just not possible to simultaneously satisfy all the plasma physics plus engineering constraints. Three possible solutions, some more well-known than others, to the problem are analyzed. These visual image generating solutions are defined as (1) the Harry Potter solution, (2) the Arnold Schwarzenegger solution, and (3) the Shaquille O'Neil solution. Each solution will be described both qualitatively and quantitatively at the meeting.

The spherical tokamak fusion power plant

International Nuclear Information System (INIS)

Wilson, H.R.; Voss, G.; Ahn, J.W.

2003-01-01

The design of a 1GW(e) steady state fusion power plant, based on the spherical tokamak concept, has been further iterated towards a fully self-consistent solution taking account of plasma physics, engineering and neutronics constraints. In particular a plausible solution to exhaust handling is proposed and the steam cycle refined to further improve efficiency. The physics design takes full account of confinement, MHD stability and steady state current drive. It is proposed that such a design may offer a fusion power plant which is easy to maintain: an attractive feature for the power plants following ITER. (author)

Mercier criterion for high-β tokamaks

International Nuclear Information System (INIS)

Galvao, R.M.O.

1984-01-01

An expression, for the application of the Mercier criterion to numerical studies of diffuse high-β tokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the high-β tokamak approximation is derived. (L.C.) [pt

Feedback control of plasma position in the HL-1 tokamak

International Nuclear Information System (INIS)

Yuan Baoshan; Jiao Boliang; Yang Kailing

1991-01-01

In the HL-1 tokamak with a thick copper shell, the control of plasma position is successfully performed by a feedback-feedforward system with dual mode regulator and the equilibrium field coils outside the shell. The plasma position can be controlled within ±2 mm in both vertical and horizontal directions under the condition that the iron core of transformer is not saturated

Magnetic evaluation of hydrogen pressures changes on MHD fluctuations in IR-T1 tokamak plasma

Science.gov (United States)

Alipour, Ramin; Ghanbari, Mohamad R.

2018-04-01

Identification of tokamak plasma parameters and investigation on the effects of each parameter on the plasma characteristics is important for the better understanding of magnetohydrodynamic (MHD) activities in the tokamak plasma. The effect of different hydrogen pressures of 1.9, 2.5 and 2.9 Torr on MHD fluctuations of the IR-T1 tokamak plasma was investigated by using of 12 Mirnov coils, singular value decomposition and wavelet analysis. The parameters such as plasma current, loop voltage, power spectrum density, energy percent of poloidal modes, dominant spatial structures and temporal structures of poloidal modes at different plasma pressures are plotted. The results indicate that the MHD activities at the pressure of 2.5 Torr are less than them at other pressures. It also has been shown that in the stable area of plasma and at the pressure of 2.5 Torr, the magnetic force and the force of plasma pressure are in balance with each other and the MHD activities are at their lowest level.

Overview on the progress of tokamak experimental research in China

International Nuclear Information System (INIS)

Xie Jikang . E-mail; Liu Yong; Wen Yizhi; Wang Long

2001-01-01

Tokamak experimental research in China has made important progress. The main efforts were related to quasi-steady-state operation, LHCD, plasma heating with ICRF, IBW, NBI and ECRH, fuelling with pellets and supersonic molecular beams, and first wall conditioning techniques. Plasma parameters in the experiments were much improved, for example n e =8x10 19 m -3 and a plasma pulse length of >10 s were achieved. ICRF boronization and conditioning resulted in Z eff close to unity. Steady state full LH wave current drive has been achieved for more than 3 s. LHCD ramp-up and recharge have also been demonstrated. The best η CD exp ∼0.5(1+0.085exp(4.8(B T -1.45)))n e I CD R p /P LH =10 19 m -2 A W -1 . Quasi-steady-state H-mode-like plasmas with a density close to the Greenwald limit were obtained by LHCD, where the energy confinement time was nearly five times longer than in the ohmic case. The synergy between IBW, pellet and LHCD was tested. Research on the mechanism of macroturbulence has been extensively carried out experimentally. AC tokamak operation has been successfully demonstrated. (author)

Design constraints for rf-driven steady-state tokamak reactors

International Nuclear Information System (INIS)

Ehst, D.A.

1979-02-01

Plasma current density profiles are computed due to electron Landau damping of lower hybrid waves launched into model tokamak density and temperature profiles. The total current and current profile shape are chosen consistent with magnetohydrodynamic equilibrium for a variety of temperature and density distributions and plasma beta values. Surface current equilibria appear attractive and are accessible to waves with n/sub z/ as low as 1.2. By suitably choosing the spectrum location and width it is possible to drive the 9.8 MA current of a 7.0-m reactor with as little as 2.8% of the fusion power recirculated as rf input from the waveguides

Maximum entropy tokamak configurations

International Nuclear Information System (INIS)

Minardi, E.

1989-01-01

The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

Reclosing of field lines and disruptive instability in tokamaks

International Nuclear Information System (INIS)

Kadomtsev, B.B.

The mechanism of field line reclosing is proposed as the most natural explanation of disruptive instability in tokamaks. This mechanism adequately accounts for the internal disruptive instability, assuming that only mode m = 1 develops. It is extended to the presence of two or several modes. When there is a large number of allowed modes, one can speak of free reclosing, which leads to a force-free magnetic field in a diffusion discharge. In a tokamak, B/sub Z/ much greater than B/sub theta/, free reclosing leads to a uniform distribution of the current over the column cross section and to ejection of part of the poloidal flux beyond the confines of the diaphragm. It may be stated that the disruptive instability in a tokamak is an MHD activity that flares up for a short time and is permanently present in a diffusion column. The geometry of magnetic surfaces during reclosing has been analyzed, and qualitative arguments are given to show that disruptive instability begins to develop as a result of the interaction of the mode m = 2 with the inner mode m = 1

Reclosing of field lines and disruptive instability in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Kadomtsev, B. B.

1976-07-01

The mechanism of field line reclosing is proposed as the most natural explanation of disruptive instability in tokamaks. This mechanism adequately accounts for the internal disruptive instability, assuming that only mode m = 1 develops. It is extended to the presence of two or several modes. When there is a large number of allowed modes, one can speak of free reclosing, which leads to a force-free magnetic field in a diffusion discharge. In a tokamak, B/sub Z/ much greater than B/sub theta/, free reclosing leads to a uniform distribution of the current over the column cross section and to ejection of part of the poloidal flux beyond the confines of the diaphragm. It may be stated that the disruptive instability in a tokamak is an MHD activity that flares up for a short time and is permanently present in a diffusion column. The geometry of magnetic surfaces during reclosing has been analyzed, and qualitative arguments are given to show that disruptive instability begins to develop as a result of the interaction of the mode m = 2 with the inner mode m = 1.

Magnetohydrodynamic helical structures in nominally axisymmetric low-shear tokamak plasmas

International Nuclear Information System (INIS)

Graves, J P; Brunetti, D; Cooper, W A; Reimerdes, H; Halpern, F; Pochelon, A; Sauter, O; Chapman, I T

2013-01-01

The primary goal of hybrid scenarios in tokamaks is to enable high performance operation with large plasma currents whilst avoiding MHD instabilities. However, if a local minimum in the safety factor is allowed to approach unity, the energy required to overcome stabilizing magnetic field line bending is very small, and as a consequence, large MHD structures can be created, with typically dominant m = n = 1 helical component. If there is no exact q = 1 rational surface the essential character of these modes can be modelled assuming ideal nested magnetic flux surfaces. The methods used to characterize these structures include linear and non-linear ideal MHD stability calculations which evaluate the departure from an axisymmetric plasma state, and also equilibrium calculations using a 3D equilibrium code. While these approaches agree favourably for simulations of ITER relevant hybrid regimes in this paper, the relevance of the ideal MHD model itself is tested through empirical examination of helical states in MAST and TCV. While long lived modes in MAST do not have island structures, some of the continuous mode oscillations exhibited in high elongation experiments in TCV indicate that resistivity may play a role in further weakening the ability of the tokamak core to remain axisymmetric. The simulations and experiments consistently highlight the need to control the safety factor in hybrid scenarios planned for future fusion grade tokamaks such as ITER. (paper)

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

Tasmaloum, Vanuatu Fossil Coral SST Reconstruction for 4200 YrBP

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Vanuatu Fossil Coral SST Reconstruction, 4200 YrBP. Paleobathymetry estimated at 10-15 m. SST reconstruction is based on the analysis of Sr/Ca and U/Ca ratios in...

Current drive by spheromak injection into a tokamak

International Nuclear Information System (INIS)

Brown, M.R.; Bellan, P.M.

1990-01-01

The authors report the first observation of current drive by spheromak injection into a tokamak due to the process of helicity injection. Current drive is observed in Caltech's ENCORE tokamak (30% increase, ΔI > 1 kA) only when both the tokamak and injected spheromak have the same sign of helicity (where helicity is defined as positive if current flows parallel to magnetic field lines and negative if anti-parallel). The initial increase (decrease) in current is accompanied by a sharp decrease (increase) in loop voltage and the increase in tokamak helicity is consistent with the helicity content of the injected spheromak. In addition, the injection of the spheromak raises the tokamak central density by a factor of six. The introduction of cold spheromak plasma causes sudden cooling of the tokamak discharge from 12 eV to 4 eV which results in a gradual decline in tokamak plasma current by a factor of three. In a second experiment, the authors inject spheromaks into the magnetized toroidal vacuum vessel (with no tokamak plasma). An m = 1 magnetic structure forms in the vessel after the spheromak undergoes a double tilt; once in the cylindrical entrance between gun and tokamak, then again in the tokamak vessel. A horizontal shift of the spheromak equilibrium is observed in the direction opposite that of the static toroidal field. In the absence of net toroidal flux, the structure develops a helical pitch as predicted by theory. Experiments with a number of refractory metal coatings have shown that tungsten and chrome coatings provide some improvement in spheromak parameters. They have also designed and will soon construct a larger, higher current spheromak gun with a new accelerator section for injection experiments on the Phaedrus-T tokamak

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Advances in multi-megawatt lower hybrid technology in support of steady-state tokamak operation

Science.gov (United States)

Delpech, L.; Achard, J.; Armitano, A.; Artaud, J. F.; Bae, Y. S.; Belo, J. H.; Berger-By, G.; Bouquey, F.; Cho, M. H.; Corbel, E.; Decker, J.; Do, H.; Dumont, R.; Ekedahl, A.; Garibaldi, P.; Goniche, M.; Guilhem, D.; Hillairet, J.; Hoang, G. T.; Kim, H. S.; Kim, J. H.; Kim, H.; Kwak, J. G.; Magne, R.; Mollard, P.; Na, Y. S.; Namkung, W.; Oh, Y. K.; Park, S.; Park, H.; Peysson, Y.; Poli, S.; Prou, M.; Samaille, F.; Yang, H. L.; The Tore Supra Team

2014-10-01

It has been demonstrated that lower hybrid current drive (LHCD) systems play a crucial role for steady-state tokamak operation, owing to their high current drive (CD) efficiency and hence their capability to reduce flux consumption. This paper describes the extensive technology programmes developed for the Tore Supra (France) and the KSTAR (Korea) tokamaks in order to bring continuous wave (CW) LHCD systems into operation. The Tore Supra LHCD generator at 3.7 GHz is fully CW compatible, with RF power PRF = 9.2 MW available at the generator to feed two actively water-cooled launchers. On Tore Supra, the most recent and novel passive active multijunction (PAM) launcher has sustained 2.7 MW (corresponding to its design value of 25 MW m-2 at the launcher mouth) for a 78 s flat-top discharge, with low reflected power even at large plasma-launcher gaps. The fully active multijunction (FAM) launcher has reached 3.8 MW of coupled power (24 MW m-2 at the launcher mouth) with the new TH2103C klystrons. By combining both the PAM and FAM launchers, 950 MJ of energy, using 5.2 MW of LHCD and 1 MW of ICRH (ion cyclotron resonance heating), was injected for 160 s in 2011. The 3.7 GHz CW LHCD system will be a key element within the W (for tungsten) environment in steady-state Tokamak (WEST) project, where the aim is to test ITER technologies for high heat flux components in relevant heat flux density and particle fluence conditions. On KSTAR, a 2 MW LHCD system operating at 5 GHz is under development. Recently the 5 GHz prototype klystron has reached 500 kW/600 s on a matched load, and studies are ongoing to design a PAM launcher. In addition to the studies of technology, a combination of ray-tracing and Fokker-Planck calculations have been performed to evaluate the driven current and the power deposition due to LH waves, and to optimize the N∥ spectrum for the future launcher design. Furthermore, an LHCD system at 5 GHz is being considered for a future upgrade of the ITER

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)

Internal m=1, n=1 helical mode in a tokamak with nonmonotonic current profile

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Mikhajlovskij, A.B.

1988-01-01

Internal helical mode in a tokamak with two resonance surfaces, on which storing coefficient reduces to unity is studied theoretically. A general criterion for the investigated perturbations stability is obtained. Dispersion equation, describing both ideal and resistive helical modes, is derived. Analytic calculations for the case of perturbations localized near the tokamak axis are made. It is shown that in the framework of standard ideal hydrodynamics such perturbations are unstable at characteristic nonmonotonous profiles of the current

The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1989-10-01

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

Magnetic Fusion Energy Program of India

International Nuclear Information System (INIS)

Sen, Abhijit

2013-01-01

The magnetic fusion energy program of India started in the early eighties with the construction of an indigenous tokamak device ADITYA at the Institute for Plasma Research in Gandhinagar. The initial thrust was on fundamental studies related to plasma instabilities and turbulence phenomena but there was also a significant emphasis on technology development in the areas of magnetics, high vacuum, radio-frequency heating and neutral beam technology. The program took a major leap forward in the late nineties with the decision to build a state-of-the-art superconducting tokamak (SST-1) that catapulted India into the mainstream of the international tokamak research effort. The SST experience and the associated technological and human resource development has now earned the country a place in the ITER collaboration as an equal partner with other major nations. Keeping in mind the rapidly growing and enormous energy needs of the future the program has also identified and launched key development projects that can lead us to a DEMO reactor and eventually a Fusion Power Plant in a systematic manner. I will give a brief overview of the early origins, the present status and some of the highlights of the future road map of the Indian Fusion Program. (author)

Plea for stellarator funding raps tokamaks

International Nuclear Information System (INIS)

Blake, M.

1992-01-01

The funding crunch in magnetic confinement fusion development has moved the editor of a largely technical publication to speak out on a policy issue. James A. Rome, who edits Stellarator News from the Fusion Energy Division at Oak Ridge National Laboratory, wrote an editorial that appeared on the front page of the May 1992 issue. It was titled open-quotes The US Stellarator Program: A Time for Renewal,close quotes and while it focused chiefly on that subject (and lamented the lack of funding for the operation of the existing ATF stellarator at Oak Ridge), it also cited some of the problems inherent in the mainline MCF approach--the tokamak--and stated that if the money can be found for further tokamak design upgrades, it should also be found for stellarators. Rome wrote, open-quotes There is growing recognition in the US, and elsewhere, that the conventional tokamak does not extrapolate to a commercially competitive energy source except with very high field coils ( 1000 MWe).close quotes He pointed up open-quotes the difficulty of simultaneously satisfying conflicting tokamak requirements for efficient current drive, high bootstrap-current fraction, complete avoidance of disruptions, adequate beta limits, and edge-plasma properties compatible with improved (H-mode) confinement and acceptable erosion of divertor plates.close quotes He then called for support for the stellarator as open-quotes the only concept that has performance comparable to that achieved in tokamaks without the plasma-current-related limitations listed above.close quotes

Tokamak devices: towards controlled fusion

International Nuclear Information System (INIS)

Trocheris, M.

1975-01-01

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

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.

Asymmetric Response of the Equatorial Pacific SST to Climate Warming and Cooling

Energy Technology Data Exchange (ETDEWEB)

Liu, Fukai [Physical Oceanography Laboratory/Qingdao Collaborative Innovation Center of Marine Science and Technology, Ocean University of China, and Qingdao National Laboratory for Marine Science and Technology, Qingdao, China; Luo, Yiyong [Physical Oceanography Laboratory/Qingdao Collaborative Innovation Center of Marine Science and Technology, Ocean University of China, and Qingdao National Laboratory for Marine Science and Technology, Qingdao, China; Lu, Jian [Atmospheric Sciences and Global Change Division, Pacific Northwest National Laboratory, Richland, Washington; Garuba, Oluwayemi [Atmospheric Sciences and Global Change Division, Pacific Northwest National Laboratory, Richland, Washington; Wan, Xiuquan [Physical Oceanography Laboratory/Qingdao Collaborative Innovation Center of Marine Science and Technology, Ocean University of China, and Qingdao National Laboratory for Marine Science and Technology, Qingdao, China

2017-09-01

The response of the equatorial Pacific Ocean to heat fluxes of equal amplitude but opposite sign is investigated using the Community Earth System Model (CESM). Results show a strong asymmetry in SST changes. In the eastern equatorial Pacific (EEP), the warming responding to the positive forcing exceeds the cooling to the negative forcing; while in the western equatorial Pacific (WEP), it is the other way around and the cooling surpasses the warming. This leads to a zonal dipole asymmetric structure, with positive values in the east and negative values in the west. A surface heat budget analysis suggests that the SST asymmetry is mainly resulted from the oceanic horizontal advection and vertical entrainment, with both of their linear and nonlinear components playing a role. For the linear component, its change appears to be more significant over the EEP (WEP) in the positive (negative) forcing scenario, favoring the seesaw pattern of the SST asymmetry. For the nonlinear component, its change acts to warm (cool) the EEP (WEP) in both scenarios, also favorable for the development of the SST asymmetry. Additional experiments with a slab ocean confirm the dominant role of ocean dynamical processes for this SST asymmetry. The net surface heat flux, in contrast, works to reduce the SST asymmetry through its shortwave radiation and latent heat flux components, with the former being related to the nonlinear relationship between SST and convection, and the latter being attributable to Newtonian damping and air-sea stability effects. The suppressing effect of shortwave radiation on SST asymmetry is further verified by partially coupled overriding experiments.

Plasma control issues for an advanced steady state tokamak reactor

International Nuclear Information System (INIS)

Moreau, D.

2001-01-01

This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

Plasma control issues for an advanced steady state tokamak reactor

International Nuclear Information System (INIS)

Moreau, D.; Voitsekhovitch, I.

1999-01-01

This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

Interannual to Decadal SST Variability in the Tropical Indian Ocean

Science.gov (United States)

Wang, G.; Newman, M.; Han, W.

2017-12-01

The Indian Ocean has received increasing attention in recent years for its large impacts on regional and global climate. However, due mainly to the close interdependence of the climate variation within the Tropical Pacific and the Indian Ocean, the internal sea surface temperature (SST) variability within the Indian Ocean has not been studied extensively on longer time scales. In this presentation we will show analysis of the interannual to decadal SST variability in the Tropical Indian Ocean in observations and Linear Inverse Model (LIM) results. We also compare the decoupled Indian Ocean SST variability from the Pacific against fully coupled one based on LIM integrations, to test the factors influence the features of the leading SST modes in the Indian Ocean. The result shows the Indian Ocean Basin (IOB) mode, which is strongly related to global averaged SST variability, passively responses to the Pacific variation. Without tropical Indo-Pacific coupling interaction, the intensity of IOB significantly decreases by 80%. The Indian Ocean Dipole (IOD) mode demonstrates its independence from the Pacific SST variability since the IOD does not change its long-term characteristics at all without inter-basin interactions. The overall SSTA variance decreases significantly in the Tropical Indian Ocean in the coupling restricted LIM runs, especially when the one-way impact from the Pacific to the Indian Ocean is turned off, suggesting that most of the variability in the Indian Ocean comes from the Pacific influence. On the other hand, the Indian Ocean could also transport anomalies to the Pacific, making the interaction a complete two-way process.

Structure of chaotic magnetic field lines in IR-T1 tokamak due to ergodic magnetic limiter

Science.gov (United States)

Ahmadi, S.; Salar Elahi, A.; Ghorannevis, M.

2018-03-01

In this paper we have studied an Ergodic Magnetic Limiter (EML) based chaotic magnetic field for transport control in the edge plasma of IR-T1 tokamak. The resonance created by the EML causes perturbation of the equilibrium field line in tokamak and as a result, the field lines are chaotic in the vicinity of the dimerized island chains. Transport barriers are formed in the chaotic field line and actually observe in tokamak with reverse magnetic shear. We used area-preserving non-twist (and twist) Poincaré maps to describe the formation of transport barriers, which are actually features of Hamiltonian systems. This transport barrier is useful in reducing radial diffusion of the field line and thus improving the plasma confinement.

Structure of chaotic magnetic field lines in IR-T1 tokamak due to ergodic magnetic limiter

Directory of Open Access Journals (Sweden)

S. Ahmadi

2018-03-01

Full Text Available In this paper we have studied an Ergodic Magnetic Limiter (EML based chaotic magnetic field for transport control in the edge plasma of IR-T1 tokamak. The resonance created by the EML causes perturbation of the equilibrium field line in tokamak and as a result, the field lines are chaotic in the vicinity of the dimerized island chains. Transport barriers are formed in the chaotic field line and actually observe in tokamak with reverse magnetic shear. We used area-preserving non-twist (and twist Poincaré maps to describe the formation of transport barriers, which are actually features of Hamiltonian systems. This transport barrier is useful in reducing radial diffusion of the field line and thus improving the plasma confinement.

Increase in beta limit in tokamak plasmas

International Nuclear Information System (INIS)

Kamada, Yutaka

2003-01-01

This paper reviews recent studies of tokamak MHD stability towards the achievement of a high beta steady-state, where the profile control of current, pressure, and rotation, and the optimization of the plasma shape play fundamental roles. The key instabilities include the neoclassical tearing mode, the resistive wall mode, the edge localized mode, etc. In order to demonstrate an economically attractive tokamak reactor, it is necessary to increase the beta value simultaneously with a sufficiently high integrated plasma performance. Towards this goal, studies of stability control in self-regulating plasma systems are essential. (author)

Physics parameter space of tokamak ignition devices

International Nuclear Information System (INIS)

Selcow, E.C.; Peng, Y.K.M.; Uckan, N.A.; Houlberg, W.A.

1985-01-01

This paper describes the results of a study to explore the physics parameter space of tokamak ignition experiments. A new physics systems code has been developed to perform the study. This code performs a global plasma analysis using steady-state, two-fluid, energy-transport models. In this paper, we discuss the models used in the code and their application to the analysis of compact ignition experiments. 8 refs., 8 figs., 1 tab

Helicity content and tokamak applications of helicity

International Nuclear Information System (INIS)

Boozer, A.H.

1986-05-01

Magnetic helicity is approximately conserved by the turbulence associated with resistive instabilities of plasmas. To generalize the application of the concept of helicity, the helicity content of an arbitrary bounded region of space will be defined. The definition has the virtues that both the helicity content and its time derivative have simple expressions in terms of the poloidal and toroidal magnetic fluxes, the average toroidal loop voltage and the electric potential on the bounding surface, and the volume integral of E-B. The application of the helicity concept to tokamak plasmas is illustrated by a discussion of so-called MHD current drive, an example of a stable tokamak q profile with q less than one in the center, and a discussion of the possibility of a natural steady-state tokamak due to the bootstrap current coupling to tearing instabilities

User's manual for DSTAR MOD1: A comprehensive tokamak disruption code

International Nuclear Information System (INIS)

Merrill, B.J.; Jardin, S.J.

1986-01-01

A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces that can occur during major tokamak plasma disruptions. The DSTAR code development effort has been accomplished by coupling a recently developed free boundary tokamak plasma transport computational model with other models developed to predict impurity transport and radiation, and the electromagnetic and thermal dynamic response of vacuum vessel components. The combined model, DSTAR, is a unique tool for predicting the consequences of tokamak disruptions. This informal report discusses the sequence of events of a resistive disruption, models developed to predict plasma transport and electromagnetic field evolution, the growth of the stochastic region of the plasma, the transport and nonequilibrium ionization/emitted radiation of the ablated vacuum vessel material, the vacuum vessel thermal and magnetic response, and user input and code output

Predictive value of serum sST2 in preschool wheezers for development of asthma with high FeNO.

Science.gov (United States)

Ketelaar, M E; van de Kant, K D; Dijk, F N; Klaassen, E M; Grotenboer, N S; Nawijn, M C; Dompeling, E; Koppelman, G H

2017-11-01

Wheezing is common in childhood. However, current prediction models of pediatric asthma have only modest accuracy. Novel biomarkers and definition of subphenotypes may improve asthma prediction. Interleukin-1-receptor-like-1 (IL1RL1 or ST2) is a well-replicated asthma gene and associates with eosinophilia. We investigated whether serum sST2 predicts asthma and asthma with elevated exhaled NO (FeNO), compared to the commonly used Asthma Prediction Index (API). Using logistic regression modeling, we found that serum sST2 levels in 2-3 years-old wheezers do not predict doctors' diagnosed asthma at age 6 years. Instead, sST2 predicts a subphenotype of asthma characterized by increased levels of FeNO, a marker for eosinophilic airway inflammation. Herein, sST2 improved the predictive value of the API (AUC=0.70, 95% CI 0.56-0.84), but had also significant predictive value on its own (AUC=0.65, 95% CI 0.52-0.79). Our study indicates that sST2 in preschool wheezers has predictive value for the development of eosinophilic airway inflammation in asthmatic children at school age. © 2017 EAACI and John Wiley and Sons A/S. Published by John Wiley and Sons Ltd.

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

The ASTRI SST-2M prototype for the Cherenkov Telescope Array: prototype technologies goals and strategies for the future SST

Science.gov (United States)

Marchiori, Gianpietro; Busatta, Andrea; Giacomel, Stefano; Folla, Ivan; Valsecchi, Marco; Canestrari, Rodolfo; Bonnoli, Giacomo; Cascone, Enrico; Conconi, Paolo; Fiorini, Mauro; Giro, Enrico; La Palombara, Nicola; Pareschi, Giovanni; Perri, Luca; Rodeghiero, Gabriele; Sironi, Giorgia; Stringhetti, Luca; Toso, Giorgio; Tosti, Gino; Pellicciari, Carlo

2014-07-01

The Cherenkov Telescope Array (CTA) observatory will represent the next generation of Imaging Atmospheric Cherenkov Telescope. Using a combination of large-, medium-, and small-scale telescopes (LST, MST, SST, respectively), it will explore the Very High Energy domain from a few tens of GeVup to about few hundreds of TeV with unprecedented sensitivity, angular resolution and imaging quality. In this framework, the Italian ASTRI program, led by the Italian National Institute of Astrophysics (INAF) developed a 4-meter class telescope, which will adopt an aplanatic, wide-field, double-reflection optical layout in a Schwarzschild- Couder configuration. Within this program INAF assigned to the consortium between Galbiati Group and EIE Group the construction, assembly and tests activities of the prototype named ASTRI SST-2M. On the basis of the lesson learnt from the prototype, other telescopes will be produced, starting from a re-design phase, in order to optimize performances and the overall costs and production schedule for the CTA-SST telescope. This paper will firstly give an overview of the concept for the SST prototype mount structure. In this contest, the technologies adopted for the design, manufacturing and tests of the entire system will be presented. Moreover, a specific focus on the challenges of the prototype and the strategies associated with it will be provided, in order to outline the near future performance goals for this type of Cherenkov telescopes employed for Gamma ray science.

Summary of the 1982 small tokamak users meeting

International Nuclear Information System (INIS)

Sprott, J.C.

1982-11-01

On November 1, 1982, the sixth in a series of approximately annual meetings of the users of small tokamaks was held in conjunction with the APS Division of Plasma Physics meeting at New Orleans. The meeting lasted three hours, with 34 people attending. The interest was on strengthening the ties between the small tokamaks and the large tokamaks. Accordingly, the latest meeting was dedicated to this theme, and in contrast to previous meetings, a few representatives from the large tokamaks were invited to attend and make presentations. Summaries of the various talks are included

HL-1 tokamak data acquisition system and its initial application in the physical experiment

International Nuclear Information System (INIS)

Deng Huichen; Fu Bo; Dong Jiafu

1989-11-01

A HL-1 tokamak data acquisition system has been developed and has been used in the physical experiment. The hardware and software configuration of the system, as well as the typical acquired data in the HL-1 experiment are introduced

Improvement of the tokamak concept

Energy Technology Data Exchange (ETDEWEB)

Laurent, L

1994-12-31

Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.

Stability of n = 1 kink modes in bean-shaped tokamaks

International Nuclear Information System (INIS)

Manickam, J.; Grimm, R.C.; Okabayashi, M.

1983-08-01

Numerical studies show that by indenting the small-major-radius side of conventional finite-aspect-ratio tokamaks, significant improvements to the stability of pressure-driven ideal MHD modes can be achieved. The internal n - 1 kink mode can be stabilized completely with quite modest indentation. Kink-ballooning mode stability is also improved, and, in the presence of a nearby conducting wall, accessibility to a second stable region at high plasma β is possible

Plasma position control in TCABR Tokamak

International Nuclear Information System (INIS)

Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

1998-01-01

The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

Fast IR diodes thermometer for tokamak

International Nuclear Information System (INIS)

Chen Xiangbo

2001-01-01

A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices

Wall conditioning of the TBR-1 Tokamak by plasma generated by microwaves

International Nuclear Information System (INIS)

Elizondo, J.I.

1986-01-01

A new system of vaccum chamber wall conditioning in the TBR-1 Tokamak, using electron cyclotron resonance plasma of hydrogen for the discharge cleaning process is presented. The construction and performance of equipments are described, and the cleaning process to otimize the conditioning efficiency by chase of plasma parameters. (author) [pt

First Results from Tests of High Temperature Superconductor Magnets on Tokamak

Energy Technology Data Exchange (ETDEWEB)

Gryaznevich, M.; Todd, T.T., E-mail: mikhail.gryaznevich@ccfe.ac.uk [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Svoboda, V.; Markovic, T.; Ondrej, G. [Czech Technical University, Prague (Czech Republic); Stockel, J.; Duran, I.; Kovarik, K. [IPP Prague, Czech Technical University, Prague (Czech Republic); Sykes, A.; Kingham, D. [Tokamak Solutions, Culham Science Centre, Abingdon (United Kingdom); Melhem, Z.; Ball, S.; Chappell, S. [Oxford Instruments, Abingdon (United Kingdom); Lilley, M. K.; De Grouchy, P.; Kim, H. -T. [Imperial College, London (United Kingdom)

2012-09-15

Full text: It has long been known that high temperature superconductors (HTS) could have an important role to play in the future of tokamak fusion research. Here we report on first results of the use of HTS in a tokamak magnet and on the progress in design and construction of the first fully-HTS tokamak. In the experiment, the two copper vertical field coils of the small tokamak GOLEM were replaced by two coils each with 6 turns of HTS (Re)BCO tape. Liquid nitrogen was used to cool the coils to below the critical temperature at which HTS becomes superconducting. Little effect on the HTS critical current has been observed for perpendicular field up to 0.5 T and superconductivity has been achieved at {approx} 90.5K during bench tests. There had been concerns that the plasma pulses and pulsed magnetic fields might cause a 'quench' in the HTS, i.e., a sudden and potentially damaging transition from superconductor to normal conductor. However, many plasma pulses were fired without any quenches even when disruptions occurred with corresponding induced electrical fields. In addition, experiments without plasma have been performed to study properties of the HTS in a tokamak environment, i.e., critical current and its dependence on magnetic and electrical fields generated in a tokamak both in DC and pulsed operations, maximum current ramp-up speed, performance of the HTS tape after number of artificially induced quenches etc. No quench has been observed at DC currents up to 200 A (1.2 kA-turns through the coil). In short pulses, current up to 1 kA through the tape (6 kA-turns) has been achieved with no subsequent degradation of the HTS performance with a current ramp rate up to 0.6 MA/s. In future experiments, increases in both the plasma current and pulse duration are planned. Considerable experience has been gained during design and fabrication of the cryostat, coils, isolation and insulation, feeds and cryosystems, and GOLEM is now routinely operated with HTS coils. The

Study on plasma visible radiation spatial distribution at the TO-1 tokamak

International Nuclear Information System (INIS)

Molotkov, L.I.; Shvindt, N.N.

1977-01-01

The results of spatial distribution measurements of radiation intensities of spectral lines of hydrogen and light plasma impurities in the visible-light spectrum TO-1 tokamak are described. The method of electrochemical scanning with the help of rotating disk with notches was used. The experiments were carried out in the stable regime of discharge and in the regime with breakdown instability. In the stable regime absolute intensities of the spectral lines were measured. The concentration of radiating atoms and ions were calculated using the measured absolute intensities. Besides in order to determine the region in which the discharge appears in the cross section of the TO-1 chamber the spatial distributions of Hsub(β) neutral hydrogen spectral line in the initial stage of the discharge (0 9 cm -3 for the C(3) ion and anti nsub(0)=1.1x10 10 cm -3 for hydrogen atoms. The investigation into spatial distributions of the Hsub(β) line radiation in the initial stage of the filament formation showed that in the overwhelming majority of cases the discharge appears near the internal wall of the tokamak chamber

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)

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)

Utilize the spectral line pair of the same ionized state ion to measure the ion temperature of tokamak plasma

International Nuclear Information System (INIS)

Lin Xiaodong

2000-01-01

Making use of a Fabry-Perot interferometer driven by a piezoelectric crystal and selecting the suitable separation of plates, the ion temperature is defined by measuring the superimposed profile of the spectral line pair of the same ionized state ions in Tokamak. The advantage of this method is to higher spectral resolution and wider spectral range select

Analysis of tokamak plasma confinement modes using the fast

Indian Academy of Sciences (India)

The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and ...

Confinement of ohmically heated plasmas and turbulent heating in high-magnetic field tokamak TRIAM-1

Energy Technology Data Exchange (ETDEWEB)

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

1979-12-01

TRIAM-1, the tokamak device with high toroidal magnetic field, has been constructed to establish the scaling laws of advanced tokamak devices such as Alcator, and to study the possibility of the turbulent heating as a further economical heating method of the fusion oriented plasmas. The plasma parameters obtained by ohmic heating alone are as follows; central electron temperature T sub(e0) = 640 eV, central ion temperature T sub(i0) = 280 eV and line-average electron density n average sub(e) = 2.2 x 10/sup 14/ cm/sup -3/. The empirical scaling laws are investigated concerning T sub(e0), T sub(i0) and n average sub(e). The turbulent heating has been carried out by applying the high electric field in the toroidal direction to the typical tokamak discharge with T sub(i0) asymptotically equals 200 eV. The efficient ion heating is observed and T sub(i0) attains to about 600 eV.

Criteria for initiation of tokamak disruptions

International Nuclear Information System (INIS)

Hopcraft, K.I.; Turner, M.F.

1986-01-01

The process by which a tokamak plasma evolves from an equilibrium state containing a saturated magnetic island to one which is disruptively unstable is discussed and illustrated by numerical simulation of a resistive magnetoplasma. Those elements which are required to initiate a disruption are delineated

Development of DEMO-FNS tokamak for fusion and hybrid technologies

Science.gov (United States)

Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

2015-07-01

The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

Tokamak engineering mechanics

CERN Document Server

Song, Yuntao; Du, Shijun

2013-01-01

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

The tokamak as a neutron source

International Nuclear Information System (INIS)

Hendel, H.W.; Jassby, D.L.

1989-11-01

This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

Tokamak poloidal-field systems. Progress report, January 1-December 31, 1982

International Nuclear Information System (INIS)

Rogers, J.D.

1983-05-01

The work performed in support of the FED and INTOR tokamak studies is reported at length and covers almost all the aspects of poloidal field (PF) design that were considered. The design work included magnetics, forces and fields, superconductor design, superconductor loss calculations, high field tokamak central solenoid parametric analysis, helium vapor release with bubble clearing and entrainment analysis, eddy current losses in dewars, structural support design for internally cooled cable superconductor (ICCS), research and technology development and manufacturing plans and milestones for poloidal field (PF) coils, fault conditions for shorted PF coils, design of 50 kA vapor cooled leads, and structural design of PF ring coils box frame dewars. Eddy current calculations in tokamak structure are being calculated. A computer code to perform stability analysis of ICCS is being written. Two water cooled switches, a vacuum interrupter and a bypass switch, were tested to develop improved higher current carrying capacities

Observation of electron temperature profile in HL-1M tokamak

International Nuclear Information System (INIS)

Cao Jianyong; Xu Deming; Ding Xuantong

2000-01-01

The principle and method of the electron temperature measurement by means of electron cyclotron emission (ECE) have been described. Several results under different conditions on HL-1M tokamak have been given. The hollow profile of electron temperature appears in some stages, such as current rising, pellet injection and impurity concentration in the plasma centre. When the bias voltage is applied, the electron temperature profile become steeper. All of the phenomena are related with the transport in plasma centre

Heat flux exchange estimation by using ATSR SST data in TOGA area

Science.gov (United States)

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

1995-12-01

The study of phenomena such as ENSO requires consideration of the dynamics and thermodynamics of the coupled ocean-atmosphere system. The dynamic and thermal properties of the atmosphere and ocean are directly affected by air-sea transfers of fluxes of momentum, heat and moisture. In this paper, we present results of turbulent heat fluxes calculated by using two years (1992 and 1993) monthly average TOGA data and ATSR SST data in TOGA area. A comparison with published results indicates good qualitative agreement. Also, we compared the results of heat flux exchange by using ATSR SST data and by using the TOGA bucket SST data. The ATSR SST data set has been shown to be useful in helping to estimate the large space scale heat flux exchange.

Improvement of confinement characteristics of tokamak plasma by controlling plasma-wall interactions

International Nuclear Information System (INIS)

Sengoku, Seio

1985-08-01

Relation between plasma-wall interactions and confinement characteristics of a tokamak plasma with respect to both impurity and fuel particle controls is discussed. Following results are obtained from impurity control studies: (1) Ion sputtering is the dominant mechanism of impurity release in a steady state tokamak discharge. (2) By applying carbon coating on entire first wall of DIVA tokamak, dominant radiative region is concentrated more in boundary plasma resulting a hot peripheral plasma with cold boundary plasma. (3) A physical model of divertor functions about impurity control is empilically obtained. By a computer simulation based on above model with respect to divertor functions for JT-60 tokamak, it is found that the allowable electron temperature of the divertor plasma is not restricted by a condition that the impurity release due to ion sputtering does not increase continuously. (4) Dense and cold divertor plasma accompanied with strong remote radiative cooling was diagnosed along the magnetic field line in the simple poloidal divertor of DOUBLET III tokamak. Strong particle recycling region is found to be localized near the divertor plate. by and from particle control studies: (1) The INTOR scaling on energy confinement time is applicable to high density region when a core plasma is fueled directly by solid deuterium pellet injection in DOUBLET III tokamak. (2) As remarkably demonstrated by direct fueling with pellet injection, energy confinement characteristics can be improved at high density range by decreasing particle deposition at peripheral plasma in order to reduce plasma-wall interaction. (3) If the particle deposition at boundary layer is necessarily reduced, the electron temperature at the boundary or divertor region increases due to decrease of the particle recycling and the electron density there. (J.P.N.)

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.)

An Assessment of the SST Simulation Using the Climate Forecast System Coupled to the SSiB Surface Model

Science.gov (United States)

Wang, Y.; Xue, Y.; Huang, B.; Lee, J.; De Sales, F.

2016-12-01

A long term simulation has been conducted using the Climate Forecast System (CFSv2) coupled to the SSiB-2 land model, which consists of the Global Forecast System atmospheric model (GFS) and the Modular Ocean model - version 4 (MOM4) as the ocean component. This study evaluates the model's performance in simulating sea surface temperature (SST) mean state, trend, and inter-annual and decadal variabilities. The model is able to produce the reasonable spatial distribution of the SST climatology; however, it has prominent large scale biases. In the middle latitude of the Northern Hemisphere, major cold biases is close to the warm side of the large SST gradients, which may be associated with the weaker Kuroshio and Gulf Stream extensions that diffuse the SST gradient. IN addition, warm biases extend along the west coast of the North America continent to the high latitude, which may be related with excessive Ekman down-welling and solar radiation fluxes reaching to the surface due to the lack of cloud there. Warm biases also exist over the tropical cold tough areas in the Pacific and Atlantic. The global SST trend and interannual variations are well captured except for that in the south Hemisphere after year 2000, which is mainly contributed by the bias from the southern Pacific Ocean. Although the model fails to accurately produce ENSO events in proper years, it does reproduce the ENSO frequency well; they are skewed toward more warm events after 1990. The model also shows ability in SST decadal variation, such as the so-called inter-decadal Pacific oscillation (IPO); however, its phases seem to go reversely compared with the observation.

Assessment of Global Forecast Ocean Assimilation Model (FOAM) using new satellite SST data

Science.gov (United States)

Ascione Kenov, Isabella; Sykes, Peter; Fiedler, Emma; McConnell, Niall; Ryan, Andrew; Maksymczuk, Jan

2016-04-01

There is an increased demand for accurate ocean weather information for applications in the field of marine safety and navigation, water quality, offshore commercial operations, monitoring of oil spills and pollutants, among others. The Met Office, UK, provides ocean forecasts to customers from governmental, commercial and ecological sectors using the Global Forecast Ocean Assimilation Model (FOAM), an operational modelling system which covers the global ocean and runs daily, using the NEMO (Nucleus for European Modelling of the Ocean) ocean model with horizontal resolution of 1/4° and 75 vertical levels. The system assimilates salinity and temperature profiles, sea surface temperature (SST), sea surface height (SSH), and sea ice concentration observations on a daily basis. In this study, the FOAM system is updated to assimilate Advanced Microwave Scanning Radiometer 2 (AMSR2) and the Spinning Enhanced Visible and Infrared Imager (SEVIRI) SST data. Model results from one month trials are assessed against observations using verification tools which provide a quantitative description of model performance and error, based on statistical metrics, including mean error, root mean square error (RMSE), correlation coefficient, and Taylor diagrams. A series of hindcast experiments is used to run the FOAM system with AMSR2 and SEVIRI SST data, using a control run for comparison. Results show that all trials perform well on the global ocean and that largest SST mean errors were found in the Southern hemisphere. The geographic distribution of the model error for SST and temperature profiles are discussed using statistical metrics evaluated over sub-regions of the global ocean.

Status of tokamak research

International Nuclear Information System (INIS)

Rawls, J.M.

1979-10-01

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

Impact of global SST gradients on the Mediterranean runoff changes across the Plio-Pleistocene transition

Science.gov (United States)

Colleoni, Florence; Cherchi, Annalisa; Masina, Simona; Brierley, Christopher M.

2015-06-01

This work explores the impact of the development of global meridional and zonal sea surface temperature (SST) gradients on the Mediterranean runoff variability during the Plio-Pleistocene transition, about 3 Ma. Results show that total annual mean Pliocene Mediterranean runoff is about 40% larger than during the preindustrial period due to more increased extratropical specific humidity. As a consequence of a weakened and extended Hadley cell, the Pliocene northwest Africa hydrological network produces a discharge 30 times larger than today. Our results support the conclusion that during the Pliocene, the Mediterranean water deficit was reduced relative to today due to a larger river discharge. By means of a stand-alone atmospheric general circulation model, we simulate the separate impact of extratropical and equatorial SST cooling on the Mediterranean runoff. While cooling the equatorial SST does not imply significant changes to the Pliocene Mediterranean hydrological budget, the extratropical SST cooling increases the water deficit due to a decrease in precipitation and runoff. Consequently, river discharge from this area reduces to preindustrial levels. The main teleconnections acting upon the Mediterranean area today, i.e., the North Atlantic Oscillation during winter and the "monsoon-desert" mechanism during summer already have a large influence on the climate of our Pliocene simulations. Finally, our results also suggest that in a climate state significantly warmer than today, changes of the Hadley circulation could potentially lead to increased water resources in northwest Africa.

Advanced commercial Tokamak optimization studies

International Nuclear Information System (INIS)

Whitley, R.H.; Berwald, D.H.; Gordon, J.D.

1985-01-01

Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units

Tokamak ARC damage

International Nuclear Information System (INIS)

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

1985-01-01

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

Tokamak ARC damage

Energy Technology Data Exchange (ETDEWEB)

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

1985-01-01

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

Planned upgrade to the coaxial plasma source facility for high heat flux plasma flows relevant to tokamak disruption simulations

International Nuclear Information System (INIS)

Caress, R.W.; Mayo, R.M.; Carter, T.A.

1995-01-01

Plasma disruptions in tokamaks remain serious obstacles to the demonstration of economical fusion power. In disruption simulation experiments, some important effects have not been taken into account. Present disruption simulation experimental data do not include effects of the high magnetic fields expected near the PFCs in a tokamak major disruption. In addition, temporal and spatial scales are much too short in present simulation devices to be of direct relevance to tokamak disruptions. To address some of these inadequacies, an experimental program is planned at North Carolina State University employing an upgrade to the Coaxial Plasma Source (CPS-1) magnetized coaxial plasma gun facility. The advantages of the CPS-1 plasma source over present disruption simulation devices include the ability to irradiate large material samples at extremely high areal energy densities, and the ability to perform these material studies in the presence of a high magnetic field. Other tokamak disruption relevant features of CPS-1U include a high ion temperature, high electron temperature, and long pulse length

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)

[99mTc]Demotate 2 in the detection of sst2-positive tumours: a preclinical comparison with [111In]DOTA-tate

International Nuclear Information System (INIS)

Maina, Theodosia; Nock, Berthold A.; Cordopatis, Paul; Bernard, Bert F.; Breeman, Wout A.P.; Gameren, Arthur van; Berg, Ria van den; Krenning, Eric P.; Jong, Marion de; Reubi, Jean-Claude

2006-01-01

The aim of this study was to evaluate [ 99m Tc]Demotate 2 ([ 99m Tc-N 4 0-1 ,Asp 0 ,Tyr 3 ]octreotate) as a candidate for in vivo imaging of sst 2 -positive tumours and to compare it with [ 111 In]DOTA-tate ([ 111 In-DOTA 0 ,Tyr 3 ]octreotate). Labelling of Demotate 2 with 99m Tc was performed at room temperature using SnCl 2 as reductant in the presence of citrate at alkaline pH. Radiochemical analysis involved ITLC and HPLC methods. Peptide conjugate affinities for sst 2 were determined by receptor autoradiography on rat brain cortex sections using [DOTA 0 , 125 I-Tyr 3 ]octreotate as the radioligand. The affinity profile of Demotate 2 for human sst 1 -sst 5 was studied by receptor autoradiography in cell preparations using the universal somatostatin radioligand [ 125 I][Leu 8 ,(D)Trp 22 ,Tyr 25 ]somatostatin-28. The internalisation rates of [ 99m Tc]Demotate 2 and [ 111 In]DOTA-tate were compared in sst 2 -positive and -negative control cell lines. Biodistribution of radiopeptides was studied in male Lewis rats bearing CA20948 tumours. Peptide conjugates showed selectivity and a high affinity binding for sst 2 (Demotate 2 IC 50 =3.2 nM and DOTA-tate IC 50 =5.4 nM). [ 99m Tc]Demotate 2, like [ 111 In]DOTA-tate, internalised rapidly in all sst 2 -positive cells tested, but not in sst 2 -negative control cells. After injection in CA20948 tumour-bearing rats both radiopeptides showed high and specific uptake in the sst 2 -positive organs and in the implanted tumour and rapid excretion from non-target tissues via the kidneys. [ 99m Tc]Demotate 2, similarly to the known sst 2 -targeting agent [ 111 In]DOTA-tate, showed promising biological qualities for application in the scintigraphy of sst 2 -positive tumours. (orig.)

Single-Shell Tank (SST) Retrieval Sequence Fiscal Year 2000 Update

International Nuclear Information System (INIS)

GARFIELD, J.S.

2000-01-01

This document describes the baseline single-shell tank (SST) waste retrieval sequence for the River Protection Project (RPP) updated for Fiscal Year 2000. The SST retrieval sequence identifies the proposed retrieval order (sequence), the tank selection and prioritization rationale, and planned retrieval dates for Hanford SSTs. In addition, the tank selection criteria and reference retrieval method for this sequence are discussed

Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

International Nuclear Information System (INIS)

Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka

2014-03-01

The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

Using CAM3 and the Alkenone Method to Understand how Pliocene SST's Affect California and Other Climates Adjacent to Upwelling Regions

Science.gov (United States)

Searles, Z. A.; Otto-Bliesner, B. L.; Rosenbloom, N. A.; Dekens, P. S.

2008-12-01

The Intergovernmental Panel on Climate Change fourth assessment report established with 90% confidence that anthropogenic climate change will result in a warmer world. In order to more fully understand possible future climate, past analogues of warm periods should be analyzed. The early Pliocene is an appropriate analogue because the continental configuration was similar to today and CO2 levels were comparable to present (~100ppm higher than pre-anthropogenic levels). This project has two objectives: 1.To test the atmospheric sensitivity to USGS PRISM2 dataset revised to better approximate proxy data indicating warmer SST's in upwelling regions and 2. To provide early-Pliocene (2.7- 3.5 Ma) SST estimates for ODP site 1018 on the California Margin (36°59.4'N, 123°16.5'W) using the UK'37 SST proxy. Recent UK'37 proxy data reveals that PRISM2 underestimates SST's in Pacific and Atlantic coastal upwelling regions. The sensitivity of the atmosphere to significantly warmer SST's in upwelling regions was tested using the National Center for Atmospheric Research (NCAR) Community Atmospheric Model, version 3 (CAM3). Initial conditions were supplied by the USGS PRISM2 project and include Pliocene SST's, vegetation cover, sea level height, topography, marine and continental ice extent. Pliocene SST's along the Californian, Peruvian, North African and South African margin were modified to reflect proxy observations from 4 sites. The lack of observational coverage (1 data point per region) required interpolation of the areal extent of the SST warming anomaly indicated by the proxies. Experiment results show increased cumulative precipitation and humidity in regions where the original PRISM2 predicted drying. The predicted North American increase in precipitation is in better agreement with geologic proxies indicating wetter conditions. Warmer SST's also causes weaker wind velocities along the North American, Peruvian, and North African margins, and stronger winds along the

Tokamak power plant burn cycle options

International Nuclear Information System (INIS)

Ehst, D.A.

1994-06-01

Experiments show that tokamaks can operate in various fashions. Economic analyses show that steady state is most attractive provided the physics and technology of current drive (CD) can be modestly improved. Even with very conservative CD assumptions a hybrid operating mode seems superior to conventional, simple inductive operation

Tokamak confinement scaling laws

International Nuclear Information System (INIS)

Connor, J.

1998-01-01

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

Steady state technologies for tokamak based fusion neutron sources and hybrids

International Nuclear Information System (INIS)

Azizov, E.A.; Kuteev, B.V.

2015-01-01

Full text of publication follows. The development of demonstration fusion neutron sources for fusion nuclear science activity and hybrid applications has reached the stage of conceptual design on the basis of tokamak device in Russia. The conceptual design of FNS-ST has been completed in details (plasma current 1.5 MA, magnetic field 1.5 T, major radius 0.5 m, aspect ratio 1.67 and auxiliary heating power up to 15 MW) [1, 2]. A comparison of physical plasma parameters and economics for FNS-ST and a conventional tokamak FNS-CT (plasma current 1.5 MA, magnetic field 6.7 T, major radius 2.25 m, aspect ratio 3 and auxiliary heating power up to 30 MW) has been fulfilled [3]. This study suggested the feasibility to reach 1-20 MW of fusion power using these magnetic configuration options. Nevertheless, the efficiency of neutron production Q remains comparable for both due to the beam fusion input. The total ST-economics for the full project including operation and utilization costs is by a factor of 2 better than of CT. Zero [4] and one-dimensional [5] models have been developed and used in this system analysis. The characteristics of plasma confinement, stability and current drive in operation have been confirmed by numerous benchmarking simulations of modern experiments. Scenarios allowing us to reach and maintain steady state operation have been considered and optimized. The results of these studies will be presented. Prospective technical solutions for SSO-technology systems have been evaluated, and the choice of enabling technologies and materials of the basic FNS options has been made. A conceptual design of a thin-wall water cooled vacuum chamber for heat loadings up to 1.5 MW/m 2 has been fulfilled. The chamber consists of 2 mm Be tiles, pre-shaped CuCrZr 1 mm shell and 1 mm of stainless steel shell as a structural material. A concept of double-null divertor for FNS-ST has been offered that is capable to withstand heat fluxes up to 6 MW/m 2 . Lithium dust

A programmatic framework for the Tokamak Physics Experiment (TPX)

International Nuclear Information System (INIS)

Thomassen, K.I.; Goldston, R.J.; Neilson, G.H.

1993-01-01

Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the open-quotes power breakevenclose quotes level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program

Nuclear fusion research at Tokamak Energy Ltd

International Nuclear Information System (INIS)

Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

2017-01-01

Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

Large Aspect Ratio Tokamak Study

International Nuclear Information System (INIS)

Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

1980-06-01

The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

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

Mechanical design of the coils encapsulated of toroidal field of Tokamak TPM1; Diseno mecanico del encapsulado de las bobinas de campo toroidal del Tokamak TPM1

Energy Technology Data Exchange (ETDEWEB)

Caldino H, U.; Francois L, J. L., E-mail: ucaldino@outlook.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

2014-10-15

The TPM1 is a small Tokamak that belongs to the Centro de Investigacion en Ciencias Aplicadas y Tecnologia Avanzada of Instituto Politecnico Nacional (CICATA-IPN); the project is under construction. Currently it has the vacuum chamber, and is intended that the machine can operate with electric pulses of 10 ms to study the behavior of plasmas in order to provide knowledge in the field of nuclear fusion by magnetic confinement. To achieve this goal is necessary to design the toroidal field coils which operate the Tokamak. This paper presents an analysis which was performed to obtain the correct configuration of coils depending on design parameters for operation of the machine. Once determined this configuration, an analysis of electromagnetic forces present in normal machine operation on one coil was conducted, this to know the stresses in the encapsulation of the same. Considering the pulsed operation, a thickness of 5 mm is determined in the encapsulated, considering fatigue failure based on studies of fatigue failures in epoxy resins. (Author)

Effects of fuelling by using high-pressure supersonic molecular beam in the HL-1M tokamak

International Nuclear Information System (INIS)

Yao Lianghua; Feng Beibin; Feng Zhen; Dong Jiafu; Li Wenzhong; Xu Deming; Hong Wenyu

2002-01-01

Supersonic molecular beam (SMB), as a new fuelling method, has been successfully developed and used in HL-1M tokamak and HT-7 superconducting tokamak. The hydrogen clusters have been found in the beam produced by high working-gas pressure in recent experiments. With a penetration depth of hydrogen particles greater than 17 cm, the rate of increase of electron density for SMB injection, dn e -bar/dt, approaches that of the small ice pellet injection. The plasma density increases step by step after multi-pulse SMB injection, just as multi-pellet fuelling. Comparison of fuelling effects was made between SMB and ice pellet injection on the same shot of ohmic discharge in HL-1M

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.

Microwave reflectrometry for electron density measurements in the TJ-1 tokamak plasma

International Nuclear Information System (INIS)

Anabitarte, E.; Bustamante, E.G.; Calderon, M.A.G.; Vegas, A.

1986-01-01

A study about microwave reflectometry to measure the outside profile of the electron plasma density on tokamak TJ-1 is presented. It is also presented the condition of applicability of this method after the characteristic parameters of the plasma and its resolution. The simulation of the plasma in laboratory by means of a metallic mirror causes the whole characterization of the reflectometer. (author)

Plasma internal inductance dynamics in a tokamak

International Nuclear Information System (INIS)

Romero, J.A.

2010-01-01

A lumped parameter model for tokamak plasma current and inductance time evolution as a function of plasma resistance, non-inductive current drive sources and boundary voltage or poloidal field coil current drive is presented. The model includes a novel formulation leading to exact equations for internal inductance and plasma current dynamics. Having in mind its application in a tokamak inductive control system, the model is expressed in state space form, the preferred choice for the design of control systems using modern control systems theory. The choice of system states allows many interesting physical quantities such as plasma current, inductance, magnetic energy, and resistive and inductive fluxes be made available as output equations. The model is derived from energy conservation theorem, and flux balance theorems, together with a first order approximation for flux diffusion dynamics. The validity of this approximation has been checked using experimental data from JET showing an excellent agreement.

Internal helical modes with m > 1 in a tokamak with a small shear and high plasma pressure

International Nuclear Information System (INIS)

Mikha lovskij, A.B.; Aburdzhaniya, G.D.; Krymskij, A.M.

1979-01-01

Internal helical modes with m>1 in a circular cross-section tokamak with a small shear and large value of the parameter β (β is the ratio between the mean plasma pressure and the mean pressure of the poloidal magnetic field) are investigated. The equations obtained are used to study the destabilizing effects leading to helical instabilities. The role of destabilizing effects is regarded both in local and in a nonlocal approximations on the assumption that the radial plasma pressure is distributed parabolically and that the radial current distribution is also parabolic though slightly varying. It has been established that the profiling of current may lead to the tokamak plasma stability with respect to the modes under investigation. A tokamak with a small shear has been shown to be more stable relative to these modes than that with a large shear

On the warm nearshore bias in Pathfinder monthly SST products over Eastern Boundary upwelling systems

CSIR Research Space (South Africa)

Dufois, F

2012-01-01

Full Text Available Using in situ sea surface temperature (SST) data and MODIS/TERRA SST, the monthly AVHRR Pathfinder (version 5.0 and 5.2) SST product was evaluated within the four main Eastern Boundary Upwelling Systems. A warm bias in the monthly Pathfinder data...

Observations of giant recombination edges on PLT tokamak induced by particle transport

International Nuclear Information System (INIS)

Brau, K.; von Goeler, S.; Bitter, M.; Cowan, R.D.; Eames, D.; Hill, K.; Sauthoff, N.; Silver, E.; Stodiek, W.

1980-03-01

Characteristic steps in the continum spectrum of high temperature tokamak plasmas associated with recombination radiation from impurity ions were observed. During special argon-seeded discharges on the Princeton Large Torus (PLT) tokamak the x-ray spectrum exhibited large enhancements over the bremsstrahlung continuum beginning with energies of 4.1 keV. This corresponds to the radiative capture of free electrons by hydrogen-like argon into the ground state of helium-like argon. A simple particle diffusion model is proposed, with the Ar XVIII radial profiles evaluated from the size of the recombination edges. For the case of moderate density ( approx. 3 x 10 13 cm -3 ) and temperature [T/sub e/(0) approx. 1.5 keV] discharges the outward radial transport velocity is found to be approximately 10 m/sec

Disruption generated secondary runaway electrons in present day tokamaks

International Nuclear Information System (INIS)

Pankratov, I.M.; Jaspers, R.

2000-01-01

An analysis of the runaway electron secondary generation during disruptions in present day tokamaks (JET, JT-60U, TEXTOR) was made. It was shown that even for tokamaks with the plasma current I approx 100 kA the secondary generation may dominate the runaway production during disruptions. In the same time in tokamaks with I approx 1 MA the runaway electron secondary generation during disruptions may be suppressed

Cryogenic analysis of forced-cooled, superconducting TF magnets for compact tokamak reactors

International Nuclear Information System (INIS)

Kerns, J.A.; Slack, D.S.; Miller, J.R.

1988-01-01

Current designs for compact tokamak reactors require the toroidal- field (TF) superconducting magnets to produce fields from 10 to 15 T at the winding pack, using high-current densities to high nuclear heat loads (greater than 1 kW/coil in some instances), which are significantly greater than the conduction and radiation heat loads for which cryogenic systems are usually designed. A cryogenic system for the TF winding pack for two such tokamak designs has been verified by performing a detailed, steady-state heat-removal analysis. Helium properties along the forced-cooled conductor flow path for a range of nuclear heat loads have been calculated. The results and implications of this analysis are presented. 12 refs., 6 figs

A dipole-like SST trend in the Somalia region during the monsoon season

Science.gov (United States)

Santos, F.; Gómez-Gesteira, M.; deCastro, M.; Días, J. M.

2015-02-01

SST trends measured in the Somalia region during the southwest monsoon season over the period 1982-2013 have shown the existence of a warming-cooling dipole. The positive spot, with a warming trend on the order of 0.37°C dec-1, is centered around 5.1°N-50.3°E and the negative one, with a trend on the order of -0.43°C dec-1, around 11.1°N-52.2°E. The migration of the Great Whirl (GW) over the last three decades at a speed of -0.3°C dec-1 in longitude and -0.6°C dec-1 in latitude was considered as the possible origin of the SST dipole. The displacement of the GW produces changes in the geostrophic currents which, in turn, generate changes in the amount of advected water from and to coast.