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Sample records for experimental tokamak text

  1. Feedback control and stabilization experiments on the Texas Experimental Tokamak (TEXT)

    International Nuclear Information System (INIS)

    Uckan, T.; Richards, B.; Wootton, A.J.; Bengtson, R.D.; Bravenec, R.; Carreras, B.A.; Li, G.X.; Hurwitz, P.; Phillips, P.E.; Rowan, W.L.; Tsui, H.Y.W.; Uglum, J.R.; Wen, Y.; Winslow, D.

    1995-01-01

    Plasma edge feedback experiments on the Texas Experimental Tokamak (TEXT) have been successful in controlling the edge plasma potential fluctuation level. The feedback wave-launcher is driven by the local edge potential fluctuations. The edge potential fluctuations are modified in a broad frequency band. Moreover, the potential fluctuations can be reduced (≤100 kHz) without enhancing other modes, or excited (10 to 12 kHz), depending on the phase difference between the driver and the launcher signal, and gain of the system. This turbulence modification is achieved not only locally but also halfway around the torus and has about 2 cm of poloidal extent. The local plasma parameters at the edge and the estimated fluctuation-induced radial particle flux are somewhat affected by the edge feedback. ((orig.))

  2. Progress of recent experimental research on the J-TEXT tokamak

    Science.gov (United States)

    Zhuang, G.; Gentle, K. W.; Chen, Z. Y.; Chen, Z. P.; Yang, Z. J.; Zheng, Wei; Hu, Q. M.; Chen, J.; Rao, B.; Zhong, W. L.; Zhao, K. J.; Gao, L.; Cheng, Z. F.; Zhang, X. Q.; Wang, L.; Jiang, Z. H.; Xu, T.; Zhang, M.; Wang, Z. J.; Ding, Y. H.; Yu, K. X.; Hu, X. W.; Pan, Y.; Huang, H.; the J-TEXT Team

    2017-10-01

    The progress of experimental research over the last two years on the J-TEXT tokamak is reviewed and reported in this paper, including: investigations of resonant magnetic perturbations (RMPs) on the J-TEXT operation region show that moderate amplitude of applied RMPs either increases the density limit from less than 0.7n G to 0.85n G (n G is the Greenwald density, {{n}\\text{G}}={{I}\\text{p}}/π {{a}2} ) or lowers edge safety factor q a from 2.15 to nearly 2.0; observations of influence of RMPs with a large m/n  =  3/1 dominant component (where m and n are the toroidal and poloidal mode numbers respectively) on electron density indicate electron density first increases (decreases) inside (around/outside) of the 3/1 rational surface, and it is increased globally later together with enhanced edge recycling; investigations of the effect of RMPs on the behavior of runaway electrons/current show that application of RMPs with m/n  =  2/1 dominant component during disruptions can reduce runaway production. Furthermore, its application before the disruption can reduce both the amplitude and the length of runaway current; experimental results in the high-density disruption plasmas confirm that local current shrinkage during a multifaceted asymmetric radiation from the edge can directly terminate the discharge; measurements by a multi-channel Doppler reflectometer show that the quasi-coherent modes in the electron diamagnetic direction occur in the J-TEXT ohmic confinement regime in a large plasma region (r/a ~ 0.3-0.8) with frequency of 30-140 kHz.

  3. Feedback control and stabilization experiments on the Texas Experimental Tokamak (TEXT)

    International Nuclear Information System (INIS)

    Uckan, T.; Carreras, B.A.; Richards, B.; Wootton, A.J.; Bengtson, R.D.; Bravenec, R.; Li, G.X.; Hurwitz, P.D.; Phillips, P.E.; Rowan, W.L.

    1994-06-01

    Plasma edge feedback experiments on the Texas Experimental Tokamak (TEXT) have been successful in controlling the edge plasma potential fluctuation level. The feedback wave-launcher, consisting of electrostatic probes located in the shadow of the limiter, is driven by the local edge potential fluctuations. In general, the edge potential fluctuations are modified in a broad frequency band. Moreover, it is observed that the potential fluctuations can be reduced (≤100 kHz) without enhancing other modes, or excited (10 to 12 kHz), depending on the phase difference between the driver and the launcher signal, and gain of the system. This turbulence modification is achieved not only locally but also halfway around the torus and has about 2 cm of poloidal extent. Experiments on the characterization of the global plasma parameters with the edge feedback are discussed. Effects of the edge feedback on the estimated fluctuation-induced radial particle flux as well as on the local plasma parameters are presented

  4. Active probing of plasma edge turbulence and feedback studies on the Texas Experimental Tokamak (TEXT)

    International Nuclear Information System (INIS)

    Uckan, T.; Richards, B.; Bengtson, R.D.

    1993-08-01

    A novel experiment is under way on the Texas Experimental Tokamak (TEXT) to actively modify the turbulence at the plasma edge by launching waves using electrostatic probes in the shadow of the limiter. The experiments are carried out with a wave launching system consisting of two Langmuir probes, which are about 1.8 cm apart in the poloidal direction, with respect to the magnetic field. These probes are operated in the electron side of the (I,V) characteristic. The probe tips are fed separately by independent ac power supplies. Measurements indicate that the wave, launched with a typical frequency image of 15--50 kHz from the edge of the machine top, is received by sensing probes located halfway around the torus. The detected signal strength depends on the frequency of the wave, the plasma current, and the phasing of the applied ac signal between the launching probes. Modifications to the spectra of the density and potential fluctuations are observed. These experiments have been extended to control of the edge plasma fluctuation level using feedback to explore its effects on confinement. When the launcher is driven by the floating potential of the fluctuating plasma at the location of the launching probes, then the fluctuations are suppressed or excited, depending on the phasing between the probe tips, both locally and at the downstream sensing probes. The fluctuation-induced particle flux also varies with the feedback phasing

  5. Active probing of plasma edge turbulence and feedback studies on the Texas Experimental Tokamak (TEXT)

    International Nuclear Information System (INIS)

    Uckan, T.; Richards, B.; Bengtson, R.D.

    1993-01-01

    The edge fluctuations play a critical role in the overall tokamak confinement. Experiments on TEXT show that electrostatic fluctuations in the edge plasma are the dominant mechanism for energy and particle transport. The basic mechanisms responsible for the edge turbulence are the subject of ongoing research in fusion devices. To understand the driving forces responsible for edge fluctuations, a novel experiment is underway on TEXT to actively modify the turbulence at the plasma edge by launching waves using electrostatic probes in the shadow of the limiter. This technique permits active probing of the spectral properties of the edge turbulence. This new approach to the study of edge fluctuations can provide more insight into the basic dynamics of the turbulence and may, in turn, enable detailed comparison with the theory. These experiments, which rely on the use of oscillating electric fields at the plasma edge, complement edge fluctuation control studies that are presently limited to the use of applied dc biasing to influence the edge electric field profile. These experiments have been extended to control of the edge plasma fluctuation level, using feedback to explore its effects on the edge turbulence characteristics as well as on confinement. (author) 8 refs., 7 figs

  6. Active probing of plasma edge turbulence and feedback studies on the Texas Experimental Tokamak (TEXT)

    International Nuclear Information System (INIS)

    Uckan, T.; Carreras, B.A.; Richards, B.; Bengtson, R.D.; Crockett, D.B.; Gentle, K.W.; Li, G.X.; Hurwitz, P.D.; Rowan, W.L.; Tsui, H.Y.W.; Wootton, A.J.

    1993-01-01

    The edge fluctuations play a critical role in the overall tokamak confinement. Experiments on TEXT show that electrostatic fluctuations in the edge plasma are the dominant mechanism for energy and particle transport. The basic mechanisms responsible for the edge turbulence are the subject of ongoing research in fusion devices. To understand the driving forces responsible for edge fluctuations, a novel experiment is underway on TEXT to actively modify the turbulence at the plasma edge by launching waves using electrostatic probes in the shadow of the limiter. This technique permits active probing of the spectral properties of the edge turbulence. This new approach to the study of edge fluctuations can provide more insight into the basic dynamics of the turbulence and may, in turn, enable detailed comparison with the theory. These experiments, which rely on the use of oscillating electric fields at the plasma edge, complement edge fluctuation control studies that are presently limited to the use of applied dc biasing to influence the edge electric field profile. These experiments have been extended to control of the edge plasma fluctuation level, using feedback to explore its effects on the edge turbulence characteristics as well as on confinement

  7. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1990-04-01

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

  8. 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

  9. Measurements on injected impurity transport in TEXT [Texas Experimental Tokamak] using multiply filtered soft x-ray detectors

    International Nuclear Information System (INIS)

    Wenzel, K.W.

    1990-01-01

    Aluminum was injected into TEXT to study trace, non-recycling impurity transport. A 92-channel, three array x-ray imaging system was constructed and installed to measure temporally-resolved density profiles of the three highest charge states. A novel krypton filter in one array discriminated between the He-like and H-like resonance lines, and a hard filter responded mostly to the fully stripped charge state. The impurity confinement time scaled approximately as τ c ∼ bar n e Z eff √m i /Z i /Ip (i denotes the background gas). Aluminum density profiles averaged over a sawtooth crashes were also measured for a few discharges. Sawteeth strongly enhanced the inward impurity flow immediately following injection, when the density was still peaked near the plasma edge. Those discharges with the longest sawtooth period obtained the most peaked aluminum density profiles; thus sawteeth were also important in ameliorating impurity accumulation on the tokamak axis. The charge state balance of the aluminum ions obtained from the measured profiles was compared to predictions of coronal equilibrium. Somewhat surprisingly the aluminum ions were close to coronal, except in those discharges with very short sawtooth periods or very large inversion radii. Preliminary evidence of up-down asymmetric density profiles was also found. Numerical simulations of aluminum transport were performed. The effect of sawtooth oscillations was taken into account with a simple flattening model. The data disagreed with a constant D anomalous model except in the plasma center; enhanced outward transport was required. The experiments did not agree with neoclassical simulations, because the theory had outward convection that was too large. 237 refs., 86 figs., 8 tabs

  10. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  11. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  12. Electromagnetic torques and forces due to misalignment effects and eddy currents in the homopolar generator, power supply for the Texas Experimental Tokamak (TEXT)

    International Nuclear Information System (INIS)

    Driga, M.D.; Bird, W.L.; Tolk, K.M.; Weldon, W.F.; Rylander, H.G.; Woodson, H.H.

    1977-01-01

    Asymmetries in the applied magnetic field due to manufacturing tolerances and rotor-stator misalignments can cause significant forces and moments in a homopolar generator. Parasitic eddy-currents in the rotor, brushes and bearings are also important effects of such asymmetries. The finite element method is used to calculate the magnetic flux distributions in the TEXT homopolar generators. The axial magnetic thrust force and the magnetic tilt moment acting on the rotor are calculated. Eddy-current torques opposing rotor motion are determined using the theory for eddy-current brakes. The results have been used in the design of the TEXT homopolar generator which have been proposed to provide the energy store and conversion for the toroidal field and ohmic heating coils of the new Texas Experimental Tokamak

  13. The recent research progress on the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Wang, Z.J.; Zhuang, G.; Gentle, K.W.

    2013-01-01

    The recent research progress on the J-TEXT tokamak is introduced. The interaction between resonant magnetic perturbations (RMPs) and plasma have been carried out on the J-TEXT tokamak and the results show that the m/n = 2/1 (m and n are the poloidal and toroidal mode numbers, respectively) mode locking is obtained with sufficiently large RMPs while suppression of the m/n = 2/1 tearing mode by moderate magnetic perturbation amplitude is also observed. With a model based on reduced magnetohydrodynamics (MHD) equations, both the mode locking and mode suppression by RMPs are simulated and the results are in good agreement with the experimental observations. To observe the current profile, a high resolution three-wave far infrared polarimeter/interferometer is set up and the first results indicate it works well. (author)

  14. The timing system on the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Zheng, Wei; Zhang, Ming; Zhuang, Ge; Ding, Tonghai; Huang, Fuqiang; Shan, Lingjie

    2014-01-01

    Highlights: •The timing system achieved tree structured timing network with only one type of timing module. •This system is integrated into J-TEXT COADC which is an EPICS based control system. •This system handles multiple timing sequences and events. •This system has been deployed on J-TEXT and working properly in daily experiments. -- Abstract: This paper describes the timing system designed to control the operation time-sequence and to generate clocks for various sub-systems on J-TEXT tokamak. The J-TEXT timing system is organized as a distributed system which is connected by a tree-structured optical fiber network. It can generate delayed triggers and gate signals (0 μs–4000 s), while providing reference clocks for other sub-systems. Besides, it provides event handling and timestamping functions. It is integrated into the J-TEXT Control, Data Access and Communication (J-TEXT CODAC) system, and it can be monitored and configured by Experimental Physics and Industrial Control System (EPICS). The configuration of this system including tree-structured network is managed in XML files by dedicated management software. This system has already been deployed on J-TEXT tokamak and it is serving J-TEXT in daily experiments

  15. The Research Progress of the J-TEXT Tokamak

    Science.gov (United States)

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

    2010-11-01

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

  16. Design and realization of the J-TEXT tokamak central control system

    International Nuclear Information System (INIS)

    Yang Zhoujun; Zhuang Ge; Hu Xiwei; Zhang Ming; Qiu Shengshun; Wang Zhijiang; Ding Yonghua; Pan Yuan

    2009-01-01

    The Joint Texas Experimental Tokamak (J-TEXT), a medium-sized conventional tokamak, serves as a user experimental facility in the China-USA fusion research community. Development of a flexible and easy-to-use J-TEXT central control system (CCS) is of supreme importance for users to coordinate the experimental scenarios with full integration into the discharge operation. This paper describes in detail the structure and functions of the J-TEXT CCS system as well as the performance in practical implementation. Results obtained from both commissioning and routine operations show that the J-TEXT CCS system can offer a satisfactory and effective control that is reliable and stable. The J-TEXT tokamak achieved high-quality performance in its first-ever experimental campaign with this CCS system.

  17. Experimental methods to study tokamak plasma stability

    International Nuclear Information System (INIS)

    Perez-Navarro, A.

    1978-01-01

    Experimental devices to measure external instability modes with small pick-up coils to detect poloidal magnetic field fluctuations, and internal modes with soft-X-ray detectors are discussed. The characteristics of these devices are calculated for a small tokamak (R 0 = 30 cm, a = 10 cm, I 0 50 KA). (author)

  18. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

    Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

    1979-01-01

    This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). However, the results of this investigation are believed to be generally applicable to the broad class of the next generation of experimental tokamak facilities such as ETF. The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties are compared to the benefits and conclusions and recommendations are developed on resolving the issue

  19. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

    Chan, V.S.

    1994-09-01

    In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

  20. The reconstruction and research progress of the TEXT-U tokamak in China

    Science.gov (United States)

    Zhuang, G.; Pan, Y.; Hu, X. W.; Wang, Z. J.; Ding, Y. H.; Zhang, M.; Gao, L.; Zhang, X. Q.; Yang, Z. J.; Yu, K. X.; Gentle, K. W.; Huang, H.; J-TEXT Team

    2011-09-01

    The TEXT/(TEXT-U) tokamak, formerly built and operated by the University of Texas at Austin in USA, was dismantled and shipped to China in 2004, and renamed as the Joint TEXT (J-TEXT) tokamak. The reconstruction work, which included reassembly of the machine and development of peripheral devices, was completed in the spring of 2007. Consequently, the first plasma was obtained at the end of 2007. At present, a typical J-TEXT ohmic discharge can produce a plasma with flattop current up to 220 kA and lasting for 300 ms, line-averaged density above 2 × 1019 m-3, and an electron temperature of about 800 eV, with a toroidal magnetic field of 2.2 T. A number of diagnostic devices used to facilitate the routine operation and experimental scenarios were developed on the J-TEXT tokamak. Hence, the measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the last closed flux surface were undertaken. The observation and simple analysis of MHD activity and disruption events were also performed. The preliminary experimental results and the future research plan for the J-TEXT are described in detail.

  1. The reconstruction and research progress of the TEXT-U tokamak in China

    International Nuclear Information System (INIS)

    Zhuang, G.; Pan, Y.; Hu, X.W.; Wang, Z.J.; Ding, Y.H.; Zhang, M.; Gao, L.; Zhang, X.Q.; Yang, Z.J.; Yu, K.X.; Gentle, K.W.; Huang, H.

    2011-01-01

    The TEXT/(TEXT-U) tokamak, formerly built and operated by the University of Texas at Austin in USA, was dismantled and shipped to China in 2004, and renamed as the Joint TEXT (J-TEXT) tokamak. The reconstruction work, which included reassembly of the machine and development of peripheral devices, was completed in the spring of 2007. Consequently, the first plasma was obtained at the end of 2007. At present, a typical J-TEXT ohmic discharge can produce a plasma with flattop current up to 220 kA and lasting for 300 ms, line-averaged density above 2 x 10 19 m -3 , and an electron temperature of about 800 eV, with a toroidal magnetic field of 2.2 T. A number of diagnostic devices used to facilitate the routine operation and experimental scenarios were developed on the J-TEXT tokamak. Hence, the measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the last closed flux surface were undertaken. The observation and simple analysis of MHD activity and disruption events were also performed. The preliminary experimental results and the future research plan for the J-TEXT are described in detail.

  2. Robust control design for the plasma horizontal position control on J-TEXT Tokamak

    International Nuclear Information System (INIS)

    Yu, W.Z.; Chen, Z.P.; Zhuang, G.; Wang, Z.J.

    2013-01-01

    It is extremely important for tokamak to control the plasma position during routine discharge. However, the model of plasma in tokamak usually contains much of the uncertainty, such as structured uncertainties and unmodeled dynamics. Compared with the traditional PID control approach, robust control theory is more suitable to handle this problem. In the paper, we propose a H ∞ robust control scheme to control the horizontal position of plasma during the flat-top phase of discharge on Joint Texas Experimental Tokamak (J-TEXT) tokamak. First, the model of our plant for plasma horizontal position control is obtained from the position equilibrium equations. Then the H ∞ robust control framework is used to synthesize the controller. Based on this, an H ∞ controller is designed to minimize the regulation/tracking error. Finally, a comparison study is conducted between the optimized H ∞ robust controller and the traditional PID controller in simulations. The simulation results of the H ∞ robust controller show a significant improvement of the performance with respect to those obtained with traditional PID controller, which is currently used on our machine

  3. Experimental investigations at the Soviet tokamaks

    International Nuclear Information System (INIS)

    Bobrovskij, G.A.; Golant, V.E.; AN SSSR, Leningrad. Fiziko-Tekhnicheskij Inst.)

    1978-01-01

    The review is devoted to the basic results obtained on the Soviet tokamaks during 1976-1977. Behaviour of impurities, tearing instability, additional methods of plasma heating, energy distribution function were investigated. A brief description of new T-7, TM-4, ''Tuman-3'' tokamaks is given. It is shown that despite inflow of impurities to the pinch periphery, no their appreciable accumulation is observed at least during the discharge time. It is shown that the helical perturbations with m=2 and 1 present the greatest danger. The suppression of the tearing instability is related with suppression of the mode with m=2. The helical perturbation prevents formation of skin configuration at the initial stage of the discharge. As a rule, the transition of an appreciable fraction of electrons to continuous acceleration does not take place, although a significant deformation of electron distribution function under the action of electric field occurs. Plasma compression by increasing magnetic field induces oscillations and improves thermal plasma isolation. It is shown experimentally that the considerable efficiency of energy contribution to the ion component at the central part of plasma may be obtained by means of HF heating under conditions of low-hybrid resonance. It is shown that the recombination has a considerable effect on concentration of neutral particles in the central region

  4. Experimental observations related to the thermodynamic properties of tokamak plasmas

    International Nuclear Information System (INIS)

    Sozzi, C.; Minardi, E.; Lazzaro, E.; Cirant, S.; Mantica, P.; Esposito, B.; Marinucci, M.; Romanelli, M.; Imbeaux, F.

    2005-01-01

    The coarse-grained tokamak plasma description derived from the magnetic entropy concept presents appealing features as it involves a simple mathematics and it identifies a limited set of characteristic parameters of the macroscopic equilibrium. In this paper a comprehensive review of the work done in order to check the reliability of the Stationary Magnetic Entropy predictions against experimental data collected from different tokamaks, plasma regimes and heating methods is reported. (author)

  5. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

    Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

    1979-01-01

    This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties to the benefits and conclusions and recommendations on resolving the issue are discussed

  6. Experimental results from the TUMAN 3 tokamak

    International Nuclear Information System (INIS)

    Golant, V.E.; Andrejko, M.V.; Askinazi, L.G.; Korneev, V.A.; Krikunov, S.V.; Lipin, B.M.; Lebedev, S.V.; Levin, L.S.; Podushnikova, K.A.; Razdobarin, G.T.; Rozhansky, V.A.; Rozhdestvensky, V.V.; Tendler, M.; Tukachinsky, A.S.; Jaroshevich, S.P.

    1995-01-01

    The open-quote open-quote TUMAN-3 close-quote close-quote Tokamak programme concentrates on issues of improved confinement. In 1989 the transition from an ordinary Ohmic regime into an improved confinement mode was achieved. The signatures of the H-mode in auxiliary heated tokamaks have been observed in this regime. The crucial role of the boundary radial electric field was found in the experiments with internal bias probe. Other techniques were demonstrated to disturb the boundary plasma which led to H-mode triggering: short increase of working gas puffing, minor radius magnetic compression and pellet injection. The role scaling of the energy confinement time in the Ohmic H-mode was obtained, which differs dramatically from the scaling for the ordinary Ohmic regime. There were found a strong dependence of τ E on plasma current and a weak dependence on density. The maximum value of τ E was 10 times longer than in the ordinary Ohmic region. The τ E scaling for the Ohmic H-mode is consistent with the scaling proposed for devices with powerful auxiliary heating. The results shows that H-mode physics is universal in tokamaks with different geometries and heating methods. (AIP) copyright 1995 American Institute of Physics

  7. Design and application of an EPICS compatible slow plant system controller in J-TEXT tokamak

    International Nuclear Information System (INIS)

    Zhang, J.; Zhang, M.; Zheng, W.; Zhuang, G.; Ding, T.

    2014-01-01

    Highlights: • Underlying functionalities are encapsulated into plug-and-play modules. • The slow controller is EPICS compatible. • The slow controller can work as PSH. - Abstract: J-TEXT tokamak has recently implemented J-TEXT COntrol, Data Access and Communication (CODAC) system on the principle of ITER CODAC. The control network in J-TEXT CODAC system is based on Experimental Physics and Industrial Control System (EPICS). However, former slow plant system controllers in J-TEXT did not support EPICS. Therefore, J-TEXT has designed an EPICS compatible slow controller. And moreover, the slow controller also acts the role of Plant System Host (PSH), which helps non-EPICS controllers to keep working in J-TEXT CODAC system. The basic functionalities dealing with user defined tasks have been modularized into driver or plug-in modules, which are plug-and-play and configured with XML files according to specific control task. In this case, developers are able to implement various kinds of control tasks with these reusable modules, regardless of how the lower-lever functions are implemented, and mainly focusing on control algorithm. And it is possible to develop custom-built modules by themselves. This paper presents design of the slow controller. Some applications of the slow controller have been deployed in J-TEXT, and will be introduced in this paper

  8. Design and application of an EPICS compatible slow plant system controller in J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, J.; Zhang, M. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zheng, W., E-mail: zhengwei@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Zhuang, G.; Ding, T. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-05-15

    Highlights: • Underlying functionalities are encapsulated into plug-and-play modules. • The slow controller is EPICS compatible. • The slow controller can work as PSH. - Abstract: J-TEXT tokamak has recently implemented J-TEXT COntrol, Data Access and Communication (CODAC) system on the principle of ITER CODAC. The control network in J-TEXT CODAC system is based on Experimental Physics and Industrial Control System (EPICS). However, former slow plant system controllers in J-TEXT did not support EPICS. Therefore, J-TEXT has designed an EPICS compatible slow controller. And moreover, the slow controller also acts the role of Plant System Host (PSH), which helps non-EPICS controllers to keep working in J-TEXT CODAC system. The basic functionalities dealing with user defined tasks have been modularized into driver or plug-in modules, which are plug-and-play and configured with XML files according to specific control task. In this case, developers are able to implement various kinds of control tasks with these reusable modules, regardless of how the lower-lever functions are implemented, and mainly focusing on control algorithm. And it is possible to develop custom-built modules by themselves. This paper presents design of the slow controller. Some applications of the slow controller have been deployed in J-TEXT, and will be introduced in this paper.

  9. Revised design for the Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1977-03-01

    A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost

  10. Neutronic scoping studies for the tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Bettis, E.S.; McAlees, D.G.; Watts, H.L.; Williams, M.L.

    1976-02-01

    One-dimensional neutron and photon radiation transport methods have been used to investigate candidate blanket configurations and compositions for use in the Tokamak Experimental Power Reactor. Seven blanket designs are compared in terms of energy recovery, radiation attenuation, potential radiation damage, and, where applicable, tritium breeding

  11. Recent Activities on the Experimental Research Programme Using Small Tokamaks

    International Nuclear Information System (INIS)

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

    2006-01-01

    A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project (CRP) is discussed in this paper. Besides the presentation of the recent activities on the experimental research programme using small tokamaks and scientific results achieved at the participating laboratories, information is provided about the organisation of the co-ordinated research project. Future plans of the co-ordinated activities within the CRP are discussed

  12. TIBER: Tokamak Ignition/Burn Experimental Research. Final design report

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Barr, W.L.

    1985-01-01

    The Tokamak Ignition/Burn Experimental Research (TIBER) device is the smallest superconductivity tokamak designed to date. In the design plasma shaping is used to achieve a high plasma beta. Neutron shielding is minimized to achieve the desired small device size, but the superconducting magnets must be shielded sufficiently to reduce the neutron heat load and the gamma-ray dose to various components of the device. Specifications of the plasma-shaping coil, the shielding, coaling, requirements, and heating modes are given. 61 refs., 92 figs., 30 tabs

  13. Characteristics of current quenches during disruptions in the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Zhang, Y; Chen, Z Y; Fang, D; Jin, W; Huang, Y H; Wang, Z J; Yang, Z J; Chen, Z P; Ding, Y H; Zhang, M; Zhuang, G

    2012-01-01

    Characteristics of tokamak current quenches are an important issue for the determination of electro-magnetic forces that act on the in-vessel components and vacuum vessel during major disruptions. The characteristics of current quenches in spontaneous disruptions in the J-TEXT tokamak have been investigated. It is shown that the waveforms for the fastest current quenches are more accurately fitted by linear current decays than exponential, although neither is a good fit in many slower cases. The minimum current quench time is about 2.4 ms for the J-TEXT tokamak. The maximum instantaneous current quench rate is more than seven times the average current quench rate in J-TEXT. (paper)

  14. Experimental operation of the KT-5C tokamak in USTC

    International Nuclear Information System (INIS)

    Wen Yizhi; Wan Shude; Zhai Kan; Liu Wandong

    1995-01-01

    Experimental operation of the KT-5C tokamak was started in early 1991. More than 3 x 10 4 shots of discharges have been performed so far. The authors deals with the major features of operation control of the KT-5C device. As the machine is usually controlled by an experimenter himself without duty operator, an audible indicator makes it much easier to control and operate. Accident and interference prevention systems prove to be reliable, and the operation system works successfully

  15. Abstracts of the International seminar 'Experimental possibilities of KTM tokamak and research programme'

    International Nuclear Information System (INIS)

    2005-01-01

    The International seminar 'Experimental possibilities of KTM tokamak and research programme' was held in 10-12 October 2005 in Astana city (Kazakhstan). The seminar was dedicated to problems of KTM tokamak commissioning. The Collection of abstracts comprises 45 papers

  16. Experimental results from the TFTR tokamak

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Arunasalam, V.; Bell, J.D.

    1986-10-01

    Recent experiments on TFTR have extended the operating regime of TFTR in both ohmic- and neutral-beam-heated discharges. The TFTR tokamak has reached its original machine design specifications (I/sub p/ = 2.5 MA and B/sub T/ = 5.2 T). Initial neutral-beam-heating experiments used up to 6.3 MW of deuterium beams. With the recent installation of two additional beamlines, the power has been increased up to 11 MW. A deuterium pellet injector was used to increase the central density to 2.5 x 10 20 m -3 in high current discharges. At the opposite extreme, by operating at low plasma current (I/sub p/ ∼ 0.8 MA) and low density (anti n/sub e/ ∼ 1 x 10 19 m -3 ), high ion temperatures (9 +- 2 keV) and rotation speeds (7 x 10 5 m/s) have been achieved during injection. In addition, plasma compression experiments have demonstrated acceleration of beam ions from 82 keV to 150 keV, in accord with expectations. The wide operating range of TFTR, together with an extensive set of diagnostics and a flexible control system, has facilitated transport and scaling studies of both ohmic- and neutral-beam-heated discharges. The results of these confinement studies are presented

  17. Electromagneto-mechanical coupling analysis of a test module in J-TEXT Tokamak during plasma disruption

    Energy Technology Data Exchange (ETDEWEB)

    Dong, Haijie; Yuan, Zhensheng; Yuan, Hongwei; Pei, Cuixiang [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Chen, Zhenmao, E-mail: chenzm@mail.xjtu.edu.cn [State Key Laboratory for Strength and Vibration of Mechanical Structures, Shanxi Engineering Research Center for NDT and Structural Integrity Evaluation Xi’an Jiaotong University, Xi’an 710049 (China); Yang, Jinhong; Wang, Weihua [Institute of Applied Physics of AOA, Hefei 230031 (China)

    2016-11-01

    In this paper, the dynamic response during plasma disruption of a test blanket module in vacuum vessel (VV) of the Joint TEXT (J-TEXT), which is an experimental Tokamak device with iron core, was simulated by applying a program developed by authors on the ANSYS platform using its parametric design language (APDL). The moving coordinate method as well as the load transfer and sequential coupling strategy were adopted to cope with the electromagneto-mechanical coupling effect. To establish the numerical model, the influence of the iron core on the eddy current and electromagnetic (EM) force during disruption was numerically investigated at first and the influence was found not significant. Together with the geometrical features of the J-TEXT Tokamak structure, 180° sector models without magnetic core were finally established for the EM field and the structural response simulations. To obtain the source plasma current, the plasma current evolution during disruption was simulated by using the Tokamak Simulation Code (TSC). With the numerical models and the source plasma current, the dynamic response of both the VV structure and the test module were calculated. The numerical results show that the maximum stress of the test module is in safe range, and the magnetic damping effect can weaken vibration of the test module. In addition, simulation without considering the coupling effect was carried out, which shows that the influence of coupling effect is not significant for the peak stress of the J-TEXT disruption problem.

  18. 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

  19. The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

    Science.gov (United States)

    Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

    2017-10-01

    Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

  20. Response of plasma rotation to resonant magnetic perturbations in J-TEXT tokamak

    Science.gov (United States)

    Yan, W.; Chen, Z. Y.; Huang, D. W.; Hu, Q. M.; Shi, Y. J.; Ding, Y. H.; Cheng, Z. F.; Yang, Z. J.; Pan, X. M.; Lee, S. G.; Tong, R. H.; Wei, Y. N.; Dong, Y. B.; J-TEXT Team

    2018-03-01

    The response of plasma toroidal rotation to the external resonant magnetic perturbations (RMP) has been investigated in Joint Texas Experimental Tokamak (J-TEXT) ohmic heating plasmas. For the J-TEXT’s plasmas without the application of RMP, the core toroidal rotation is in the counter-current direction while the edge rotation is near zero or slightly in the co-current direction. Both static RMP experiments and rotating RMP experiments have been applied to investigate the plasma toroidal rotation. The core toroidal rotation decreases to lower level with static RMP. At the same time, the edge rotation can spin to more than 20 km s-1 in co-current direction. On the other hand, the core plasma rotation can be slowed down or be accelerated with the rotating RMP. When the rotating RMP frequency is higher than mode frequency, the plasma rotation can be accelerated to the rotating RMP frequency. The plasma confinement is improved with high frequency rotating RMP. The plasma rotation is decelerated to the rotating RMP frequency when the rotating RMP frequency is lower than the mode frequency. The plasma confinement also degrades with low frequency rotating RMP.

  1. Test particle calculations for the Texas experimental tokamak with resonant magnetic fields

    International Nuclear Information System (INIS)

    Wootton, A.J.; McCool, S.C.; Zheng, S.

    1991-01-01

    This paper presents a simple test particle model that attempts to describe particle motion in the presence of intrinsic electrostatic fluctuations in a prescribed tokamak magnetic field. In particular, magnetic field configurations that include externally produced magnetic islands and stochastic regions are considered. The resulting test particle transport is compared with the predictions of analytic models and with the experimentally measured electron heat and particle transport on the Texas Experimental Tokamak (TEXT). Agreement between the test particle results and applicable analytic theories is found. However, there is only partial agreement with the experimental results, and possible reasons for the discrepancies are explored. Good agreement is found between predicted and measured spatially asymmetric particle distributions. The particle collection efficiency of an apertured limiter inside a magnetic island (an intra-island pump limiter) is discussed

  2. Measurement of the hydrogen recombination coefficient in the TEXT tokamak as a function of outgassing and power radiated during tokamak discharges

    International Nuclear Information System (INIS)

    Langley, R.A.; Rowan, W.L.; Bravenec, R.V.; Nelin, K.

    1986-10-01

    The global recombination rate coefficient k/sub r/ for hydrogen has been measured in the TEXT tokamak vacuum vessel for various surface conditions. An attempt was made to correlate the measured values of k/sub r/ with residual gas analyzer (RGA) data taken before each measurement of k/sub r/ and with the power radiated during tokamak discharges produced after each measurement of k/sub r/. The results show that k/sub r/ increases during a series of tokamak discharges, k/sub r/ is relatively insensitive to power radiated during tokamak discharges, and k/sub r/ increases with the RGA measurements of mass 28 and 40 but not with those of mass 18. In addition, it was found that the mass 18 (H 2 O) signal decreases as glow discharge experiments with hydrogen were performed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

  4. 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

  5. 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)

  6. Plasma density remote control system of experimental advanced superconductive tokamak

    International Nuclear Information System (INIS)

    Zhang Mingxin; Luo Jiarong; Li Guiming; Wang Hua; Zhao Dazheng; Xu Congdong

    2007-01-01

    In Tokamak experiments, experimental data and information on the density control are stored in the local computer system. Therefore, the researchers have to be in the control room for getting the data. Plasma Density Remote Control System (DRCS), which is implemented by encapsulating the business logic on the client in the B/S module, conducts the complicated science computation and realizes the synchronization with the experimental process on the client. At the same time, Web Services and Data File Services are deployed for the data exchange. It is proved in the experiments that DRCS not only meets the requirements for the remote control, but also shows an enhanced capability on the data transmission. (authors)

  7. Physics design and experimental study of tokamak divertor

    International Nuclear Information System (INIS)

    Yan Jiancheng; Gao Qingdi; Yan Longwen; Wang Mingxu; Deng Baiquan; Zhang Fu; Zhang Nianman; Ran Hong; Cheng Fayin; Tang Yiwu; Chen Xiaoping

    2007-06-01

    The divertor configuration of HL-2A tokamak is optimized, and the plasma performance in divertor is simulated with B2-code. The effects of collisionality on plasma-wall transition in the scrape-off layer of divertor are investigated, high performances of the divertor plasma in HL-2A are simulated, and a quasi- stationary RS operation mode is established with the plasma controlled by LHCD and NBI. HL-2A tokamak has been successfully operated in divertor configuration. The major parameters: plasma current I p =320 kA, toroidal field B t =2.2 T, plasma discharger duration T d =1580 ms ware achieved at the end of 2004. The preliminary experimental researches of advanced diverter have been carried out. Design studies of divertor target plate for high power density fusion reactor have been carried out, especially, the physical processes on the surface of flowing liquid lithium target plate. The exploration research of improving divertor ash removal efficiency and reducing tritium inventory resulting from applying the RF ponderomotive force potential is studied. The optimization structure design studies of FEB-E reactor divertor are performed. High flux thermal shock experiments were carried on tungsten and carbon based materials. Hot Isostatic Press (HIP) method was employed to bond tungsten to copper alloys. Electron beam simulated thermal fatigue tests were also carried out to W/Cu bondings. Thermal desorption and surface modification of He + implanted into tungsten have been studied. (authors)

  8. Hybrid neural network for density limit disruption prediction and avoidance on J-TEXT tokamak

    Science.gov (United States)

    Zheng, W.; Hu, F. R.; Zhang, M.; Chen, Z. Y.; Zhao, X. Q.; Wang, X. L.; Shi, P.; Zhang, X. L.; Zhang, X. Q.; Zhou, Y. N.; Wei, Y. N.; Pan, Y.; J-TEXT team

    2018-05-01

    Increasing the plasma density is one of the key methods in achieving an efficient fusion reaction. High-density operation is one of the hot topics in tokamak plasmas. Density limit disruptions remain an important issue for safe operation. An effective density limit disruption prediction and avoidance system is the key to avoid density limit disruptions for long pulse steady state operations. An artificial neural network has been developed for the prediction of density limit disruptions on the J-TEXT tokamak. The neural network has been improved from a simple multi-layer design to a hybrid two-stage structure. The first stage is a custom network which uses time series diagnostics as inputs to predict plasma density, and the second stage is a three-layer feedforward neural network to predict the probability of density limit disruptions. It is found that hybrid neural network structure, combined with radiation profile information as an input can significantly improve the prediction performance, especially the average warning time ({{T}warn} ). In particular, the {{T}warn} is eight times better than that in previous work (Wang et al 2016 Plasma Phys. Control. Fusion 58 055014) (from 5 ms to 40 ms). The success rate for density limit disruptive shots is above 90%, while, the false alarm rate for other shots is below 10%. Based on the density limit disruption prediction system and the real-time density feedback control system, the on-line density limit disruption avoidance system has been implemented on the J-TEXT tokamak.

  9. Prediction of density limit disruptions on the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Wang, S Y; Chen, Z Y; Huang, D W; Tong, R H; Yan, W; Wei, Y N; Ma, T K; Zhang, M; Zhuang, G

    2016-01-01

    Disruption mitigation is essential for the next generation of tokamaks. The prediction of plasma disruption is the key to disruption mitigation. A neural network combining eight input signals has been developed to predict the density limit disruptions on the J-TEXT tokamak. An optimized training method has been proposed which has improved the prediction performance. The network obtained has been tested on 64 disruption shots and 205 non-disruption shots. A successful alarm rate of 82.8% with a false alarm rate of 12.3% can be achieved at 4.8 ms prior to the current spike of the disruption. It indicates that more physical parameters than the current physical scaling should be considered for predicting the density limit. It was also found that the critical density for disruption can be predicted several tens of milliseconds in advance in most cases. Furthermore, if the network is used for real-time density feedback control, more than 95% of the density limit disruptions can be avoided by setting a proper threshold. (paper)

  10. First time observation of local current shrinkage during the MARFE behavior on the J-TEXT tokamak

    Science.gov (United States)

    Shi, Peng; Zhuang, G.; Gentle, K.; Hu, Qiming; Chen, Jie; Li, Qiang; Liu, Yang; Gao, Li; Zhang, Xiaolong; Liu, Hai; Chen, Zhipeng; Zhu, Lizhi; Li, Fuming; Zhou, Yinan; Zeng, Zhong; Liu, Linzi; He, Jiyang

    2017-11-01

    Multifaceted asymmetric radiation as well as strong poloidal asymmetry of the electron density from the edge, dubbed as ‘MARFE’, has been observed in high electron density Ohmically heated plasmas on J-TEXT tokamak. Equilibrium reconstruction based on the measured data from the 17-channel FIR polarimeter-interferometer indicates that an asymmetric plasma current density distribution forms at the edge region and the plasma current shrinkage locates at the MARFE affected region. Furthermore, associated with the localized plasma current shrinkage, a locked mode MHD activity is excited, which then terminate the discharge with a major disruption. Localized plasma current shrinkage at the MARFE region is considered to be the direct cause for the density limit disruptions, and the proposed interpretation is consistent with the experimental observations.

  11. Neural Network Prediction of Disruptions Caused by Locked Modes on J-TEXT Tokamak

    International Nuclear Information System (INIS)

    Ding Yonghua; Jin Xuesong; Chen Zhenzhen; Zhuang Ge

    2013-01-01

    Prediction of disruptions caused by locked modes using the Back-Propagation (BP) neural network is completed on J-TEXT tokamak. The network, which is based on the BP neural network, uses Mirnov coils and locked mode coils signals as input data, and outputs a signal including information of prediction of locked mode. The rate of successful prediction of locked modes is more than 90%. For intrinsic locked mode disruptions, the network can give a prewarning signal about 1 ms ahead of the locking-time. For the disruption caused by resonant magnetic perturbation (RMPs) locked modes, the network can give a prewarning signal about 10 ms ahead of the locking-time

  12. Note: Measurement of the runaway electrons in the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Chen, Z. Y.; Zhang, Y.; Zhang, X. Q.; Luo, Y. H.; Jin, W.; Li, J. C.; Chen, Z. P.; Wang, Z. J.; Yang, Z. J.; Zhuang, G.

    2012-01-01

    The runaway electrons have been measured by hard x-ray detectors and soft x-ray array in the J-TEXT tokamak. The hard x-ray radiations in the energy ranges of 0.5-5 MeV are measured by two NaI detectors. The flux of lost runaway electrons can be obtained routinely. The soft x-ray array diagnostics are used to monitor the runaway beam generated in disruptions since the soft x-ray is dominated by the interaction between runaway electrons and metallic impurities inside the plasma. With the aid of soft x-ray array, runaway electron beam has been detected directly during the formation of runaway current plateau following the disruptions.

  13. Design of the power supply system for the plasma current modulation on J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, M.; Shao, J.; Ma, S.X., E-mail: mashaoxiang@hust.edu.cn; Liang, X.; Yu, K.X.; Pan, Y.

    2016-10-15

    Highlights: • A modification scheme of heating field power supply system for plasma current modulation. • High-power fast control power supply with multilevel cascade circuit. • Restraining circulating current with coupled inductors in cyclic symmetric structure. - Abstract: In order to further study the influence of current modulation parameters on suppressing tearing instability, the plasma current should be modulated in a wider range. So a modification scheme is designed to improve the performance of ohmic heating power supply system on J-TEXT tokamak. A multilevel cascade circuit with carrier phase-shifted PWM technique has been proposed. Coupled inductors are connected in the form of cyclic symmetry to restrain the circulating current caused by multiple paralleled branches. The simulation proves this proposed current modulation power supply system matches output requirement and achieves good current sharing effect. Finally, a prototype is designed, and the experiment results can verify the correctness of the simulation model well.

  14. Dispersion relations of density fluctuations observed by heavy ion beam probe in the TEXT tokamak

    International Nuclear Information System (INIS)

    Ross, D.W.

    1990-09-01

    Wave numbers as functions of frequency for density fluctuations in the core of the TEXT tokamak are measured in Heavy Ion Beam Probe experiments by analyzing the relative phases of signals originating from nearby points in the plasma. The adjacent points are typically 2 cm apart, with their relative orientation (δr, δθ) depending on position (r,θ). for angular frequencies ω ≤ 10 6 /s the signals are quite coherent, leading to reasonably well-defined ''dispersion relations.'' These do not correspond to known modes of the drift wave type, i.e., ballooning or slab-like electron drift waves or ion temperature gradient modes. The effect of finite sample volume size does not significantly alter this conclusion. 25 refs., 6 figs., 3 tabs

  15. Study on paralleled inverters with current-sharing coupled inductors on J-TEXT Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Shao, J.; Rao, B., E-mail: borao@hust.edu.cn; Zhang, M.; Ma, S.X.; Liang, X.; Yu, K.X.; Pan, Y.

    2016-12-15

    Highlights: • A modification scheme of heating field power supply system for plasma current modulation. • High-power fast control power supply with multilevel cascade circuit. • Restraining circulating current with coupled inductors in cyclic symmetric structure. • Analysis on the topology with current-sharing coupled inductors. - Abstract: The coupled inductors in paralleled inverters are applied to restrain the high frequency circulating current on J-TEXT Tokamak. Compared with individual inductor, this method has the benefit of high voltage utilization, less volume and weight of the inductor. In this paper, circuit topology of coupled inductors in cyclic symmetry structure for steady-state operation is analyzed and then the design of the inductor is introduced. The maximum circulating current is related to number of parallel branch, DC side voltage, self-inductance of the inductor and the frequency of carrier wave. The simulation and prototype experiment results verify the design.

  16. 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

  17. Experimental device for the X-ray energetic distribution measurement in a tokamak plasma

    International Nuclear Information System (INIS)

    Perez-Navarro, A.

    1977-01-01

    An experimental system to measure the X-ray spectrum in a tokamak plasma is described, emphasizing its characteristics: resolution, dead time and the pulse pile-up distortion effects on the X-ray spectra. (author) [es

  18. Experimental result of poloidal limiter baking of Aditya tokamak

    International Nuclear Information System (INIS)

    Jadeja, K.A.; Arambhadiya, B.G.; Bhatt, S.B.; Bora, D.

    2005-01-01

    In tokamak Aditya, Poloidal limiter function as the operational limiter and are subjected to very high particles load and heat flux during plasma discharge. In addition, Poloidal limiter is the first material surface to come in contact with the hot plasma. In plasma discharge, the impurity generations from limiter are mostly by adsorbed particles. The baking of limiter provides high degassing rate and thermal desorption of adsorbed particles of limiter to reduce impurities from the limiter tiles. The series of experiments are done with different conditions like, Baking of limiter SS ring by heating element with and without limiter tiles in atmosphere and vacuum. Than Poloidal limiter is structured with 14 numbers of graphite tiles and electrical isolated to the vessel and support structure. As a heating element and for electrical isolation, Nychrome wire and ceramic block with ceramic tubes are used. In addition, Thermo couple and two DC power supply (0-10 Ampere) are used for limiter baking. Mass analyzer gives partial pressures of different species to observe effect of limiter baking. For the period of Poloidal limiter baking in Aditya, the partial pressures of different species like hydrogen, water vapor, and oxygen are extremely increased with time duration. This paper presents series of experimental results of poloidal limiter baking. (author)

  19. Design study of toroidal magnets for tokamak experimental power reactors

    International Nuclear Information System (INIS)

    Stekly, Z.J.J.; Lucas, E.J.

    1976-12-01

    This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed

  20. Experimental progress and innovation on the HL-2A tokamak

    International Nuclear Information System (INIS)

    Ding Xuantong

    2010-01-01

    The HL-2A Tokamak is the first large controlled fusion experiment device with divertor in China. In this paper, the main experimental results on this device will be presented. Since its establishment, the operation conditions have been improved greatly. First, the divertor configuration was realized, the electron temperature was increased to 5.5 keV by electron cyclotron heating, and then the high confinement edge localized mode was achieved. With the development of high power auxiliary heating and advanced plasma diagnostic systems,innovative contributions have been made in certain plasma physics areas. The 3-dimension structure of the zonal flow has been identified, which is very important in the transport of fusion plasma. Supersonic molecular beam injection has also been developed and successfully used for plasma transport. The tearing mode has been suppressed by electron cyclotron resonance heating with low frequency modulation, and the confinement has been improved. The new phenomena of the internal kink mode and Alfen mode excited by energetic electrons have been observed. Future plans and new experiments on the device will also be briefly presented. (authors)

  1. Oak Ridge Tokamak experimental power reactor study scoping report

    International Nuclear Information System (INIS)

    Roberts, M.

    1977-03-01

    This report presents the scoping studies performed as the initial part of the program to produce a conceptual design for a Tokamak Experimental Power Reactor (EPR). The EPR as considered in this study is to employ all systems necessary for significant electric power production at continuous high duty cycle operation; it is presently scheduled to be the final technological step before a Demonstration Reactor Plant (Demo). The scoping study tasks begin with an exploration and identification of principal problem areas and then concentrate on consideration and evaluation of alternate design choices for each of the following major systems: Plasma Engineering and Physics, Nuclear, Electromagnetics, Neutral Beam Injection, and Tritium Handling. In addition, consideration has been given to the integration of these systems and requirements arising out of their incorporation into an EPR. One intent of this study is to document the paths explored in search of the appropriate EPR characteristics. To satisfy this intent, the explorations are presented in chart form outlining possible options in key areas with extensive supporting footnotes. An important result of the scoping study has been the development and definition of an EPR reference design to serve as (1) a common focus for the continuing design study and (2) a guide for associated development programs. In addition, the study has identified research and development requirements essential to facilitate the successful conceptual design, construction, and operation of an EPR

  2. Beam heating requirements for a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Bertoncini, P.J.; Brooks, J.N.; Fasolo, J.A.; Stacey, W.M. Jr.

    1976-01-01

    Typical beam heating requirements for effective tokamak experimental power reactor (TEPR) operation have been studied in connection with the Argonne preliminary conceptual TEPR design. For an ignition level plasma (approximately 100 MWt fusion power) for the nominal case envisioned, the neutral beam is only used to heat the plasma to ignition. This typically requires a beam power output of 40 MW at 180 keV for about 3 sec with a total energy of 114 MJ supplied to the plasma. The beam requirements for an ignition device are not very sensitive to changes in wall-sputtered impurity levels or plasma resistivity. For a plasma that must be driven due to poor confinement, the beam must remain on for most of the burn cycle. For representative cases, beam powers of approximately 23 MW are required for a total on-time of 20 to 50 sec. Reqirements on power level, beam energy, on-time, and beam-generation efficiency all represent considerable advances over present technology. For the Argonne TEPR design, a total of 16 to 32 beam injectors is envisioned. For a 40-MW, 180-keV, one-component beam, each injector supplies about 7 to 14 A of neutrals to the plasma. For positive ion sources, about 50 to 100 A of ions are required per injector and some form of particle and/or energy recycling appears to be essential in order to meet the power and efficiency requirements

  3. Tokamak experimental power reactor conceptual design. Volume I

    International Nuclear Information System (INIS)

    1976-08-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters

  4. Space-time statistics of the turbulence in the PRETEXT and TEXT tokamak edge plasmas

    International Nuclear Information System (INIS)

    Levinson, S.J.

    1986-01-01

    A study of the statistical space-time properties of the turbulence observed in the edge regions of the PRETEXT and the TEXT tokamaks are reported. Computer estimates of the particle-transport spectrum T(omega), and of the local wavenumber-frequency spectra S(K,omega) for poloidal (k/sub y/) and toroidal (k/sub z/) wavenumbers was determined. A conventional fast-Fourier-transform technique is used initially for the analyses of the potential and density fluctuations obtained from spatially fixed Langmuir-probe pairs. Measurements of the fluctuation-induced particle transport revealed that the particle flux is outward for both PRETEXT and TEXT, and it results primarily from the low-frequency, long-wavelength components of the turbulence. The S(K/sub y/, omega) spectra are dominated by low frequencies ( -1 ) and appear broadened about an approximately linear statistical dispersion relation, anti k(omega). The broadening is characterized by a spectral width sigma/sub k/(omega) (rms deviation about anti k(omega)). In PRETEXT, sigma/sub k/(omega) is of the order of anti k(omega), and the turbulence appears to propagate poloidally with an apparent mean phase velocity of 1-2 x 10 5 cm/s in the ion diamagnetic drift direction. In TEXT, a reversal in the phase velocity of the turbulence in the edge plasma was observed

  5. Design of the 2D electron cyclotron emission imaging instrument for the J-TEXT tokamak.

    Science.gov (United States)

    Pan, X M; Yang, Z J; Ma, X D; Zhu, Y L; Luhmann, N C; Domier, C W; Ruan, B W; Zhuang, G

    2016-11-01

    A new 2D Electron Cyclotron Emission Imaging (ECEI) diagnostic is being developed for the J-TEXT tokamak. It will provide the 2D electron temperature information with high spatial, temporal, and temperature resolution. The new ECEI instrument is being designed to support fundamental physics investigations on J-TEXT including MHD, disruption prediction, and energy transport. The diagnostic contains two dual dipole antenna arrays corresponding to F band (90-140 GHz) and W band (75-110 GHz), respectively, and comprises a total of 256 channels. The system can observe the same magnetic surface at both the high field side and low field side simultaneously. An advanced optical system has been designed which permits the two arrays to focus on a wide continuous region or two radially separate regions with high imaging spatial resolution. It also incorporates excellent field curvature correction with field curvature adjustment lenses. An overview of the diagnostic and the technical progress including the new remote control technique are presented.

  6. Design of the 2D electron cyclotron emission imaging instrument for the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pan, X. M.; Yang, Z. J., E-mail: yangzj@hust.edu.cn; Ma, X. D.; Ruan, B. W.; Zhuang, G. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan, Hubei 430074 (China); Zhu, Y. L. [School of Physics, University of Science and Technology of China, Anhui 230026 (China); Luhmann, N. C.; Domier, C. W. [Davis Millimeter Wave Research Center, University of California, Davis, California 95616 (United States)

    2016-11-15

    A new 2D Electron Cyclotron Emission Imaging (ECEI) diagnostic is being developed for the J-TEXT tokamak. It will provide the 2D electron temperature information with high spatial, temporal, and temperature resolution. The new ECEI instrument is being designed to support fundamental physics investigations on J-TEXT including MHD, disruption prediction, and energy transport. The diagnostic contains two dual dipole antenna arrays corresponding to F band (90-140 GHz) and W band (75-110 GHz), respectively, and comprises a total of 256 channels. The system can observe the same magnetic surface at both the high field side and low field side simultaneously. An advanced optical system has been designed which permits the two arrays to focus on a wide continuous region or two radially separate regions with high imaging spatial resolution. It also incorporates excellent field curvature correction with field curvature adjustment lenses. An overview of the diagnostic and the technical progress including the new remote control technique are presented.

  7. Characteristics of edge-localized modes in the experimental advanced superconducting tokamak (EAST)

    DEFF Research Database (Denmark)

    Jiang, M.; Xu, G.S.; Xiao, C.

    2012-01-01

    Edge-localized modes (ELMs) are the focus of tokamak edge physics studies because the large heat loads associated with ELMs have great impact on the divertor design of future reactor-grade tokamaks such as ITER. In the experimental advanced superconducting tokamak (EAST), the first ELMy high...... confinement modes (H-modes) were obtained with 1 MW lower hybrid wave power in conjunction with wall conditioning by lithium (Li) evaporation and real-time Li powder injection. The ELMs in EAST at this heating power are mostly type-III ELMs. They were observed close to the H-mode threshold power and produced...

  8. Oak Ridge Tokamak experimental power reactor study reference design

    International Nuclear Information System (INIS)

    Roberts, M.; Bettis, E.S.

    1975-11-01

    A Tokamak EPR Reference Design is presented as a basis for further design study leading to a Conceptual Design. The set of basic plasma parameters selected--minor radius of 2.25 m, major radius of 6.75 m, magnetic field on axis of 4.8 T and plasma current of 7.2 MA--should produce a reactor-grade plasma with a significant neutron flux, even with the great uncertainty in plasma physics scaling from present experience to large sizes. Neutronics and heat transfer calculations coupled with mechanical design and materials considerations were used to develop a blanket and shield capable of operating at high temperature, protecting the surrounding coils, being maintained remotely and, in a few experimental modules, breeding tritium. Nb 3 Sn and NbTi superconductors are used in the toroidal field coil design. The coil system was developed for a maximum field of 11 T at the winding (to give a field on axis of 4.8 T), and combines multifilamentary superconducting cable with forced flow of supercritical helium enclosed in a steel conduit. The structural system uses a stainless steel center bucking ring and intercoil box beam bracing to provide rigid support for coils against the centering force, overturning moments from poloidal fields and faults, other external forces, and thermal stresses. The poloidal magnetics system is specially designed both to reduce the total volt-second energy requirements and to reduce the magnitude of the rate of field change at the toroidal field coils. The rate of field change imposed upon the toroidal field coils is reduced by at least a factor of 3.3 compared to that due to the plasma alone. Tritium processing, tritium containment and vacuum systems employ double containment and atmospheric cleanup to minimize releases. The document also contains discussions of systems integration and assembly, key research and development needs, and schedule considerations

  9. 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

  10. Design of FPGA based high-speed data acquisition and real-time data processing system on J-TEXT tokamak

    International Nuclear Information System (INIS)

    Zheng, W.; Liu, R.; Zhang, M.; Zhuang, G.; Yuan, T.

    2014-01-01

    Highlights: • It is a data acquisition system for polarimeter–interferometer diagnostic on J-TEXT tokamak based on FPGA and PXIe devices. • The system provides a powerful data acquisition and real-time data processing performance. • Users can implement different data processing applications on the FPGA in a short time. • This system supports EPICS and has been integrated into the J-TEXT CODAC system. - Abstract: Tokamak experiment requires high-speed data acquisition and processing systems. In traditional data acquisition system, the sampling rate, channel numbers and processing speed are limited by bus throughput and CPU speed. This paper presents a data acquisition and processing system based on FPGA. The data can be processed in real-time before it is passed to the CPU. It provides processing ability for more channels with higher sampling rates than the traditional data acquisition system while ensuring deterministic real-time performance. A working prototype is developed for the newly built polarimeter–interferometer diagnostic system on the Joint Texas Experimental Tokamak (J-TEXT). It provides 16 channels with 120 MHz maximum sampling rate and 16 bit resolution. The onboard FPGA is able to calculate the plasma electron density and Faraday rotation angel. A RAID 5 storage device is adopted providing 700 MB/s read–write speed to buffer the data to the hard disk continuously for better performance

  11. Comparison between 2D turbulence model ESEL and experimental data from AUG and COMPASS tokamaks

    DEFF Research Database (Denmark)

    Ondac, Peter; Horacek, Jan; Seidl, Jakub

    2015-01-01

    In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtain...... for an extension of the ESEL model from 2D to 3D to fully resolve the parallel dynamics, and the coupling from the plasma to the sheath....

  12. Design of the power system for dynamic resonant magnetic perturbation coils on the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yi, B. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); Ding, Y.H., E-mail: yhding@mail.hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); Zhang, M.; Rao, B.; Nan, J.Y.; Zeng, W.B.; Zheng, M.Y.; Xu, H.Y.; Zhuang, G.; Pan, Y. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2013-10-15

    Highlights: ► We introduce the dynamic resonant magnetic perturbation coils system on J-TEXT. ► Details of the design of the power supply system have been presented. ► At DC mode, two antiparallel 6-pulse phase thyristor rectifiers were chosen. ► An AC–DC–AC converter including a series resonant inverter was adopted for AC mode. ► Some engineering testing result was given in this paper. -- Abstract: A set of in-vessel saddle coils has been installed on J-TEXT tokamak. They are proposed for further researches on controlling tearing modes and driving plasma rotation by static and dynamic resonant magnetic perturbations (RMPs). The saddle coils will be energized by DC with the amplitude up to 10 kA, or AC with maximum amplitude up to 5 kA within the frequency range of 1–5 kHz. At DC mode two antiparallel 6-pulse phase thyristor rectifiers are chosen to obtain bidirectional current, while at AC mode an AC–DC–AC converter including a series resonant inverter can generate current of various amplitudes and frequencies. The paper presents the design of the power supply system, based on the definition of the power supply requirements and the feasibility of implementation of the topology and control strategy. Some simulation and experimental results are given in the end.

  13. Accessibility and replacement as prime constraints in the design of large experimental tokamaks

    International Nuclear Information System (INIS)

    Challender, R.S.; Reynolds, P.

    1976-01-01

    An attempt is made to bring together those design features of large, experimental Tokamaks, which would lead to better accessibility during non-active operation and, in particular, permit replacement and repair after activation, thereby making possible an extended period of experimental operation into the ignition phase

  14. System engineering and design of a pulsed homopolar generator power supply for the Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Bird, W.L.; Grant, G.B.; Weldon, W.F.; Rylander, H.G.; Woodson, H.H.

    1977-01-01

    The design of a homopolar generator power supply for the Texas Experimental Tokamak (TEXT) is presented. Four series-connected disk type homopolar machines serve as inertial energy storage and conversion devices to supply 50 to 70 MW peak power to the toroidal field coil and ohmic heating coil circuits. The system is nominally operated at 150 MJ, 430 V to provide a 0.5 sec flat top, 160 kA TF current pulse and a 0.3 sec, 10 kA OH current pulse every 2.0 min on a continuous basis. The system has a maximum capacity of 200 MJ at a maximum open circuit voltage of 500 V. The homopolar machine design is described

  15. Breakdown assisted by a novel electron drift injection in the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Wang, Nengchao; Jin, Hai; Zhuang, Ge; Ding, Yonghua; Pan, Yuan; Cen, Yishun; Chen, Zhipeng; Huang, Hai; Liu, Dequan; Rao, Bo; Zhang, Ming; Zou, Bichen

    2014-01-01

    A novel electron drift injection (EDI) system aiming to improve breakdown behavior has been designed and constructed on the Joint Texas EXperiment Tokamak Tokamak. Electrons emitted by the system undergo the E×B drift, ∇B drift and curvature drift in sequence in order to traverse the confining magnetic field. A local electrostatic well, generated by a concave-shaped plate biased more negative than the cathode, is introduced to interrupt the emitted electrons moving along the magnetic field line (in the parallel direction) in an attempt to bring an enhancement of the injection efficiency and depth. A series of experiments have demonstrated the feasibility of this method, and a penetration distance deeper than 9.5 cm is achieved. Notable breakdown improvements, including the reduction of breakdown delay and average loop voltage, are observed for discharges assisted by EDI. The lower limit of successfully ionized pressure is expanded

  16. Experimental measurement of electron heat diffusivity in a tokamak

    International Nuclear Information System (INIS)

    Callen, J.D.; Jahns, G.L.

    1976-06-01

    The electron temperature perturbation produced by internal disruptions in the center of the Oak Ridge Tokamak (ORMAK) is followed with a multi-chord soft x-ray detector array. The space-time evolution is found to be diffusive in character, with a conduction coefficient larger by a factor of 2.5 - 15 than that implied by the energy containment time, apparently because it is a measurement for the small group of electrons whose energies exceed the cut-off energy of the detectors

  17. TIBER II: an upgraded tokamak igntion/burn experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Perkins, L.J.

    1986-01-01

    We are disIgning a minimum-size Tokamak ignition/Burn Reactor (TIBER II). This design incorporates physics requirements, neutron wall loading and fluence parameters that will make it compatible with a nuclear testing mission. Reactor relevant physics will be tested by using current drive and steady-state operation. Although the design accommodates several current drive options, including neutral beams, the base case uses a combination of lower hybrid and electron-cyclotron radio frequency power. Minimum neutron shielding, compact structures, high magnet-current densities, and remotely maintainable vacuum seals, all contribute to the compact size

  18. Comments on experimental results of energy confinement of tokamak plasmas

    International Nuclear Information System (INIS)

    Chu, T.K.

    1989-04-01

    The results of energy-confinement experiments on steady-state tokamak plasmas are examined. For plasmas with auxiliary heating, an analysis based on the heat diffusion equation is used to define heat confinement time (the incremental energy confinement time). For ohmically sustained plasmas, experiments show that the onset of the saturation regime of energy confinement, marfeing, detachment, and disruption are marked by distinct values of the parameter /bar n//sub e///bar j/. The confinement results of the two types of experiments can be described by a single surface in 3-dimensional space spanned by the plasma energy, the heating power, and the plasma density: the incremental energy confinement time /tau//sub inc/ = ΔW/ΔP is the correct concept for describing results of heat confinement in a heating experiment; the commonly used energy confinement time defined by /tau//sub E/ = W/P is not. A further examination shows that the change of edge parameters, as characterized by the change of the effective collision frequency ν/sub e/*, governs the change of confinement properties. The totality of the results of tokamak experiments on energy confinement appears to support a hypothesis that energy transport is determined by the preservation of the pressure gradient scale length. 70 refs., 6 figs., 1 tab

  19. Experimental test of far-infrared polarimetry for Faraday rotation measurements on the TFR 600 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Soltwisch, H [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Plasmaphysik; Association Euratom-Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.)); Equipe, T F.R. [Association Euratom-CEA sur la Fusion, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France). Dept. de Recherches sur la Fusion Controlee

    1981-09-01

    The results are reported on the feasibility of using far-infrared polarimetry for Faraday rotation diagnostic measurements on the TRF Tokamak. Precise quantitative results were not obtained but a satisfactory agreement with a simple theoretical model leads to a good understanding of the experimental limitations of the method.

  20. Experimental observation of current generation by asymmetrical heating of ions in a tokamak plasma

    International Nuclear Information System (INIS)

    Gahl, J.; Ishihara, O.; Wong, K.L.; Kristiansen, M.; Hagler, M.

    1986-01-01

    The first experimental observation of current generation by asymmetrical heating of ions is reported. Ions were asymmetrically heated by a unidirectional fast Alfven wave launched by a slow wave antenna inside a tokamak. Current generation was detected by measuring the asymmetry of the toroidal plasma current with probes at the top and bottom of the toroidal plasma column

  1. Wavelength calibration of x-ray imaging crystal spectrometer on Joint Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Yan, W.; Chen, Z. Y.; Jin, W.; Huang, D. W.; Ding, Y. H.; Li, J. C.; Zhang, X. Q.; Zhuang, G.; Lee, S. G.; Shi, Y. J.

    2014-01-01

    The wavelength calibration of x-ray imaging crystal spectrometer is a key issue for the measurements of plasma rotation. For the lack of available standard radiation source near 3.95 Å and there is no other diagnostics to measure the core rotation for inter-calibration, an indirect method by using tokamak plasma itself has been applied on joint Texas experimental tokamak. It is found that the core toroidal rotation velocity is not zero during locked mode phase. This is consistent with the observation of small oscillations on soft x-ray signals and electron cyclotron emission during locked-mode phase

  2. A cryogenic system for TIBER II [Tokamak Ignition/Burn Experimental Reactor

    International Nuclear Information System (INIS)

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

    1987-01-01

    Phase II of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) study describes one option for a small, economical, next-generation tokamak [1,2]. Because of its small size, minimum shielding is used between the plasma and the toroidal-field (TF) coils. Consequently, a large cryogenic system (approximately 70 kW at 4.5 K) capable of delivering forced-flow helium is required. This paper describes a cryogenic system that meets this requirement and includes TIBER-II requirements. 3 refs

  3. Experimental study of external kink instabilities in the Columbia High Beta Tokamak

    International Nuclear Information System (INIS)

    Ivers, T.H.

    1991-01-01

    The generation of power through controlled thermonuclear fusion reactions in a magnetically confined plasma holds promise as a means of supplying mankind's future energy needs. The device most technologically advanced in pursuit of this goal is the tokamak, a machine in which a current-carrying toroidal plasma is thermally isolated from its surroundings by a strong magnetic field. To be viable, the tokamak reactor must produce a sufficiently large amount of power relative to that needed to sustain the fusion reactions. Plasma instabilities may severely limit this possibility. In this work, I describe experimental measurements of the magnetic structure of large-scale, rapidly-growing instabilities that occur in a tokamak when the current or pressure of the plasma exceeds a critical value relative to the magnetic field, and I compare these measurements with theoretical predictions

  4. Overview of recent experimental results from the Aditya tokamak

    Science.gov (United States)

    Tanna, R. L.; Ghosh, J.; Chattopadhyay, P. K.; Raj, Harshita; Patel, Sharvil; Dhyani, P.; Gupta, C. N.; Jadeja, K. A.; Patel, K. M.; Bhatt, S. B.; Panchal, V. K.; Patel, N. C.; Chavda, Chhaya; Praveenlal, E. V.; Shah, K. S.; Makawana, M. N.; Jha, S. K.; Gopalkrishana, M. V.; Tahiliani, K.; Sangwan, Deepak; Raju, D.; Nagora, Umesh; Pathak, S. K.; Atrey, P. K.; Purohit, S.; Raval, J.; Joisa, Y. S.; Rao, C. V. S.; Chowdhuri, M. B.; Banerjee, S.; Ramaiya, N.; Manchanda, R.; Thomas, J.; Kumar, Ajai; Ajay, Kumar; Sharma, P. K.; Kulkarni, S. V.; Sathyanarayana, K.; Shukla, B. K.; Das, Amita; Jha, R.; Saxena, Y. C.; Sen, A.; Kaw, P. K.; Bora, D.; the ADITYA Team

    2017-10-01

    Several experiments, related to controlled thermonuclear fusion research and highly relevant for large size tokamaks, including ITER, have been carried out in ADITYA, an ohmically heated circular limiter tokamak. Repeatable plasma discharges of a maximum plasma current of ~160 kA and discharge duration beyond ~250 ms with a plasma current flattop duration of ~140 ms have been obtained for the first time in ADITYA. The reproducibility of the discharge reproducibility has been improved considerably with lithium wall conditioning, and improved plasma discharges are obtained by precisely controlling the position of the plasma. In these discharges, chord-averaged electron density ~3.0-4.0  ×  1019 m-3 using multiple hydrogen gas puffs, with a temperature of the order of ~500-700 eV, have been achieved. Novel experiments related to disruption control are carried out and disruptions, induced by hydrogen gas puffing, are successfully mitigated using the biased electrode and ion cyclotron resonance pulse techniques. Runaway electrons are successfully mitigated by applying a short local vertical field (LVF) pulse. A thorough disruption database has been generated by identifying the different categories of disruption. Detailed analysis of several hundred disrupted discharges showed that the current quench time is inversely proportional to the q edge. Apart from this, for volt-sec recovery during the plasma formation phase, low loop voltage start-up and current ramp-up experiments have been carried out using electron cyclotron resonance heating (ECRH). Successful recovery of volt-sec leads to the achievement of longer plasma discharge durations. In addition, the neon gas puff assisted radiative improved confinement mode has also been achieved in ADITYA. All of the above mentioned experiments will be discussed in this paper.

  5. Numerical simulation of plasma response to externally applied resonant magnetic perturbation on the J-TEXT tokamak

    Science.gov (United States)

    Bicheng, LI; Zhonghe, JIANG; Jian, LV; Xiang, LI; Bo, RAO; Yonghua, DING

    2018-05-01

    Nonlinear magnetohydrodynamic (MHD) simulations of an equilibrium on the J-TEXT tokamak with applied resonant magnetic perturbations (RMPs) are performed with NIMROD (non-ideal MHD with rotation, open discussion). Numerical simulation of plasma response to RMPs has been developed to investigate magnetic topology, plasma density and rotation profile. The results indicate that the pure applied RMPs can stimulate 2/1 mode as well as 3/1 mode by the toroidal mode coupling, and finally change density profile by particle transport. At the same time, plasma rotation plays an important role during the entire evolution process.

  6. Experimental modeling of eddy currents and deflections for tokamak limiters

    International Nuclear Information System (INIS)

    Hua, T.Q.; Knott, M.J.; Turner, L.R.; Wehrle, R.B.

    1986-01-01

    In this study, experiments were performed to investigate deflection, current, and material stress in cantilever beams with the Fusion ELectromagnetic Induction eXperiment (FELIX) at the Argonne National Laboratory. Since structures near the plasma are typically cantilevered, the beams provide a good model for the limiter blades of a tokamak fusion reactor. The test pieces were copper, aluminum, phosphor bronze, and brass cantilever beams, clamped rigidly at one end with a nonconducting support frame inside the FELIX test volume. The primary data recorded as functions of time were the beam deflection measured with a noncontact electro-optical device, the total eddy current measured with a Rogowski coil and linking through a central hole in the beam, and the material stress extracted from strain gauges. Measurements of stress and deflection were taken at selected positions along the beam. The extent of the coupling effect depends on several factors. These include the size, the electrical and mechanical properties of the beam, segmenting of the beam, the decay rate of the dipole field, and the strength of the solenoid field

  7. Texas Experimental Tokamak, a plasma research facility: Technical progress report

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1995-08-01

    In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively

  8. Development of hard X-ray spectrometer with high time resolution on the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Ma, T.K.; Chen, Z.Y., E-mail: zychen@hust.edu.cn; Huang, D.W.; Tong, R.H.; Yan, W.; Wang, S.Y.; Dai, A.J.; Wang, X.L.

    2017-06-01

    A hard X-ray (HXR) spectrometer has been developed to study the runaway electrons during the sawtooth activities and during the runaway current plateau phase on the J-TEXT tokamak. The spectrometer system contains four NaI scintillator detectors and a multi-channel analyzer (MCA) with 0.5 ms time resolution. The dedicated peak detection circuit embedded in the MCA provides a pulse height analysis at count rate up to 1.2 million counts per second (Mcps), which is the key to reach the high time resolution. The accuracy and reliability of the system have been verified by comparing with the hardware integrator of HXR flux. The temporal evolution of HXR flux in different energy ranges can be obtained with high time resolution by this dedicated HXR spectrometer. The response of runaway electron transport with different energy during the sawtooth activities can be studied. The energy evolution of runaway electrons during the plateau phase of runaway current can be obtained. - Highlights: • A HXR spectrometer with high time resolution has been developed on J-TEXT tokamak. • The response of REs transport during the sawtooth activities can be investigated. • The energy evolution of REs following the disruptions can be monitored.

  9. Accessibility and replacement as prime constraints in the design of large experimental Tokamaks

    International Nuclear Information System (INIS)

    Challender, R.S.; Reynolds, P.

    1976-01-01

    Many of the designs being developed for large Tokamak experiments are based on the classical geometry of the small machines in laboratories throughout the world: a circular array of coils interlinked by an inner toroidal vessel. An attempt is made to bring together those design features which would lead to better accessibility during non-active operation and in particular permit replacement and repair after activation, thereby making possible an extended period of experimental operation into the ignition phase

  10. Absolute intensity calibration of the 32-channel heterodyne radiometer on experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.; Zhao, H. L.; Liu, Y., E-mail: liuyong@ipp.ac.cn; Li, E. Z.; Han, X.; Ti, A.; Hu, L. Q.; Zhang, X. D. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Domier, C. W.; Luhmann, N. C. [Department of Electrical and Computer Engineering, University of California at Davis, Davis, California 95616 (United States)

    2014-09-15

    This paper presents the results of the in situ absolute intensity calibration for the 32-channel heterodyne radiometer on the experimental advanced superconducting tokamak. The hot/cold load method is adopted, and the coherent averaging technique is employed to improve the signal to noise ratio. Measured spectra and electron temperature profiles are compared with those from an independent calibrated Michelson interferometer, and there is a relatively good agreement between the results from the two different systems.

  11. Electron cyclotron emission radiometer upgrade on the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Z. J.; Pan, X. M., E-mail: panxiaoming@hust.edu.cn; Ma, X. D.; Ruan, B. W.; Zhou, R. B.; Zhang, C. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan, Hubei 430074 (China)

    2016-11-15

    To meet experimental requirements, the J-TEXT electron cyclotron emission (ECE) diagnostic is being upgraded. The front end antenna and transmission line have been modified and a new 8-channel W-band detecting unit has been developed. The improved ECE system will extend the frequency range from 94.5-124.5 GHz to 80.5-124.5 GHz. This will enable the system to cover the most plasma in the radius direction for B{sub T} = 1.8–2.2 T, and it even can cover almost the whole plasma range ρ = − 0.8–0.9 (minus means the high field side) at B{sub T} = 1.8 T. A new auxiliary channel bank with 8 narrow band, tunable yttrium iron garnet filters is planned to add to the ECE system. Due to observations along a major radius, perpendicular to B{sub T}, and relatively low electron temperature, Doppler and relativistic broadening are minimal and thus high spatial resolution measurements can be made at variable locations with these tunable channels.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wootton, A.J.

    1994-07-01

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

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

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1994-07-01

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

  14. First experimental results with the Current Limit Avoidance System at the JET tokamak

    Energy Technology Data Exchange (ETDEWEB)

    De Tommasi, G. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Galeani, S. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Jachmich, S. [Association EURATOM-Belgian State, Koninklijke Militaire School - Ecole Royale Militaire, B-1000 Brussels (Belgium); Joffrin, E. [IRFM-CEA, Centre de Cadarache, 13108 Saint-paul-lez-Durance (France); Lennholm, M. [EFDA Close Support Unit, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); European Commission, B-1049 Brussels (Belgium); Lomas, P.J. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Neto, A.C. [Associazione EURATOM-IST, Instituto de Plasmas e Fusao Nuclear, IST, 1049-001 Lisboa (Portugal); Maviglia, F. [Associazione EURATOM-ENEA-CREATE, Via Claudio 21, 80125 Napoli (Italy); McCullen, P. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Pironti, A. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Rimini, F.G. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Sips, A.C.C. [European Commission, B-1049 Brussels (Belgium); Varano, G.; Vitelli, R. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Zaccarian, L. [CNRS, LAAS, 7 Avenue du Colonel Roche, F-31400 Toulouse (France); Universitè de Toulouse, LAAS, F-31400 Toulouse (France)

    2013-06-15

    The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.

  15. Experimental investigation of turbulent transport at the edge of a tokamak plasma

    International Nuclear Information System (INIS)

    Fedorczak, N.

    2010-01-01

    This manuscript is devoted to the experimental investigation of particle transport in the edge region of the tokamak Tore Supra. The first part introduces the motivations linked to energy production, the principle of a magnetic confinement and the elements of physics essential to describe the dynamic of the plasma at the edge region. From data collected by a set of Langmuir probes and a fast visible imaging camera, we demonstrate that the particle transport is dominated by the convection of plasma filaments, structures elongated along magnetic field lines. They present a finite wave number, responsible for the high enhancement of the particle flux at the low field side of the tokamak. This leads to the generation of strong parallel flows, and the strong constraint of filament geometry by the magnetic shear. (author)

  16. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-11-01

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  17. Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

    International Nuclear Information System (INIS)

    Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

    1976-01-01

    This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended

  18. Superconducting coil design for a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Turner, L.R.; Wang, S.T.; Smelser, P.

    1977-01-01

    Superconducting toroidal field (TF) and polodial-field (PF) coils have been designed for the proposed Argonne National Laboratory experimental power reactor (EPR). Features of the design include: (1) Peak field of 8 T at 4.2 K or 10 T at 3.0 K. (2) Constant-tension shape for the TF coils, corrected for the finite number (16) of coils. (3) Analysis of errors in coil alignment. (4) Comparison of safety aspects of series-connected and parallel-connected coils. (5) A 60 kA sheet conductor of NbTi with copper stabilizer and stainless steel for support. (6) Superconducting PF coils outside the TF coils. (7) The TF coils shielded from pulsed fields by high-purity aluminum

  19. Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

    International Nuclear Information System (INIS)

    Onozuka, M.; Takeda, N.; Nakahira, M.; Shimizu, K.; Nakamura, T.

    2003-01-01

    The most recent assessment method to evaluate the dynamic behavior of the International Thermonuclear Experimental Reactor (ITER) tokamak assembly is outlined. Three experimental models, including a 1/5.8-scale tokamak model, have been considered to validate the numerical analysis methods for dynamic events, particularly seismic ones. The experimental model has been evaluated by numerical calculations and the results are presented. In the calculations, equivalent linearization has been applied for the non-linear characteristics of the support flange connection, caused by the effects of the bolt-fastening and the friction between the flanges. The detailed connecting conditions for the support flanges have been developed and validated for the analysis. Using the conditions, the eigen-mode analysis has shown that the first and second eigen-mode are horizontal vibration modes with the natural frequency of 39 Hz, while the vertical vibration mode is the fourth mode with the natural frequency of 86 Hz. Dynamic analysis for seismic events has shown the maximum acceleration of approximately twofold larger than that of the applied acceleration, and the maximum stress of 104 MPa found in the flange connecting bolt. These values will be examined comparing with experimental results in order to validate the analysis methods

  20. Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: masanori_onozuka@mhi.co.jp; Takeda, N.; Nakahira, M.; Shimizu, K.; Nakamura, T

    2003-09-01

    The most recent assessment method to evaluate the dynamic behavior of the International Thermonuclear Experimental Reactor (ITER) tokamak assembly is outlined. Three experimental models, including a 1/5.8-scale tokamak model, have been considered to validate the numerical analysis methods for dynamic events, particularly seismic ones. The experimental model has been evaluated by numerical calculations and the results are presented. In the calculations, equivalent linearization has been applied for the non-linear characteristics of the support flange connection, caused by the effects of the bolt-fastening and the friction between the flanges. The detailed connecting conditions for the support flanges have been developed and validated for the analysis. Using the conditions, the eigen-mode analysis has shown that the first and second eigen-mode are horizontal vibration modes with the natural frequency of 39 Hz, while the vertical vibration mode is the fourth mode with the natural frequency of 86 Hz. Dynamic analysis for seismic events has shown the maximum acceleration of approximately twofold larger than that of the applied acceleration, and the maximum stress of 104 MPa found in the flange connecting bolt. These values will be examined comparing with experimental results in order to validate the analysis methods.

  1. Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

    International Nuclear Information System (INIS)

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-01-01

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology

  2. 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

  3. Realization of toroidal field power supply control system for J-TEXT tokamak

    International Nuclear Information System (INIS)

    Qiu Shengshun; Zhuang Ge; Zhang Ming; Feng Jianming

    2009-01-01

    Based on the integrated development environment provided by QNX real-time operation system, the control system of toroidal field power supply is designed and developed. The system is proved to be reliable, stable and in real-time. It can control the power supply successfully to produce a constant current up to 92.5kA lasting for 1s and 1.74T at the magnetic axis. In conclusion, the control system can meet the requirements of the J-TEXT routine operation at present. (authors)

  4. Plasma response to m/n  =  3/1 resonant magnetic perturbation at J-TEXT Tokamak

    Science.gov (United States)

    Hu, Qiming; Li, Jianchao; Wang, Nengchao; Yu, Q.; Chen, Jie; Cheng, Zhifeng; Chen, Zhipeng; Ding, Yonghua; Jin, Hai; Li, Da; Li, Mao; Liu, Yang; Rao, Bo; Zhu, Lizhi; Zhuang, Ge; the J-TEXT Team

    2016-09-01

    The influence of resonant magnetic perturbations (RMPs) with a large m/n  =  3/1 component on electron density has been studied at J-TEXT tokamak by using externally applied static and rotating RMPs, where m and n are the poloidal and toroidal mode number, respectively. The detailed time evolution of electron density profile, measured by the polarimeter-interferometer, shows that the electron density n e first increases (decreases) inside (around/outside) of the 3/1 rational surface (RS), and it is increased globally later together with enhanced edge recycling. Associated with field penetration, the toroidal rotation around the 3/1 RS is accelerated in the co-I p direction and the poloidal rotation is changed from the electron to ion diamagnetic drift direction. Spontaneous unlocking-penetration circles occur after field penetration if the RMPs amplitude is not strong enough. For sufficiently strong RMPs, the 2/1 locked mode is also triggered due to mode coupling, and the global density is increased. The field penetration threshold is found to be linearly proportional to n eL (line-integrated density) at the 3/1 RS but to (n eL)0.73 for n e at the plasma core. In addition, for rotating RMPs with a large 3/1 component, field penetration causes a global increase in electron density.

  5. Data acquisition, processing and display of experimental data for the Tokamak de Varennes

    International Nuclear Information System (INIS)

    Robins, E.S.; Larsen, J.M.; Lee, A.; Somers, G.

    1985-01-01

    The Tokamak de Varennes is to be a national facility for research into magnetic nuclear fusion. A centralised computer system is currently under development to facilitate the remote control, acquisition, processing and display of experimental data. The software (GALE-V) consists of a set of tasks to build data structures which mirror the physical arrangement of each experiment and provide the bases for the interpretation and presentation of the data to each experimenter. Data retrieval is accomplished through the graphics subsystem, and an interface for user-written data processing programs allows for the varied needs of data analysis of each experiment. Other facilities being developed provide the tools for a user to retrieve, process and view the data in a simple manner

  6. The upgrade of the J-TEXT experimental data access and management system

    International Nuclear Information System (INIS)

    Yang, C.; Zhang, M.; Zheng, W.; Liu, R.; Zhuang, G.

    2014-01-01

    Highlights: • The J-TEXT DAMS is developed based on B/S model, which makes it conveniently access the system. • The JWeb-Scope adopts segment strategy to read data that improve the speed of reading data. • DAMS have integrated the management and JWeb-Scope and make an easy way for visitors to access the experiment data. • The JWeb-Scope can be visited all over the world, plot experiment data and zoom in or out smoothly. - Abstract: The experimental data of J-TEXT tokamak are stored in the MDSplus database. The old J-TEXT data access system is based on the tools provided by MDSplus. Since the number of signals is huge, the data retrieval for an experiment is difficult. To solve this problem, the J-TEXT experimental data access and management system (DAMS) based on MDSplus has been developed. The DAMS left the old MDSplus system unchanged providing new tools, which can help users to handle all signals as well as to retrieve signals they need thanks to the user information requirements. The DAMS also offers users a way to create their jScope configuration files which can be downloaded to the local computer. In addition, the DAMS provides a JWeb-Scope tool to visualize the signal in a browser. JWeb-Scope adopts segment strategy to read massive data efficiently. Users can plot one or more signals on their own choice and zoom-in, zoom-out smoothly. The whole system is based on B/S model, so that the users only need of the browsers to access the DAMS. The DAMS has been tested and it has a better user experience. It will be integrated into the J-TEXT remote participation system later

  7. Conversion of magnetic energy to runaway kinetic energy during the termination of runaway current on the J-TEXT tokamak

    Science.gov (United States)

    Dai, A. J.; Chen, Z. Y.; Huang, D. W.; Tong, R. H.; Zhang, J.; Wei, Y. N.; Ma, T. K.; Wang, X. L.; Yang, H. Y.; Gao, H. L.; Pan, Y.; the J-TEXT Team

    2018-05-01

    A large number of runaway electrons (REs) with energies as high as several tens of mega-electron volt (MeV) may be generated during disruptions on a large-scale tokamak. The kinetic energy carried by REs is eventually deposited on the plasma-facing components, causing damage and posing a threat on the operation of the tokamak. The remaining magnetic energy following a thermal quench is significant on a large-scale tokamak. The conversion of magnetic energy to runaway kinetic energy will increase the threat of runaway electrons on the first wall. The magnetic energy dissipated inside the vacuum vessel (VV) equals the decrease of initial magnetic energy inside the VV plus the magnetic energy flowing into the VV during a disruption. Based on the estimated magnetic energy, the evolution of magnetic-kinetic energy conversion are analyzed through three periods in disruptions with a runaway current plateau.

  8. The study of heat flux for disruption on experimental advanced superconducting tokamak

    International Nuclear Information System (INIS)

    Yang, Zhendong; Fang, Jianan; Luo, Jiarong; Cui, Zhixue; Gong, Xianzu; Gan, Kaifu; Zhao, Hailin; Zhang, Bin; Chen, Meiwen

    2016-01-01

    Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dR_s_e_p = −2 cm, while it changes to upper single null (dR_s_e_p = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m"2.

  9. The study of heat flux for disruption on experimental advanced superconducting tokamak

    Science.gov (United States)

    Yang, Zhendong; Fang, Jianan; Gong, Xianzu; Gan, Kaifu; Luo, Jiarong; Zhao, Hailin; Cui, Zhixue; Zhang, Bin; Chen, Meiwen

    2016-05-01

    Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dRsep = -2 cm, while it changes to upper single null (dRsep = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m2.

  10. Phenomena of non-thermal electrons from the X-ray imaging crystal spectrometer on J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yan, W. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan (China); Chen, Z.Y., E-mail: zychen@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan (China); Jin, W. [Center of Interface Dynamics for Sustainability, China Academy of Engineering Physics, Chengdu 610200, Sichuan (China); Huang, D.W. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan (China); Lee, S.G.; Shi, Y.J. [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Tong, R.H.; Wang, S.Y.; Wei, Y.N.; Ma, T.K.; Zhuang, G. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan (China)

    2016-11-01

    Highlights: • Some lines from X-ray imaging crystal spectrometer (XICS) can be enhanced by non-thermal electrons, such as q, r satellite lines and z lines. • Analyze the non-thermal phenomena can reduce the error of electron temperature deduced from the intensity ratio of different lines of the He-like argon spectra from XICS. • XICS can be a tool to measure the non-thermal phenomena from these enhanced lines. - Abstract: A high spectra resolution X-ray imaging crystal spectrometer has been implemented on J-TEXT Tokamak for the measurements of K{sub α} spectra of helium-like argon and its satellite lines. The wavelength range of K{sub α} spectra of helium-like argon is from 3.9494 Å to 3.9944 Å that includes the resonance line w, intercombination lines x and y, forbidden line z and numerous satellite lines, referenced using standard Gabriel notation. In low-density discharge, the intensity of q, r satellite lines and z lines can be significantly enhanced by non-thermal electrons. Non-thermal electrons are produced due to the low plasma density. The high hard X-ray flux from NaI detector and significant downshift electron cyclotron emissions from energetic runaway electrons also indicated that there is a large population of runaway electrons in the low-density discharge. The non-thermal part of electrons can affect the excitation/transition equilibrium or ionization/recombination equilibrium. The q line is mainly produced by inner-shell excitation of lithium-like argon, and the r line is partially produced by inner-shell excitation of lithium-like argon and dielectronic recombination of helium-like argon.

  11. Phenomena of non-thermal electrons from the X-ray imaging crystal spectrometer on J-TEXT tokamak

    International Nuclear Information System (INIS)

    Yan, W.; Chen, Z.Y.; Jin, W.; Huang, D.W.; Lee, S.G.; Shi, Y.J.; Tong, R.H.; Wang, S.Y.; Wei, Y.N.; Ma, T.K.; Zhuang, G.

    2016-01-01

    Highlights: • Some lines from X-ray imaging crystal spectrometer (XICS) can be enhanced by non-thermal electrons, such as q, r satellite lines and z lines. • Analyze the non-thermal phenomena can reduce the error of electron temperature deduced from the intensity ratio of different lines of the He-like argon spectra from XICS. • XICS can be a tool to measure the non-thermal phenomena from these enhanced lines. - Abstract: A high spectra resolution X-ray imaging crystal spectrometer has been implemented on J-TEXT Tokamak for the measurements of K_α spectra of helium-like argon and its satellite lines. The wavelength range of K_α spectra of helium-like argon is from 3.9494 Å to 3.9944 Å that includes the resonance line w, intercombination lines x and y, forbidden line z and numerous satellite lines, referenced using standard Gabriel notation. In low-density discharge, the intensity of q, r satellite lines and z lines can be significantly enhanced by non-thermal electrons. Non-thermal electrons are produced due to the low plasma density. The high hard X-ray flux from NaI detector and significant downshift electron cyclotron emissions from energetic runaway electrons also indicated that there is a large population of runaway electrons in the low-density discharge. The non-thermal part of electrons can affect the excitation/transition equilibrium or ionization/recombination equilibrium. The q line is mainly produced by inner-shell excitation of lithium-like argon, and the r line is partially produced by inner-shell excitation of lithium-like argon and dielectronic recombination of helium-like argon.

  12. High resolution polarimeter-interferometer system for fast equilibrium dynamics and MHD instability studies on Joint-TEXT tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Li, Q.; Liu, Y.; Gao, L.; Zhou, Y. N.; Jian, X.; Xiong, C. Y.; Wang, Z. J. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); Brower, D. L.; Ding, W. X. [Department of Physics and Astronomy, University of California Los Angeles, Los Angeles, California 90095 (United States)

    2014-11-15

    A high-performance Faraday-effect polarimeter-interferometer system has been developed for the J-TEXT tokamak. This system has time response up to 1 μs, phase resolution < 0.1° and minimum spatial resolution ∼15 mm. High resolution permits investigation of fast equilibrium dynamics as well as magnetic and density perturbations associated with intrinsic Magneto-Hydro-Dynamic (MHD) instabilities and external coil-induced Resonant Magnetic Perturbations (RMP). The 3-wave technique, in which the line-integrated Faraday angle and electron density are measured simultaneously by three laser beams with specific polarizations and frequency offsets, is used. In order to achieve optimum resolution, three frequency-stabilized HCOOH lasers (694 GHz, >35 mW per cavity) and sensitive Planar Schottky Diode mixers are used, providing stable intermediate-frequency signals (0.5–3 MHz) with S/N > 50. The collinear R- and L-wave probe beams, which propagate through the plasma poloidal cross section (a = 0.25–0.27 m) vertically, are expanded using parabolic mirrors to cover the entire plasma column. Sources of systematic errors, e.g., stemming from mechanical vibration, beam non-collinearity, and beam polarization distortion are individually examined and minimized to ensure measurement accuracy. Simultaneous density and Faraday measurements have been successfully achieved for 14 chords. Based on measurements, temporal evolution of safety factor profile, current density profile, and electron density profile are resolved. Core magnetic and density perturbations associated with MHD tearing instabilities are clearly detected. Effects of non-axisymmetric 3D RMP in ohmically heated plasmas are directly observed by polarimetry for the first time.

  13. Theoretical and experimental studies of runaway electrons in the TEXTOR tokamak

    International Nuclear Information System (INIS)

    Abdullaev, S.S.; Finken, K.H.; Wongrach, K.; Willi, O.

    2016-01-01

    Theoretical and experimental studies of runaway electrons in tokamaks and their mitigations, particularly the recent studies performed by a group of the Heinrich-Heine University Duesseldorf in collaboration with the Institute of Energy and Climate Research of the Research Centre (Forschungszentrum) of Juelich are reviewed. The main topics focus on (i) runaway generation mechanisms, (ii) runaway orbits in equilibrium plasma, (iii) transport in stochastic magnetic fields, (iv) diagnostics and investigations of transport of runaway electron and their losses in low density discharges (v) runaway electrons during plasma disruptions, and (vi) runaway mitigation methods. The development of runaway diagnostics enables the measurement of runaway electrons in both the centre and edge of the plasma. The diagnostics provide an absolute runaway energy resolved measurement, the radial decay length of runaway electrons and, the structure and dynamics of runaway electron beams. The new mechanism of runaway electron formation during plasma disruptions is discussed.

  14. Inward particle transport at high collisionality in the Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Wang, G. Q.; Ma, J.; Weiland, J.; Zang, Q.

    2013-01-01

    We have made the first drift wave study of particle transport in the Experimental Advanced Superconducting Tokamak (Wan et al., Nucl. Fusion 49, 104011 (2009)). The results reveal that collisions make the particle flux more inward in the high collisionality regime. This can be traced back to effects that are quadratic in the collision frequency. The particle pinch is due to electron trapping which is not very efficient in the high collisionality regime so the approach to equilibrium is slow. We have included also the electron temperature gradient (ETG) mode to give the right electron temperature gradient, since the Trapped Electron Mode (TE mode) is weak in this regime. However, at the ETG mode number ions are Boltzmann distributed so the ETG mode does not give particle transport

  15. Theoretical and experimental studies of runaway electrons in the TEXTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Abdullaev, S.S.; Finken, K.H.; Wongrach, K.; Willi, O.

    2016-07-01

    Theoretical and experimental studies of runaway electrons in tokamaks and their mitigations, particularly the recent studies performed by a group of the Heinrich-Heine University Duesseldorf in collaboration with the Institute of Energy and Climate Research of the Research Centre (Forschungszentrum) of Juelich are reviewed. The main topics focus on (i) runaway generation mechanisms, (ii) runaway orbits in equilibrium plasma, (iii) transport in stochastic magnetic fields, (iv) diagnostics and investigations of transport of runaway electron and their losses in low density discharges (v) runaway electrons during plasma disruptions, and (vi) runaway mitigation methods. The development of runaway diagnostics enables the measurement of runaway electrons in both the centre and edge of the plasma. The diagnostics provide an absolute runaway energy resolved measurement, the radial decay length of runaway electrons and, the structure and dynamics of runaway electron beams. The new mechanism of runaway electron formation during plasma disruptions is discussed.

  16. Development of an alternating integrator for magnetic measurements for experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Liu, D. M., E-mail: dmliu@live.cn; Zhao, W. Z.; He, Y. G.; Chen, B. [School of Electrical Engineering and Automation, Hefei University of Technology, Hefei 230009 (China); Wan, B. N.; Shen, B.; Huang, J.; Liu, H. Q. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2014-11-15

    A high-performance integrator is one of the key electronic devices for reliably controlling plasma in the experimental advanced superconducting tokamak for long pulse operation. We once designed an integrator system of real-time drift compensation, which has a low integration drift. However, it is not feasible for really continuous operations due to capacitive leakage error and nonlinearity error. To solve the above-mentioned problems, this paper presents a new alternating integrator. In the new integrator, the integrator system of real-time drift compensation is adopted as one integral cell while two such integral cells work alternately. To achieve the alternate function, a Field Programmable Gate Array built in the digitizer is utilized. The performance test shows that the developed integrator with the integration time constant of 20 ms has a low integration drift (<15 mV) for 1000 s.

  17. Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor

    International Nuclear Information System (INIS)

    Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.

    1976-01-01

    As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield

  18. Advanced limiter test (ALT-I) in the TEXTOR tokamak - concept and experimental design

    International Nuclear Information System (INIS)

    Conn, R.W.; Grotz, S.P.; Prinja, A.K.

    1983-01-01

    The concept and experimental design of a pump-limiter for the TEXTOR tokamak is described. The module is constructed of stainless steel with a compound curvature head designed to limit the maximum heat flux to 300 W/cm 2 . The head is made of TiC-coated graphite containing a variable aperture slot to admit plasma to a deflector plate for ballistic pumping action. The assembly is actively pumped using Zr-Al getters with an estimated hydrogen pumping speed of 2x10 4 1/s. The aspect ratio of the pump duct and the length of the plasma channel are both variable to permit study of plasma plugging, ballistic scattering, and enhanced gas conduction effects. The module can be moved radially by 10 cm to permit its operation either as the primary or secondary limiter. Major diagnostics include Langmuir and solid state probes, bolometers, infrared thermography, thermocouples, ion gauges, manometers, and a gas mass analyzer. (author)

  19. 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

  20. Experimental and theoretical study of particle transport in the TCV Tokamak

    International Nuclear Information System (INIS)

    Fable, E.

    2009-06-01

    The main scope of this thesis work is to compare theoretical models with experimental observations on particle transport in particular regimes of plasma operation from the Tokamak à Configuration Variable (TCV) located at CRPP–EPFL in Lausanne. We introduce the main topics in Tokamak fusion research and the challenging problems in the first Chapter. A particular attention is devoted to the modelling of heat and particle transport. In the second Chapter the experimental part is presented, including an overview of TCV capabilities, a brief review of the relevant diagnostic systems, and a discussion of the numerical tools used to analyze the experimental data. In addition, the numerical codes that are used to interpret the experimental data and to compare them with theoretical predictions are introduced. The third Chapter deals with the problem of understanding the mechanisms that regulate the transport of energy in TCV plasmas, in particular in the electron Internal Transport Barrier (eITB) scenario. A radial transport code, integrated with an external module for the calculation of the turbulence-induced transport coefficients, is employed to reproduce the experimental scenario and to understand the physics at play. It is shown how the sustainment of an improved confinement regime is linked to the presence of a reversed safety factor profile. The improvement of confinement in the eITB regime is visible in the energy channel and in the particle channel as well. The density profile shows strong correlation with the temperature profile and has a large local logarithmic gradient. This is an important result obtained from the TCV eITB scenario analysis and is presented in the fourth Chapter. In the same chapter we present the estimate of the particle diffusion and convection coefficients obtained from density transient experiments performed in the eITB scenario. The theoretical understanding of the strong correlation between density and temperature observed in the e

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

    International Nuclear Information System (INIS)

    Evans, T.E.

    1984-09-01

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

  2. 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

  3. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    1982-03-01

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

  4. Structural analysis and manufacture for the vacuum vessel of experimental advanced superconducting tokamak (EAST) device

    International Nuclear Information System (INIS)

    Song Yuntao; Yao Damao; Wu Songata; Weng Peide

    2006-01-01

    The experimental advanced superconducting tokamak (EAST) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and is being constructed as the Chinese national nuclear fusion research project. The vacuum vessel, that is one of the key components, will have to withstand not only the electromagnetic force due to the plasma disruption and the Halo current, but also the pressure of boride water and the thermal stress due to the 250 deg. C baking out by the hot pressure nitrogen gas, or the 100 deg. C hot wall during plasma operation. This paper is a report of the mechanical analyses of the vacuum vessel. According to the allowable stress criteria of American Society of Mechanical Engineers, Boiler and Pressure Vessel Committee (ASME), the maximum integrated stress intensity on the vacuum vessel is 396 MPa, less than the allowable design stress intensity 3S m (441 MPa). At the same time, some key R and D issues are presented, which include supporting system, bellows and the assembly of the whole vacuum vessel

  5. Collaboration on Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Text Reactor. Final report

    International Nuclear Information System (INIS)

    Intrator, T.

    2000-01-01

    This proposal was peer reviewed and funded as a Collaboration on ''Low Phase Speed Radio Frequency Current Drive Experiments at the Tokamak Fusion Test Reactor''. The original plans we had were to carry out the collaboration proposal by including a post doctoral scientist stationed at PPPL. In response to a 60+% funding cut, all expenses were radically pruned. The post doctoral position was eliminated, and the Principal Investigator (T. Intrator) carried out the brunt of the collaboration. Visits to TFTR enabled T. Intrator to set up access to the TFTR computing network, database, and get familiar with the new antennas that were being installed in TFTR during an up to air. One unfortunate result of the budget squeeze that TFTR felt for its last year of operation was that the experiments that we specifically got funded to perform were not granted run time on TFTR., On the other hand we carried out some modeling of the electric field structure around the four strap direct launch Ion Bernstein Wave (IBW) antenna that was operated on TFTR. This turned out to be a useful exercise and shed some light on the operational characteristics of the IBW antenna and its coupling to the plasma. Because of this turn of events, the project was renamed ''Modeling of Ion Bernstein Wave Antenna Array and Coupling to Plasma on Tokamak Fusion Test Reactor''

  6. HTMR: an experimental tokamak reactor with hybrid copper/superconductor toroidal field magnet

    International Nuclear Information System (INIS)

    Avanzini, P.G.; Raia, G.; Rosatelli, F.; Zampaglione, V.

    1985-01-01

    The feasibility of a hybrid configuration superconducting coils/copper coils for a next generation tokamak TF magnet has been investigated. On the basis of this hybrid solution, the conceptual design has been developed for a medium-high toroidal field tokamak reactor (HTMR). The results of this study show the possibility of designing a tokamak reactor with reduced size in comparison with other INTOR like devices, still gaining some margins in front of the uncertainties in the scaling laws for plasma physics parameters and retaining the presence of a blanket with a tritium breeding ratio of about 1

  7. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    Willem Janssens, Ir.; Crutzen, Y.; Farfaletti-Casali, F.; Matera, R.

    1991-01-01

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  9. Experimental observations of MHD instabilities in the high-beta tokamak Torus-II

    International Nuclear Information System (INIS)

    Machida, M.

    1982-01-01

    The CO 2 laser scattering and interferometry diagnostics have been used to study the MHD instabilities in the high-beta tokamak Torus-II. Detailed measurements of the density and density fluctuation profiles have been performed. In order to measure density fluctuations with wavelengths longer than 2 cm, an interferometric like, phase matching technique has been developed. The toroidal and poloidal mode numbers have been measured using a double-beam, two-position technique. Working at high-beta values, average β greater than or equal to 10%, we have found parameters where the growing instabilities are created or suppressed. The plasma lifetime for both cases is seen to be about the same and the loss of the plasma appears to be caused by the decay in the external fields. The growing instability parameters are within the MHD regime, and it only grows at the outer edge of the plasma. This is in agreement with the theoretical Ballooning mode instability. The frequency and mode number measurements also agree with the Kinetic theory description of Ballooning modes. The comparison with possible other modes, such as Tearing and Drift instabilities, is performed and the Ballooning growth rate is shown to be the best fit to the experimental values

  10. Cross-section sensitivity analyses for a Tokamak Experimental Power Reactor

    International Nuclear Information System (INIS)

    Simmons, E.L.; Gerstl, S.A.W.; Dudziak, D.J.

    1977-09-01

    The objectives of this report were (1) to determine the sensitivity of neutronic responses in the preliminary design of the Tokamak Experimental Power Reactor by Argonne National Laboratory, and (2) to develop the use of a neutron-gamma coupled cross-section set in the calculation of cross-section sensitivity analysis. Response functions such as neutron plus gamma kerma, Mylar dose, copper transmutation, copper dpa, and activation of the toroidal field coil dewar were investigated. Calculations revealed that the responses were most sensitive to the high-energy group cross sections of iron in the innermost regions containing stainless steel. For example, both the neutron heating of the toroidal field coil and the activation of the toroidal field coil dewar show an integral sensitivity of about -5 with respect to the iron total cross sections. Major contributors are the scattering cross sections of iron, with -2.7 and -4.4 for neutron heating and activation, respectively. The effects of changes in gamma cross sections were generally an order of 10 lower

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  12. Experimental characteristics of ion Bernstein wave heating on JIPP T-IIU tokamak

    International Nuclear Information System (INIS)

    Ogawa, Y.; Kawahata, K.; Ando, R.

    1986-03-01

    The directly launched Ion Bernstein Wave (IBW) heating experiments have been carried out on JIPP T-IIU tokamak for two experimental conditions; (a) the ''3rd-branch'' of the IBW between 3rd- and 4th-cyclotron harmonics of the deuterium, and (b) the ''2nd-branch'' of the IBW between 2nd- and 3rd-cyclotron harmonics. In the case (a), the direct hydrogen heating at ω = 1.5 Ω H has been found in previous experiments. Here we present additional data to support this subharmonics heating, i.e., the spectroscopic measurement of Fe XVIII lines and mass separated analysis of charge-exchange neutrals. While, in the case (b), the remarkable increase of the electron temperature has been observed, especially at the central region of the plasma, and it has been estimated from the global energy balance that almost all of IBW power is delivered to the electron. To investigate this difference of the heating mode, the power absorption has been calculated with the ray tracing code, taking into account of the effect of the plasma/antenna coupling. It is concluded from the consideration of the electron Landau damping that the transition from the ion heating mode to the electron one would be explained by the difference of the electron temperature at the ohmic phase; i.e., T e (0) = 0.7 keV for the case (a) and T e (0) = 1.3 keV for the case (b). (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  14. MHD activity in the ISX-B tokamak: experimental results and theoretical interpretation

    Energy Technology Data Exchange (ETDEWEB)

    Carreras, B.A.; Dunlap, J.L.; Bell, J.D.; Charlton, L.A.; Cooper, W.A.; Dory, R.A.; Hender, T.C.; Hicks, H.R.; Holmes, J.A.; Lynch, V.E.

    1982-01-01

    The observed spectrum of MHD fluctuations in the ISX-B tokamak is clearly dominated by the n=1 mode when the q=1 surface is in the plasma. This fact agrees well with theoretical predictions based on 3-D resistive MHD calculations. They show that the (m=1; n=1) mode is then the dominant instability. It drives other n=1 modes through toroidal coupling and n>1 modes through nonlinear couplings. These theoretically predicted mode structures have been compared in detail with the experimentally measured wave forms (using arrays of soft x-ray detectors). The agreement is excellent. More detailed comparisons between theory and experiment have required careful reconstructions of the ISX-B equilibria. The equilibria so constructed have permitted a precise evaluation of the ideal MHD stability properties of ISX-B. The present results indicate that the high ..beta.. ISX-B equilibria are marginally stable to finite eta ideal MHD modes. The resistive MHD calculations also show that at finite ..beta.. there are unstable resistive pressure driven modes.

  15. New steady-state quiescent high-confinement plasma in an experimental advanced superconducting tokamak.

    Science.gov (United States)

    Hu, J S; Sun, Z; Guo, H Y; Li, J G; Wan, B N; Wang, H Q; Ding, S Y; Xu, G S; Liang, Y F; Mansfield, D K; Maingi, R; Zou, X L; Wang, L; Ren, J; Zuo, G Z; Zhang, L; Duan, Y M; Shi, T H; Hu, L Q

    2015-02-06

    A critical challenge facing the basic long-pulse high-confinement operation scenario (H mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as the edge localized mode (ELM), which leads to cyclical high peak heat and particle fluxes at the plasma facing components. A breakthrough is made in the Experimental Advanced Superconducting Tokamak in achieving a new steady-state H mode without the presence of ELMs for a duration exceeding hundreds of energy confinement times, by using a novel technique of continuous real-time injection of a lithium (Li) aerosol into the edge plasma. The steady-state ELM-free H mode is accompanied by a strong edge coherent MHD mode (ECM) at a frequency of 35-40 kHz with a poloidal wavelength of 10.2 cm in the ion diamagnetic drift direction, providing continuous heat and particle exhaust, thus preventing the transient heat deposition on plasma facing components and impurity accumulation in the confined plasma. It is truly remarkable that Li injection appears to promote the growth of the ECM, owing to the increase in Li concentration and hence collisionality at the edge, as predicted by GYRO simulations. This new steady-state ELM-free H-mode regime, enabled by real-time Li injection, may open a new avenue for next-step fusion development.

  16. Investigation of particle transport through the measurement of the electron source in the Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Klepper, C.C.

    1985-01-01

    The spatial distribution of the electron source was measured spectroscopically in the Texas Experimental Tokamak. The method used involves the measurement of the emissivity of the Balmer α and β lines of neutral hydrogen. Modeling of the corresponding atomic transitions provides a relation between the emissivities and the electron source from the ionization of neutrals. Toroidal distributions were obtained by means of a set of relatively calibrated photodiode amplifier-filter packages referred to as plasma light monitors. Such monitors were distributed toroidally, and attached primarily to radial ports. Specially constructed, absolutely calibrated monitors provided absolute calibration. A scanning, rotating mirror system provided in-out brightness profiles. A TV camera system, viewing the limiter through a tangential port, provided a qualitative description of the poloidal asymmetry. Such description was necessary for the inversion of the rotating mirror data. Using electron density profiles obtained by means of far-infrared interferometry, and integrating the electron sources, the global particle confinement time (tau/sub p/) was computed. Parameter scans were performed in ohmically heated plasmas, varying the toroidal field, the plasma current, the electron density, and the plasma position with respect to the center of the poloidal ring limiter. It was found that tau/sub p/ peaks for a critical density that is independent of the other parameters

  17. Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

    International Nuclear Information System (INIS)

    1976-01-01

    This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn 3 conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended

  18. 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

  19. Development and experimental evaluation of theoretical models for ion cyclotron resonance frequency heating of tokamak plasmas

    International Nuclear Information System (INIS)

    Mantsinen, M.

    1999-01-01

    Heating with electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is a well-established method for auxiliary heating of present-day tokamak plasmas and is envisaged as one of the main heating techniques for the International Thermonuclear Experimental Reactor (ITER) and future reactor plasmas. In order to predict the performance of ICRF heating in future machines, it is important to benchmark present theoretical modelling with experimental results on present tokamaks. This thesis reports on development and experimental evaluation of theoretical models for ICRF heating at the Joint European Torus (JET). Several ICRF physics effects and scenarios have been studied. Direct importance to the ITER is the theoretical analysis of ICRF heating experiments with deuterium-tritium (D-T) plasmas. These experiments clearly demonstrate the potential of ICRF heating for auxiliary heating of reactor plasmas. In particular, scenarios with potential for good bulk ion heating and enhanced D-T fusion reactivity have been identified. Good bulk ion heating is essential for reactor plasmas in order to obtain a high ion temperature and a high fusion reactivity. In JET good bulk ion heating with ICRF waves has been achieved in high-performance discharges by adding ICRF heating to neutral beam injection. In these experiments, as in other JET discharges where damping at higher harmonics of the ion cyclotron frequency takes place, so-called finite Larmor radius (FLR) effects play an important role. Due to FLR effects, the resonating ion velocity distribution function can have a strong influence on the power deposition. Evidence for this effect has been obtained from the third harmonic deuterium heating experiments. Because of FLR effects, the wave-particle interaction can also become weak at certain ion energies, which prevents resonating ions from reaching higher energies. When interacting with the wave, an ion receives not only a change in energy but also a change in

  20. Development and experimental evaluation of theoretical models for ion cyclotron resonance frequency heating of tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Mantsinen, M. [Helsinki Univ. of Technology, Espoo (Finland). Dept. of Technical Physics

    1999-06-01

    Heating with electromagnetic waves in the ion cyclotron range of frequencies (ICRF) is a well-established method for auxiliary heating of present-day tokamak plasmas and is envisaged as one of the main heating techniques for the International Thermonuclear Experimental Reactor (ITER) and future reactor plasmas. In order to predict the performance of ICRF heating in future machines, it is important to benchmark present theoretical modelling with experimental results on present tokamaks. This thesis reports on development and experimental evaluation of theoretical models for ICRF heating at the Joint European Torus (JET). Several ICRF physics effects and scenarios have been studied. Direct importance to the ITER is the theoretical analysis of ICRF heating experiments with deuterium-tritium (D-T) plasmas. These experiments clearly demonstrate the potential of ICRF heating for auxiliary heating of reactor plasmas. In particular, scenarios with potential for good bulk ion heating and enhanced D-T fusion reactivity have been identified. Good bulk ion heating is essential for reactor plasmas in order to obtain a high ion temperature and a high fusion reactivity. In JET good bulk ion heating with ICRF waves has been achieved in high-performance discharges by adding ICRF heating to neutral beam injection. In these experiments, as in other JET discharges where damping at higher harmonics of the ion cyclotron frequency takes place, so-called finite Larmor radius (FLR) effects play an important role. Due to FLR effects, the resonating ion velocity distribution function can have a strong influence on the power deposition. Evidence for this effect has been obtained from the third harmonic deuterium heating experiments. Because of FLR effects, the wave-particle interaction can also become weak at certain ion energies, which prevents resonating ions from reaching higher energies. When interacting with the wave, an ion receives not only a change in energy but also a change in

  1. Experimental investigation of density behaviors in front of the lower hybrid launcher in experimental advanced superconducting tokamak

    International Nuclear Information System (INIS)

    Zhang, L.; Ding, B. J.; Li, M. H.; Kong, E. H.; Wei, W.; Liu, F. K.; Shan, J. F.; Wu, Z. G.; Zhu, L.; Ma, W. D.; Tong, Y. Y.; Li, Y. C.; Wang, M.; Zhao, L. M.; Hu, H. C.; Liu, L.

    2013-01-01

    A triple Langmuir probe is mounted on the top of the Lower Hybrid (LH) antenna to measure the electron density near the LH grills in Experimental Advanced Superconducting Tokamak. In this work, the LH power density ranges from 2.3 MWm −2 to 10.3 MWm −2 and the rate of puffing gas varies from 1.7 × 10 20 el/s to 14 × 10 20 el/s. The relation between the edge density (from 0.3 × n e-cutoff to 20 × n e-cutoff , where n e-cutoff is the cutoff density, n e-cutoff = 0.74 × 10 17 m −3 for 2.45 GHz lower hybrid current drive) near the LH grill and the LH power reflection coefficients is investigated. The factors, including the gap between the LH grills and the last closed magnetic flux surface, line-averaged density, LH power, edge safety factor, and gas puffing, are analyzed. The experiments show that injection of LH power is beneficial for increasing edge density. Gas puffing is beneficial for increasing grill density but excess gas puffing is unfavorable for coupling and current drive

  2. Experimental study of the interaction between RF antennas and the edge plasma of a tokamak

    International Nuclear Information System (INIS)

    Kubic, Martin

    2013-01-01

    Antennas operating in the ion cyclotron range of frequency (ICRF) provide a useful tool for plasma heating in many tokamaks and are foreseen to play an important role in ITER. However, in addition to the desired heating in the core plasma, spurious interactions with the plasma edge and material boundary are known to occur. Many of these deleterious effects are caused by the formation of radio-frequency (RF) sheaths. The aim of this thesis is to study, mainly experimentally, scrape-off layer (SOL) modifications caused by RF sheaths effects by means of Langmuir probes that are magnetically connected to a powered ICRH antenna. Effects of the two types of Faraday screens' operation on RF-induced SOL modifications are studied for different plasma and antenna configurations - scans of strap power ratio imbalance, injected power and SOL density. In addition to experimental work, the influence of RF sheaths on retarding field analyzer (RFA) measurements of sheath potential is investigated with one-dimensional particle-in-cell code. One-dimensional particle-in-cell simulations show that the RFA is able to measure reliably the sheath potential only for ion plasma frequencies ω π similar to RF cyclotron frequency ω rf , while for the real SOL conditions (ω π ≥ ω rf ), when the RFA is magnetically connected to RF region, it is strongly underestimated. An alternative method to investigate RF sheaths effects is proposed by using broadening of the ion distribution function as an evidence of the RF electric fields in the sheath. RFA measurements in Tore Supra indicate that RF potentials do indeed propagate from the antenna 12 m along magnetic field lines. (author) [fr

  3. Recent experimental studies of edge and internal transport barriers in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Gohil, P; Baylor, L R; Burrell, K H; Casper, T A; Doyle, E J; Greenfield, C M; Jernigan, T C; Kinsey, J E; Lasnier, C J; Moyer, R A; Murakami, M; Rhodes, T L; Rudakov, D L; Staebler, G M; Wang, G; Watkins, J G; West, W P; Zeng, L

    2003-01-01

    Results from recent experiments on the DIII-D tokamak have revealed many important details on transport barriers at the plasma edge and in the plasma core. These experiments include: (a) the formation of the H-mode edge barrier directly by pellet injection; (b) the formation of a quiescent H-mode edge barrier (QH-mode) which is free from edge localized modes, but which still exhibits good density and radiative power control; (c) the formation of multiple transport barriers, such as the quiescent double barrier (QDB) which combines an internal transport barrier with the quiescent H-mode edge barrier. Results from the pellet-induced H-mode experiments indicate that: (a) the edge temperature (electron or ion) does not need to attain a critical value for the formation of the H-mode barrier, (b) pellet injection leads to an increased gradient in the radial electric field, E r , at the plasma edge; (c) the experimentally determined edge parameters at barrier transition are well below the predictions of several theories on the formation of the H-mode barrier, (d) pellet injection can lower the threshold power required to form the H-mode barrier. The quiescent H-mode barrier exhibits good density control as the result of continuous magnetohydrodynamic activity at the plasma edge called the edge harmonic oscillation (EHO). The EHO enhances the edge particle transport whilst maintaining a good energy transport barrier. The ability to produce multiple barriers in the QDB regime has led to long duration, high-performance plasmas with β N H 89 values of 7 for up to 10 times the confinement time. Density profile control in the plasma core of QDB plasmas has been demonstrated using on-axis electron cyclotron heating

  4. Experimental and modeling researches of dust particles in the HL-2A tokamak

    Science.gov (United States)

    Huang, Zhi-Hui; Yan, Long-Wen; Tomita, Yukihiro; Feng, Zhen; Cheng, Jun; Hong, Wen-Yu; Pan, Yu-Dong; Yang, Qing-Wei; Duan, Xu-Ru

    2015-02-01

    The investigation of dust particle characteristics in fusion devices has become more and more imperative. In the HL-2A tokamak, the morphologies and compositions of dust particles are analyzed by using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDX) with mapping. The results indicate that the sizes of dust particles are in a range from 1 μm to 1 mm. Surprisingly, stainless steel spheres with a diameter of 2.5 μm-30 μm are obtained. The production mechanisms of dust particles include flaking, disintegration, agglomeration, and arcing. In addition, dynamic characteristics of the flaking dust particles are observed by a CMOS fast framing camera and simulated by a computer program. Both of the results display that the ion friction force is dominant in the toroidal direction, while the centrifugal force is crucial in the radial direction. Therefore, the visible dust particles are accelerated toriodally by the ion friction force and migrated radially by the centrifugal force. The averaged velocity of the grain is on the order of ˜ 100 m/s. These results provide an additional supplement for one of critical plasma-wall interaction (PWI) issues in the framework of the International Thermonuclear Experimental Reactor (ITER) programme. Project supported by the National Magnetic Confinement Fusion Science Program of China (Grant Nos. 2014GB107000 and 2013GB112008), the National Natural Science Foundation of China (Grant Nos. 11320101005, 11175060, 11375054, and 11075046), and the China-Korean Joint Foundation (Grant No. 2012DFG02230).

  5. Nuclear design of the blanket/shield system for a Tokamak Experimental Power Reactor

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1976-01-01

    The various options and trade-offs in the nuclear design of the blanket/shield for a Tokamak Experimental Power Reactor (TEPR) are investigated. The TEPR size and cost are particularly sensitive to the blanket/shield thickness, Δ/sub BS/, on the inner side of the torus. Radition damage to the components of the superconducting magnet and refrigeration power requirements set lower limits on Δ/sub BS/. These limits are developed in terms of TEPR design parameters such as the wall loading, duty cycle, and frequency of magnet anneals. The study of the nuclear performance of various material compositions shows that mixtures of tungsten, or tantalum, or stainless-steel alloys and boron carbide require the smallest Δ/sub BS/ for a given attenuation. This Δ/sub BS/ has to be doubled if the low induced activation materials graphite and aluminum are used. The space problems are greatly eased in the Argonne National Laboratory ANL-TEPR reference design by using two separate segments of the blanket/shield. The inner segment occupies the region of the high magnetic field, uses very efficient attenuators (tungsten- or tantalum- or stainless-steel-boron carbide mixtures), and is only 1 m thick. The outer blanket/shield is 131 cm and consists of an optimized composition of stainless steel and boron carbide. For the design parameters of 0.2 MW/m 2 neutron wall loading and 50 percent duty cycle, the reactor components can operate satisfactorily up to (a) 10 yr for the stainless-steel first wall, (b) 10 yr for the superconductor composite after which magnet warmup becomes necessary, and (c) 30 yr for the Mylar insulation. Nuclear heat generation rates in the blanket/shield and magnet are well within the practical limits for heat removal

  6. Magnetic field shielding system in a tokamak experimental power reactor (EPR): concept and calculations

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Marcus, F.B.; Dory, R.A.; Moore, J.R.

    1975-01-01

    A poloidal magnetic field shielding system is proposed for a tokamak EPR. This coil system minimizes the pulsed poloidal field that intersects the TF (toroidal field) coils and hence reduces the risk of superconductor quenching and structural failure of the coils. Based on an idealized shielding model, we have determined the configurations for the OH (ohmic heating), the S-VF (shield-vertical field), and the T-VF (trimming-vertical field) coils in a typical tokamak EPR. It is found that the pulsed poloidal field strength is greatly reduced in the TF coil region. The overall requirement in stored plasma and vertical field energy is also substantially reduced when compared with conventional EPR designs. Use of this field shielding system is expected to enhance reliability of the superconducting TF coils in a tokamak EPR

  7. Advanced Tokamak Stability Theory

    Science.gov (United States)

    Zheng, Linjin

    2015-03-01

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

  8. Experimental observations of surface electrostatic wave on KT-5B tokamak

    International Nuclear Information System (INIS)

    Zhu Shiyao; Han Shensheng

    1991-01-01

    Shear Alfven waves have been successfully excited in KT-5B small tokamak by means of the one turn longitudinal loop antenna located in the shadow area. The measured antenna loadings show their rich structure, and the loadings are also found to be sensitive to the plasma current. Preliminary evidence of surface electrostatic wave was observed

  9. Numerical and experimental analysis of eddy currents induced in tokamak machines

    International Nuclear Information System (INIS)

    Takahashi, T.; Takahashi, G.; Kazawa, Y.; Suzuki, Y.

    1977-01-01

    This paper deals with eddy current phenomena in Tokamak machines. A numerical method is presented which will permit eddy currents to be calculated. Examples of numerical results and a discussion of the JT-60 are shown. Calculations are checked by measurements in basic models

  10. New dual gas puff imaging system with up-down symmetry on experimental advanced superconducting tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Shao, L. M.; Zweben, S. J.

    2012-01-01

    advanced superconducting tokamak (EAST). The two views are up-down symmetric about the midplane and separated by a toroidal angle of 66.6 degrees. A linear manifold with 16 holes apart by 10 mm is used to form helium gas cloud at the 130x130 mm (radial versus poloidal) objective plane. A fast camera...

  11. Experimental study of parametric dependence of electron-scale turbulence in a spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Ren, Y.; Guttenfelder, W.; Kaye, S. M.; Mazzucato, E.; Bell, R. E.; Diallo, A.; LeBlanc, B. P. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Domier, C. W.; Lee, K. C. [University of California at Davis, Davis, California 95616 (United States); Smith, D. R. [University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Yuh, H. [Nova Photonics, Inc., Princeton, New Jersey 08540 (United States)

    2012-05-15

    Electron-scale turbulence is predicted to drive anomalous electron thermal transport. However, experimental study of its relation with transport is still in its early stage. On the National Spherical Tokamak Experiment (NSTX), electron-scale density fluctuations are studied with a novel tangential microwave scattering system with high radial resolution of {+-}2 cm. Here, we report a study of parametric dependence of electron-scale turbulence in NSTX H-mode plasmas. The dependence on density gradient is studied through the observation of a large density gradient variation in the core induced by an edge localized mode (ELM) event, where we found the first clear experimental evidence of density gradient stabilization of electron-gyro scale turbulence in a fusion plasma. This observation, coupled with linear gyro-kinetic calculations, leads to the identification of the observed instability as toroidal electron temperature gradient (ETG) modes. It is observed that longer wavelength ETG modes, k{sub Up-Tack }{rho}{sub s} Less-Than-Or-Equivalent-To 10 ({rho}{sub s} is the ion gyroradius at electron temperature and k{sub Up-Tack} is the wavenumber perpendicular to local equilibrium magnetic field), are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in electron thermal diffusivity. Comparisons with nonlinear ETG gyrokinetic simulations show ETG turbulence may be able to explain the experimental electron heat flux observed before the ELM event. The collisionality dependence of electron-scale turbulence is also studied by systematically varying plasma current and toroidal field, so that electron gyroradius ({rho}{sub e}), electron beta ({beta}{sub e}), and safety factor (q{sub 95}) are kept approximately constant. More than a factor of two change in electron collisionality, {nu}{sub e}{sup *}, was achieved, and we found that the spectral power of electron-scale turbulence appears to increase as {nu}{sub e}{sup *} is

  12. Measurement of the electron and ion temperatures by the x-ray imaging crystal spectrometer on joint Texas experimental tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yan, W.; Chen, Z. Y., E-mail: zychen@hust.edu.cn; Huang, D. W.; Tong, R. H.; Wang, S. Y.; Wei, Y. N.; Ma, T. K.; Zhuang, G. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan (China); Jin, W. [Center of Interface Dynamics for Sustainability, China Academy of Engineering Physics, Chengdu, Sichuan 610200 (China); Lee, S. G. [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Shi, Y. J. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of)

    2016-11-15

    An x-ray imaging crystal spectrometer has been developed on joint Texas experimental tokamak for the measurement of electron and ion temperatures from the K{sub α} spectra of helium-like argon and its satellite lines. A two-dimensional multi-wire proportional counter has been applied to detect the spectra. The electron and ion temperatures have been obtained from the Voigt fitting with the spectra of helium-like argon ions. The profiles of electron and ion temperatures show the dependence on electron density in ohmic plasmas.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  14. Overview of recent experimental results from the DIII-D advanced tokamak program

    International Nuclear Information System (INIS)

    Burrell, K.H.

    2003-01-01

    The D III-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: 1) We have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; 2) Using this rotational stabilization, we have achieved β N H 89 ≥ 10 for 4 τ E limited by the neoclassical tearing mode; 3) Using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased β T by 60%; 4) We have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; 5) We have made the first integrated AT demonstration discharges with current profile control using ECCD; 6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and 7) We have demonstrated stationary tokamak operation for 6.5 s (36 τ E ) at the same fusion gain parameter of β N H 89 /q 95 2 ≅ 0.4 as ITER but at much higher q 95 = 4.2. We have developed general improvements applicable to conventional and advanced tokamak operating modes: 1) We have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 τ E ) with constant density and constant radiated power; 2) We have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; 3) We have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much smaller than to the outer strike points. (author)

  15. Experimental investigation on electron cyclotron absorption at down-shifted frequency in the PLT tokamak

    International Nuclear Information System (INIS)

    Mazzucato, E.; Fidone, I.; Cavallo, A.; von Goeler, S.; Hsuan, H.

    1986-05-01

    The absorption of 60 GHz electron cyclotron waves, with the extraordinary mode and an oblique angle of propagation, has been investigated in the PLT tokamak in the regime of down-shifted frequencies. The production of energetic electrons, with energies of up to 300 to 400 keV, peaks at values of toroidal field (approx. =29 kG) for which the wave frequency is significantly smaller than the electron cyclotron frequency in the whole plasma region. The observations are consistent with the predictions of the relativistic theory of electron cyclotron damping at down-shifted frequency. Existing rf sources make this process a viable method for assisting the current ramp-up, and for heating the plasma of present large tokamaks

  16. Experimental study on practicability of self-created spherical tokamak in coilless STPC-EX machine

    International Nuclear Information System (INIS)

    Sinman, S.

    2002-01-01

    The aim of this study is to recognize the physical basis of the alternative self organization mechanism occurred STPC-EX machine. The conventional diagnostic tools are used in this study and for photographic recording, open shutter integrated post-fogging method is preferred. The annular coaxial two plasma current sheets one within other at the same direction are created and flowed on the surface of floating conductive central rod. Consequently, spherical tokamak configurated by new creation mechanism of Dual Axial Z-Pinch. (DAZP) yields fairly high beta of 0.4-0.6 at self created spherical tokamak plasma. Sustainment time of DAZP is 5.6-6.3 mili second. (author)

  17. Comparison Between 2D Turbulence Model ESEL and Experimental Data from AUG and Compass Tokamaks

    Czech Academy of Sciences Publication Activity Database

    Ondáč, Peter; Horáček, Jan; Seidl, Jakub; Vondráček, Petr; Müller, H.W.; Adámek, Jiří; Nielsen, A.H.

    2015-01-01

    Roč. 55, č. 2 (2015), s. 128-135 ISSN 1210-2709 R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : turbulence * tokamak * computer model * probe measurements Subject RIV: BL - Plasma and Gas Discharge Physics https://ojs.cvut.cz/ojs/index.php/ap/article/viewFile/2257/2816

  18. Tokamak control simulator

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  19. Experimental studies of thermal and non-thermal electron cyclotron phenomena in tokamaks

    International Nuclear Information System (INIS)

    McDermott, F.S.

    1984-12-01

    A direct measurement of wave absorption in the ISX-B tokamak at the second harmonic of the electron cyclotron frequency is reported. Measurements of the absorption of a wave polarized in the extraordinary mode and propagating perpendicular to the toroidal magnetic field are in agreement with the absorption predicted by the linearized Vlasov equation for a thermal plasma. Agreement is found both for an analytic approximation to the wave absorption and for a numerical simulation of ray propagation in toroidal geometry. Observations are also reported on a non-linear, three-wave interaction process occurring during high power electron cyclotron resonance heating in the Versator II tokamak. The measured spectra and the threshold power are consistent with a model in which the incident power in the extraordinary mode of polarization decays at the upper hybrid resonance layer into a lower hybrid wave and an electron Bernstein wave. Finally, measurements of non-thermal emission at the second harmonic of the electron cyclotron frequency and below the electron plasma frequency are reported from low density, non-Maxwellian plasma in the Versator II tokamak. The emission spectra are in agreement with a model in which waves are driven unstable at the anomalous Doppler resonance, while only weakly damped at the Cerenkov resonance

  20. Experimental study of the β-limit in ohmic H-mode in the TUMAN-3M tokamak

    International Nuclear Information System (INIS)

    Lebedev, S.V.; Andreiko, M.V.; Askinazi, L.G.; Golant, V.E.; Kornev, V.A.; Krikunov, S.V.; Levin, L.S.; Rozhdestvensky, V.V.; Tukachinsky, A.S.; Yaroshevich, S.P.

    1998-01-01

    Because of its high confinement properties, the H-mode provides good opportunities to achieve high beta values in a tokamak. In this paper the results of an experimental study of β T and β N limits in the H-mode, obtained in a circular cross section tokamak without auxiliary heating are presented. The experiments were performed in the TUMAN-3M tokamak. The device has the following parameters: R 0 =0.53m, a s =0.22m (limiter configuration), B T ≤1.2T, I p ≤175kA, n-bar e ≤6.2x10 19 m -3 . The stored energy was measured using diamagnetic loops and compared with W calculated from kinetic data obtained by Thomson scattering and microwave interferometry. Measurements of the stored energy and of the β were performed in the ohmic H-mode before and after boronization and in the scenario with fast current ramp-down in ohmic H-mode. A maximum value of β T of 2.0% and β N of 2.0 were achieved. The β N limit achieved reveals itself as a 'soft' (non-disruptive) limit. The stored energy slowly decays after the current ramp-down. No correlation was found between beta restriction and MHD phenomena. Internal transport barrier (ITB) formation was observed in ohmic H-mode. An enhancement factor of 2.0 over ITER93H(ELM-free) was found in the ohmic H-mode with ITB. (author)

  1. Simulations of the L-H transition on experimental advanced superconducting Tokamak

    International Nuclear Information System (INIS)

    Weiland, Jan

    2014-01-01

    We have simulated the L-H transition on the EAST tokamak [Baonian Wan, EAST and HT-7 Teams, and International Collaborators, “Recent experiments in the EAST and HT-7 superconducting tokamaks,” Nucl. Fusion 49, 104011 (2009)] using a predictive transport code where ion and electron temperatures, electron density, and poloidal and toroidal momenta are simulated self consistently. This is, as far as we know, the first theory based simulation of an L-H transition including the whole radius and not making any assumptions about where the barrier should be formed. Another remarkable feature is that we get H-mode gradients in agreement with the α – α d diagram of Rogers et al. [Phys. Rev. Lett. 81, 4396 (1998)]. Then, the feedback loop emerging from the simulations means that the L-H power threshold increases with the temperature at the separatrix. This is a main feature of the C-mod experiments [Hubbard et al., Phys. Plasmas 14, 056109 (2007)]. This is also why the power threshold depends on the direction of the grad B drift in the scrape off layer and also why the power threshold increases with the magnetic field. A further significant general H-mode feature is that the density is much flatter in H-mode than in L-mode

  2. Experimental determination of the dimensionless scaling parameter of energy transport in tokamaks

    International Nuclear Information System (INIS)

    Luce, T.C.; Petty, C.C.

    1995-07-01

    Controlled fusion experiments have focused on the variation of the plasma characteristics as the engineering or control parameters are systematically changed. This has led to the development of extrapolation formulae for prediction of future device performance using these same variables as a basis. Recently, it was noticed that present-day tokamaks can operate with all of the dimensionless variables which appear in the Vlasov-Maxwell system of equations at values projected for a fusion powerplant with the exception of the parameter ρ * , the gyroradius normalized to the machine size. The scaling with this parameter is related to the benefit of increasing the size of the machine either directly or effectively by increasing the magnetic field. It is exactly this scaling which is subject to systematic error in the inter-machine databases and the cost driver for any future machine. If this scaling can be fixed by a series of single machine experiments, much as the current and power scalings have been, the confidence in the prediction of future device performance would be greatly enhanced. While carrying out experiments of this type, it was also found that the ρ * scaling can illuminate the underlying physics of energy transport. Conclusions drawn from experiments on the DIII-D tokamak in these two areas are the subject of this paper

  3. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1977-01-01

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

  4. Hydrocarbon deposition in gaps of tungsten and graphite tiles in Experimental Advanced Superconducting Tokamak edge plasma parameters

    International Nuclear Information System (INIS)

    Xu Qian; Yang Zhongshi; Luo Guangnan

    2015-01-01

    The three-dimensional (3D) Monte Carlo code PIC-EDDY has been utilized to investigate the mechanism of hydrocarbon deposition in gaps of tungsten tiles in the Experimental Advanced Superconducting Tokamak (EAST), where the sheath potential is calculated by the 2D in space and 3D in velocity particle-in-cell method. The calculated results for graphite tiles using the same method are also presented for comparison. Calculation results show that the amount of carbon deposited in the gaps of carbon tiles is three times larger than that in the gaps of tungsten tiles when the carbon particles from re-erosion on the top surface of monoblocks are taken into account. However, the deposition amount is found to be larger in the gaps of tungsten tiles at the same CH 4 flux. When chemical sputtering becomes significant as carbon coverage on tungsten increases with exposure time, the deposition inside the gaps of tungsten tiles would be considerable. (author)

  5. Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Minato, Akio; Tone, Tatsuzo

    1985-12-01

    A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

  6. Simulation of Heating with the Waves of Ion Cyclotron Range of Frequencies in Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Yang Cheng; Zhu Sizheng; Zhang Xinjun

    2010-01-01

    Simulation on the heating scenarios in experimental advanced superconducting tokamak (EAST) was performed by using a full wave code TORIC. The locations of resonance layers for these heating schemes are predicted and the simulations for different schemes in ICRF experiments in EAST, for example, ion heating (both fundamental and harmonic frequency) or electron heating (by direct fast waves or by mode conversion waves), on-axis or off-axis heating, and high-field-side (HFS) launching or low-field-side (LFS) launching, etc, were conducted. For the on-axis minority ion heating of 3 He in D( 3 He) plasma, the impacts of both density and temperature on heating were discussed in the EAST parameter ranges.

  7. 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

  8. A current-pulsed power supply with rapid rising and falling edges for magnetic perturbation coils on the J-TEXT tokamak

    International Nuclear Information System (INIS)

    Yan, M.X.; Rao, B.; Ding, Y.H.; Hu, Q.M.; Hu, F.R.; Li, D.; Li, M.; Ji, X.K.; Xu, G.; Zheng, W.; Jiang, Z.H.

    2017-01-01

    Highlights: • The power supply is required to have rapid rising and falling edges. • A modified topology based on the buck chopper of current-pulsed power supply is presented and analyzed. • An entity meeting the electrical requirements has been constructed. • The spike voltage of IGBT is qualitatively analyzed. - Abstract: This study presents the design and principle of a current-pulsed power supply (CPPS) for the tearing mode (TM) feedback control of the J-TEXT tokamak. CPPS is a new method of stabilizing large magnetic islands and accelerating mode rotation through the use of modulated magnetic perturbation. In this application, continuous magnetic perturbation pulse trains with frequency of 1 kHz to kHz, amplitude of 0.25 G, and duty ratio of 20%–50% are required generating via in-vessel magnetic coils. A modified topology based on buck chopper is raised to satisfy the demands of inductive load. This modified topology is characterized by high frequency, rapid rising and falling edges, and large amplitude of current pulses. Appropriate RCD snubber circuit is applied to protect the Insulated Gate Bipolar Transistor (IGBT) switch device. Equipment with peak current that reaches 1 kA, frequency that ranges from 1 kHz to 3 kHz, and rising and falling time within 100 μs was constructed and applied to physical experiment.

  9. A current-pulsed power supply with rapid rising and falling edges for magnetic perturbation coils on the J-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yan, M.X. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Rao, B., E-mail: borao@hust.edu.cn [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Ding, Y.H.; Hu, Q.M.; Hu, F.R.; Li, D.; Li, M.; Ji, X.K.; Xu, G.; Zheng, W.; Jiang, Z.H. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China); College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2017-02-15

    Highlights: • The power supply is required to have rapid rising and falling edges. • A modified topology based on the buck chopper of current-pulsed power supply is presented and analyzed. • An entity meeting the electrical requirements has been constructed. • The spike voltage of IGBT is qualitatively analyzed. - Abstract: This study presents the design and principle of a current-pulsed power supply (CPPS) for the tearing mode (TM) feedback control of the J-TEXT tokamak. CPPS is a new method of stabilizing large magnetic islands and accelerating mode rotation through the use of modulated magnetic perturbation. In this application, continuous magnetic perturbation pulse trains with frequency of 1 kHz to kHz, amplitude of 0.25 G, and duty ratio of 20%–50% are required generating via in-vessel magnetic coils. A modified topology based on buck chopper is raised to satisfy the demands of inductive load. This modified topology is characterized by high frequency, rapid rising and falling edges, and large amplitude of current pulses. Appropriate RCD snubber circuit is applied to protect the Insulated Gate Bipolar Transistor (IGBT) switch device. Equipment with peak current that reaches 1 kA, frequency that ranges from 1 kHz to 3 kHz, and rising and falling time within 100 μs was constructed and applied to physical experiment.

  10. Experimental results on pellet injection and MHD from the RTP tokamak

    International Nuclear Information System (INIS)

    Oomens, A.A.M.; Kloe, J. de; Salzedas, F.J.B.

    2001-01-01

    The ablation of hydrogen pellets has been studied in the Rijnhuizen Tokamak Project RTP with a diagnostic with high spatial and temporal resolution. It has been observed that (part of the) ablation cloud drifts away from the pellet in opposite direction. These drifts occur in semi-periodical bursts. A summary of a detailed analysis of this drift of the cloud and its implications for the fueling profile is presented. Stabilization of m/n=2/1 tearing modes preceding density limit disruptions, has been studied with modulated and continuous ECRH. The results indicate that EC heating of the islands under these conditions is very inefficient. The time dependence of the growth rate of the precursor mode is first algebraic, but becomes exponential in a later phase. (author)

  11. Experimental measurements of lower-hybrid wave propagation in the Versator II tokamak using microwave scattering

    International Nuclear Information System (INIS)

    Rohatgi, R.; Chen, K.; Bekefi, G.; Bonoli, P.; Luckhardt, S.C.; Mayberry, M.; Porkolab, M.; Villasenor, J.

    1991-01-01

    A series of 139 GHz microwave scattering experiments has been performed on the Versator II tokamak (B. Richards, Ph.D. thesis, Massachusetts Institute of Technology, 1981) to study the propagation of externally launched 0.8 GHz lower-hybrid waves. During lower-hybrid current drive, the launched waves are found to follow a highly directional resonance cone in the outer portion of the plasma. Wave power is also detected near the center of the plasma, and evidence of wave absorption is seen. Scattering of lower-hybrid waves in k space by density fluctuations appears to be a weak effect, although measurable frequency broadening by density fluctuations is found, Δω/ω=3x10 -4 . In the detectable range (2.5 parallel parallel spectra inferred from the scattering measurements are quite similar above and below the current drive density limit. Numerical modeling of these experiments using ray tracing is also presented

  12. Experimental survey of the L-H transition conditions in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Carlstrom, T.N.; Gohil, P.; Watkins, J.C.

    1994-01-01

    We present the global analysis of a recent survey of the H-mode power threshold in DIII-D using D o → D + NBI after boronization of the vacuum vessel. Single parameter scans of B T , I p , density, and plasma shape have been carried out on the DIII-D tokamak for neutral beam heated single-null and double-null diverted plasmas. In single-null discharges, the power threshold is found to increase approximately linearly with B T and n e but remains independent of I p . In double-null discharges, the power threshold is found to be approximately independent of both B T and n e . Various shape parameters such as plasma-wall gaps had only a weak effect on the power threshold. Imbalancing the double null configuration resulted in a large increase in the threshold power

  13. An experimental study of turbulence by phase-contrast imaging in the DIII-D tokamak

    Science.gov (United States)

    Coda, Stefano

    1997-10-01

    A CO2-laser imaging system employing the Zernike phase-contrast technique was designed, built, installed, and operated on the DIII-D tokamak. This system measures the line integrals of plasma density fluctuations along 16 vertical chords at the outer edge of the tokamak (0.85 Mechanical vibrations are damped by a novel dual-axis focal-spot feedback stabilization system. The theoretical treatment of scattering and imaging techniques was extended to finite-frequency fluctuations in the Rytov approximation. An extensive comparative analysis of the properties of phase-contrast imaging (PCI) and of other imaging and scintillation techniques was also carried out. Studies of edge turbulence were performed. The radial- wave-number spectrum peaks at finite wave numbers, both positive and negative. This first observation of radial modes is in agreement with recent predictions from theoretical and numerical work. The dependence of the correlation length and peak wave number on plasma parameters and on the frequency was studied in detail. Frequency spectra typically obey an inverse square law, consistent with a Lorentzian distribution. At the transition from L to H mode the amplitude and correlation length of the turbulence decrease, while the decorrelation time remains approximately constant. The Biglari-Diamond-Terry shear-decorrelation criterion was verified quantitatively; theoretical scaling laws for the correlation parameters were also tested. The turbulence amplitude follows a mixing-length scaling in L mode only: the lower level seen in H mode may indicate a weaker turbulence regime. The fluctuation content of Edge Localized Modes (ELMs) was thoroughly characterized, and systematic differences between type-I and type-III ELMs were discovered. Future applications of PCI, including crossed-beam localization and heterodyne radio-frequency-wave detection, are also discussed. (Copies available exclusively from MIT Libraries, Rm. 14-0551, Cambridge, MA 02139-4307. Ph. 617

  14. Overview of recent experimental results from the DIII-D advanced tokamak program

    International Nuclear Information System (INIS)

    Allen, S.L.

    2001-01-01

    The goals of DIII-D Advanced Tokamak (AT) experiments are to investigate and optimize the upper limits of energy confinement and MHD stability in a tokamak plasma, and to simultaneously maximize the fraction of non-inductive current drive. Significant overall progress has been made in the past 2 years, as the performance figure of merit β N H 89P of 9 has been achieved in ELMing H-mode for over 16 τ E without sawteeth. We also operated at β N ∼7 for over 35 τ E or 3 τ R , with the duration limited by hardware. Real-time feedback control of β (at 95% of the stability boundary), optimizing the plasma shape (e.g., δ, divertor strike- and X-point, double/single null balance), and particle control (n e /n GW ∼0.3, Z eff N H 89P of 7. The QDB regime has been obtained to date only with counter neutral beam injection. Further modification and control of internal transport barriers (ITBs) has also been demonstrated with impurity injection (broader barrier), pellets, and ECH (strong electron barrier). The new Divertor-2000, a key ingredient in all these discharges, provides effective density, impurity and heat flux control in the high-triangularity plasma shapes. Discharges at n e /n GW ∼1.4 have been obtained with gas puffing by maintaining the edge pedestal pressure; this operation is easier with Divertor-2000. We are developing several other tools required for AT operation, including real-time feedback control of resistive wall modes (RWMs) with external coils, and control of neoclassical tearing modes (NTMs) with electron cyclotron current drive (ECCD). (author)

  15. An experimental study of plasma fluctuations in the TCV and TEXTOR Tokamaks

    International Nuclear Information System (INIS)

    Mejeire de, C. A.

    2013-01-01

    The main body of this thesis reports on the commissioning and first measurements with a novel tangential phase-contrast imaging (TPCI) diagnostic, which had previously been installed in the TCV tokamak. The instrument measures fluctuations in line-integrated electron density along 9 parallel chords within a 6 cm diameter CO 2 laser beam. TPCI measurements reveal the first evidence in TCV of the geodesic acoustic mode (GAM), which is an oscillating zonal flow. Frequency, radial wavelength, radial extent and propagation are all in qualitative agreement with a gyro-kinetic simulation and recent theoretical work. The mode is found to have a modest, but measurable magnetic component, whose spatial structure is characterised for the first time in a toroidal plasma. For some experiments, clear evidence is found of the theoretically expected m/n = 2/0 mode structure, although in others the structure appears to be more complex. Electron energy confinement in X 2 heated TCV L-mode plasmas had previously been observed to increase on changing the triangularity (δ) of the poloidal plasma cross-section from δ = +0.4 to δ = −0.4. Measurements with the TPCI diagnostic reveal that this change coincides with a clear decrease in both the absolute level and the decorrelation time of broadband electron density fluctuations. This is in agreement with the conjecture that the increased confinement time is caused by a change in the turbulent state. The second part of the thesis reports on a fluctuation study in the TEXTOR tokamak. At sufficiently weak toroidal magnetic field, NBI heated, limited TEXTOR plasmas exhibit bursts of beam-ion driven ‘fishbone’ and Alfvén modes, which are characterised using the multi-antenna reflectometer and Mirnov coils. In H-mode the fishbone triggers ELMs and in L-mode it triggers previously unobserved bursts of particle recycling, resembling the ELMs. The reflectometer phase shows statistically significant bispectral coherence between the fishbone

  16. 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)

  17. Experimental study on plasma parameters in the mushroom limiter shadow in the T-10 tokamak

    International Nuclear Information System (INIS)

    Alferov, A.A.; Vershkov, V.A.; Grashin, S.A.; Chankin, A.V.

    1988-01-01

    Plasma parameters in the shadow of mashroom limiter installed in the lower tokamak outlet are studied. Investigation into asymmetry of plasma fluxes to the ion and electrone limiter sides leads to a consumption concerning two meachanisms of its occurrance-toroidal plasma rotation and prevailing plasma departure to the wall through the external torus encirclement. Asymmetry of plasma drift potentials near the limiter observed during the experiment leads to current drift through the limiter close to Spitzer j s one. It is shown that with the increase of mean plasma density the plasma density in the limiter channels grows and its temperature is decreased so the charged particle losses for the limiter are weakly dependent on the mean density which is connected with plasma confinement degradation under the density reduction. A complete flux of charged particles to the limiter is comparable to their flux from plasma filament. Plasma flux into the channels is close to ambipolar one and the power fluxes to neutralization plates are of the order of 10 j s Te/e. Neutral gas pressure dependence in the volume under the limiter on the plasma fluxes to channels is nonlinear, the maximum pressure achieves 3x10 -2 T

  18. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

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

  19. Impact of E × B flow shear on turbulence and resulting power fall-off width in H-mode plasmas in experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Q. Q., E-mail: yangqq@ipp.ac.cn; Zhong, F. C., E-mail: gsxu@ipp.ac.cn, E-mail: fczhong@dhu.edu.cn; Jia, M. N. [College of Science, Donghua University, Shanghai 201620 (China); Xu, G. S., E-mail: gsxu@ipp.ac.cn, E-mail: fczhong@dhu.edu.cn; Wang, L.; Wang, H. Q.; Chen, R.; Yan, N.; Liu, S. C.; Chen, L.; Li, Y. L.; Liu, J. B. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2015-06-15

    The power fall-off width in the H-mode scrape-off layer (SOL) in tokamaks shows a strong inverse dependence on the plasma current, which was noticed by both previous multi-machine scaling work [T. Eich et al., Nucl. Fusion 53, 093031 (2013)] and more recent work [L. Wang et al., Nucl. Fusion 54, 114002 (2014)] on the Experimental Advanced Superconducting Tokamak. To understand the underlying physics, probe measurements of three H-mode discharges with different plasma currents have been studied in this work. The results suggest that a higher plasma current is accompanied by a stronger E×B shear and a shorter radial correlation length of turbulence in the SOL, thus resulting in a narrower power fall-off width. A simple model has also been applied to demonstrate the suppression effect of E×B shear on turbulence in the SOL and shows relatively good agreement with the experimental observations.

  20. Influence of INCONEL 625 composition on the activation characteristics of the vacuum vessel of experimental fusion tokamaks

    International Nuclear Information System (INIS)

    Cambi, G.; Cepraga, D.G.; Boeriu, S.; Maganzani, I.

    1995-01-01

    The radioactive inventory, the decay heat and the contact dose rate of permanent components such as the vacuum vessel of two experimental fusion tokamaks, the compact IGNITOR-ULT and the ITER-EDA fusion machines, are evaluated by using the ENEA-Bologna integrated methodology. The vacuum vessel material considered is the INCONEL 625. The neutron flux is calculated using the VITAMIN-C 171-group library, based on EFF-2 data and the 1-D transport code XSDRNPM in the S 8 -P 3 approximation. The ANITA-2 code, using updated cross sections and decay data libraries based on EAF-3 and IRDF90 evaluation files is used for activation calculations. The fusion neutron source has been normalised to a neutron first wall load of 2 MW/m 2 and 1 MW/m 2 for IGNITOR-ULT and ITER, respectively. The material irradiation have been described by multistep time histories, resulting in the designed total fluence. Variations in the composition of INCONEL 625 have been assessed and their impact on the activation characteristics are discussed, also from the point of view of waste disposal. (orig.)

  1. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R S [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  2. Simulations of toroidal Alfvén eigenmode excited by fast ions on the Experimental Advanced Superconducting Tokamak

    Science.gov (United States)

    Pei, Youbin; Xiang, Nong; Shen, Wei; Hu, Youjun; Todo, Y.; Zhou, Deng; Huang, Juan

    2018-05-01

    Kinetic-MagnetoHydroDynamic (MHD) hybrid simulations are carried out to study fast ion driven toroidal Alfvén eigenmodes (TAEs) on the Experimental Advanced Superconducting Tokamak (EAST). The first part of this article presents the linear benchmark between two kinetic-MHD codes, namely MEGA and M3D-K, based on a realistic EAST equilibrium. Parameter scans show that the frequency and the growth rate of the TAE given by the two codes agree with each other. The second part of this article discusses the resonance interaction between the TAE and fast ions simulated by the MEGA code. The results show that the TAE exchanges energy with the co-current passing particles with the parallel velocity |v∥ | ≈VA 0/3 or |v∥ | ≈VA 0/5 , where VA 0 is the Alfvén speed on the magnetic axis. The TAE destabilized by the counter-current passing ions is also analyzed and found to have a much smaller growth rate than the co-current ions driven TAE. One of the reasons for this is found to be that the overlapping region of the TAE spatial location and the counter-current ion orbits is narrow, and thus the wave-particle energy exchange is not efficient.

  3. Investigation and experimental data de-noising of Damavand tokamak by using fourier series expansion and wavelet code

    International Nuclear Information System (INIS)

    Sadeghi, Y.

    2006-01-01

    Computer Programs are important tools in physics. Analysis of the experimental data and the control of complex handle physical phenomenon and the solution of numerical problem in physics help scientist to the behavior and simulate the process. In this paper, calculation of several Fourier series gives us a visual and analytic impression of data analyses from Fourier series. One of important aspect in data analyses is to find optimum method for de-noising. Wavelets are mathematical functions that cut up data into different frequency components, and then study each component with a resolution corresponding to its scale. They have advantages over usual traditional methods in analyzing physical situations where the signal contains discontinuities and sharp spikes. Transformed data by wavelets in frequency space has time information and can clearly show the exact location in time of the discontinuity. This aspect makes wavelets an excellent tool in the field of data analysis. In this paper, we show how Fourier series and wavelets can analyses data in Damavand tokamak. ?

  4. Effect of Wave Accessibility on Lower Hybrid Wave Current Drive in Experimental Advanced Superconductor Tokamak with H-Mode Operation

    International Nuclear Information System (INIS)

    Li Xin-Xia; Xiang Nong; Gan Chun-Yun

    2015-01-01

    The effect of the wave accessibility condition on the lower hybrid current drive in the experimental advanced superconductor Tokamak (EAST) plasma with H-mode operation is studied. Based on a simplified model, a mode conversion layer of the lower hybrid wave between the fast wave branch and the slow wave branch is proved to exist in the plasma periphery for typical EAST H-mode parameters. Under the framework of the lower hybrid wave simulation code (LSC), the wave ray trajectory and the associated current drive are calculated numerically. The results show that the wave accessibility condition plays an important role on the lower hybrid current drive in EAST plasma. For wave rays with parallel refractive index n ‖ = 2.1 or n ‖ = 2.5 launched from the outside midplane, the wave rays may penetrate the core plasma due to the toroidal geometry effect, while numerous reflections of the wave ray trajectories in the plasma periphery occur. However, low current drive efficiency is obtained. Meanwhile, the wave accessibility condition is improved if a higher confined magnetic field is applied. The simulation results show that for plasma parameters under present EAST H-mode operation, a significant lower hybrid wave current drive could be obtained for the wave spectrum with peak value n ‖ = 2.1 if a toroidal magnetic field B T = 2.5 T is applied. (paper)

  5. TEXT Energy Storage System

    International Nuclear Information System (INIS)

    Weldon, W.F.; Rylander, H.G.; Woodson, H.H.

    1977-01-01

    The Texas Experimental Tokamak (TEXT) Enery Storage System, designed by the Center for Electromechanics (CEM), consists of four 50 MJ, 125 V homopolar generators and their auxiliaries and is designed to power the toroidal and poloidal field coils of TEXT on a two-minute duty cycle. The four 50 MJ generators connected in series were chosen because they represent the minimum cost configuration and also represent a minimal scale up from the successful 5.0 MJ homopolar generator designed, built, and operated by the CEM

  6. Scoping studies of tritium handling in a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Cherdack, R.; Watson, J.S.; Clinton, S.D.; Fisher, P.W.

    1975-01-01

    Tritium handling techniques in an experimental fusion power reactor (EPR) are evaluated to determine the requirements of the system and to compare different equipment and techniques for meeting those requirements. Tritium process equipment is needed to (1) evacuate and maintain a vacuum in the plasma vessel and the neutral beam injectors, (2) purify and recycle tritium and deuterium for the plasma fuel cycle, (3) recover tritium from experimental breeding modules, and (4) provide tritium containment and atmospheric cleanup. A development program is outlined to develop and demonstrate the required techniques and equipment and to permit confident design of an EPR for operation by the mid-1980s

  7. Development and integration of a 50 Hz pellet injection system for the Experimental Advanced Superconducting Tokamak (EAST)

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Xingjia [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Science Island Branch of Graduate School, University of Science and Technology of China, Hefei 230029 (China); Chen, Yue [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Jiansheng, E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Vinyar, Igor; Lukin, Alexander [PELIN, Saint-Petersburg (Russian Federation); Yuan, Xiaoling; Li, Changzheng; Liu, Haiqing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2017-01-15

    Highlights: • The design of the pumping system fits the operation requirement well not only theoretically but also experimentally. • The data showed that the averaged pellet injection velocity and propellant gas pressure had a relationship submitting to the power function. • The reliability of the injected pellet was mostly around 90% which is higher than the PI-20 system thanks to the improved pumping system and the new pellet fabrication and acceleration system. - Abstract: A 50 Hz pellet injection system, which is designed for edge-localized mode (ELM) control, has been successfully developed and integrated for the Experimental Advanced Superconducting Tokamak (EAST). Pellet injection is achieved by two separated injection system modules that can be operated independently from 1 to 25 Hz. The nominal injection velocity is 250 m/s with a scatter of ±50 m/s at a repetition rate of 50 Hz. A buffer tank and a two-stage differential pumping system of the pellet injection system was designed to increase hydrogen/deuterium ice quality and eliminate the influence of propellant gas on plasma operation, respectively. The pressure of the buffer tank could be pumped to 1 × 10{sup 2} Pa, and the pressure in the second differential chamber could reach 1 × 10{sup −4} Pa during the experiment. Engineering experiments, which consisted of 50 Hz pellet injection and guiding tube mock-up experiments, were also systematically carried out in a laboratory environment and demonstrated that the pellet injection system can reliably inject pellets at a repetitive frequency of 50 Hz.

  8. Study on lower hybrid current drive efficiency at high density towards long-pulse regimes in Experimental Advanced Superconducting Tokamak

    International Nuclear Information System (INIS)

    Li, M. H.; Ding, B. J.; Zhang, J. Z.; Gan, K. F.; Wang, H. Q.; Zhang, L.; Wei, W.; Li, Y. C.; Wu, Z. G.; Ma, W. D.; Jia, H.; Chen, M.; Yang, Y.; Feng, J. Q.; Wang, M.; Xu, H. D.; Shan, J. F.; Liu, F. K.; Peysson, Y.

    2014-01-01

    Significant progress on both L- and H-mode long-pulse discharges has been made recently in Experimental Advanced Superconducting Tokamak (EAST) with lower hybrid current drive (LHCD) [J. Li et al., Nature Phys. 9, 817 (2013) And B. N. Wan et al., Nucl. Fusion 53, 104006 (2013).]. In this paper, LHCD experiments at high density in L-mode plasmas have been investigated in order to explore possible methods of improving current drive (CD) efficiency, thus to extend the operational space in long-pulse and high performance plasma regime. It is observed that the normalized bremsstrahlung emission falls much more steeply than 1/n e-av (line-averaged density) above n e-av  = 2.2 × 10 19  m −3 indicating anomalous loss of CD efficiency. A large broadening of the operating line frequency (f = 2.45 GHz), measured by a radio frequency (RF) probe located outside the EAST vacuum vessel, is generally observed during high density cases, which is found to be one of the physical mechanisms resulting in the unfavorable CD efficiency. Collisional absorption of lower hybrid wave in the scrape off layer (SOL) may be another cause, but this assertion needs more experimental evidence and numerical analysis. It is found that plasmas with strong lithiation can improve CD efficiency largely, which should be benefited from the changes of edge parameters. In addition, several possible methods are proposed to recover good efficiency in future experiments for EAST

  9. 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

  10. The density limit in Tokamaks

    International Nuclear Information System (INIS)

    Alladio, F.

    1985-01-01

    A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs

  11. START: the creation of a spherical tokamak

    International Nuclear Information System (INIS)

    Sykes, Alan

    1992-01-01

    The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)

  12. Conditional analysis of floating potential fluctuations at the edge of the Texas Experimental Tokamak Upgrade (TEXT-U)

    International Nuclear Information System (INIS)

    Filippas, A.V.; Bengston, R.D.; Li, G.; Meier, M.; Ritz, C.P.; Powers, E.J.

    1995-01-01

    Fluctuations in floating potential in the scrape-off layer and plasma edge were analyzed using a conditional statistical analysis technique. The floating potential fluctuations had a nearly Gaussian probability density function with the largest deviation from a Gaussian at the shear layer. The conditional averaging technique followed the statistical evolution of selected conditions in the floating potential signal. The decay rate of a conditional feature in time or space showed a small systematic variation with the amplitude of condition chosen. Either long-lived coherent structures are not present in statistically significant numbers, or the fluctuations are dominated by a large number of coherent structures with a nearly Gaussian distribution of fluctuation amplitudes, or conditional analysis using the amplitude of the floating potential as a condition is not a sensitive technique for identifying coherent structures

  13. New System For Tokamak T-10 Experimental Data Acquisition, Data Handling And Remote Access

    International Nuclear Information System (INIS)

    Sokolov, M. M.; Igonkina, G. B.; Koutcherenko, I. Yu.; Nurov, D. N.

    2008-01-01

    For carrying out the experiments on nuclear fusion devices in the Institute of Nuclear Fusion, Moscow, a system for experimental data acquisition, data handling and remote access (further 'DAS-T10') was developed and has been used in the Institute since the year 2000. The DAS-T10 maintains the whole cycle of experimental data handling: from configuration of data measuring equipment and acquisition of raw data from the fusion device (the Device), to presentation of math-processed data and support of the experiment data archive. The DAS-T10 provides facilities for the researchers to access the data both at early stages of an experiment and well afterwards, locally from within the experiment network and remotely over the Internet.The DAS-T10 is undergoing a modernization since the year 2007. The new version of the DAS-T10 will accommodate to modern data measuring equipment and will implement improved architectural solutions. The innovations will allow the DAS-T10 to produce and handle larger amounts of experimental data, thus providing the opportunities to intensify and extend the fusion researches. The new features of the DAS-T10 along with the existing design principles are reviewed in this paper

  14. An Experimental Text in Transformational Geometry, Student Text; Cambridge Conference on School Mathematics Feasibility Study No. 43a.

    Science.gov (United States)

    Cambridge Conference on School Mathematics, Newton, MA.

    This is part of a student text which was written with the aim of reflecting the thinking of The Cambridge Conference on School Mathematics (CCSM) regarding the goals and objectives for mathematics. The instructional materials were developed for teaching geometry in the secondary schools. This document is chapter six and titled Motions and…

  15. Experimental studies of processes with vibrationally excited hydrogen molecules that are important for tokamak edge plasma

    International Nuclear Information System (INIS)

    Cadez, I.; Markelj, S.; Rupnik, Z.; Pelicon, P.

    2006-01-01

    We are currently conducting a series of different laboratory experimental studies of processes involving vibrationally excited hydrogen molecules that are relevant to fusion edge plasma. A general overview of our activities is presented together with results of studies of hydrogen recombination on surfaces. This includes vibrational spectroscopy of molecules formed by recombination on metal surfaces exposed to the partially dissociated hydrogen gas and recombination after hydrogen permeation through metal membrane. The goal of these studies is to provide numerical parameters needed for edge plasma modelling and better understanding of plasma wall interaction processes. (author)

  16. Experimental studies and modelling of high radiation and high density plasmas in the ASDEX upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Casali, Livia

    2015-11-24

    Fusion plasmas contain impurities, either intrinsic originating from the wall, or injected willfully with the aim of reducing power loads on machine components by converting heat flux into radiation. The understanding and the prediction of the effects of these impurities and their radiation on plasma performances is crucial in order to retain good confinement. In addition, it is important to understand the impact of pellet injection on plasma performance since this technique allows higher core densities which are required to maximise the fusion power. This thesis contributes to these efforts through both experimental investigations and modelling. Experiments were conducted at ASDEX Upgrade which has a full-W wall. Impurity seeding was applied to H-modes by injecting nitrogen and also medium-Z impurities such as Kr and Ar to assess the impact of both edge and central radiation on confinement. A database of about 25 discharges has been collected and analysed. A wide range of plasma parameters was achieved up to ITER relevant values such as high Greenwald and high radiation fractions. Transport analyses taking into account the radiation distribution reveal that edge localised radiation losses do not significantly impact confinement as long as the H-mode pedestal is sustained. N seeding induces higher pedestal pressure which is propagated to the core via profile stiffness. Central radiation must be limited and controlled to avoid confinement degradation. This requires reliable control of the impurity concentration but also possibilities to act on the ELM frequency which must be kept high enough to avoid an irreversible impurity accumulation in the centre and the consequent radiation collapse. The key role of the f{sub ELM} is confirmed also by the analysis of N+He discharges. Non-coronal effects affect the radiation of low-Z impurities at the plasma edge. Due to the radial transport, the steep temperature gradients and the ELM flush out, a local equilibrium cannot be

  17. 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

  18. Comparison of explicit calculations for n = 3 to 8 dielectronic satellites of the FeXXV Kα resonance line with experimental data from the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Decaux, V.; Bitter, M.; Hsuan, H.; Hill, K.W.; von Goeler, S.; Park, H.; Bhalla, C.P.

    1991-12-01

    Dielectronic satellite spectra of the FeXXV Kα resonance line observed from the Tokamak Fusion Test Reactor (TFTR) plasmas have been compared with recent explicit calculations for the n = 3 to 8 dielectronic satellites as well as the earlier theoretical predictions, which were based on the 1/n 3 scaling law for n > 4 satellites. The analysis has been performed by least-squares fits of synthetic spectra to the experimental data. The synthetic spectra constructed from both theories are in good agreement with the observed data. However, the electron temperature values obtained from the fit of the present explicit calculations are in better agreement with independent measurements. 20 refs., 4 figs

  19. Combined Quitline Counseling and Text Messaging for Smoking Cessation: A Quasi-Experimental Evaluation.

    Science.gov (United States)

    Boal, Ashley L; Abroms, Lorien C; Simmens, Samuel; Graham, Amanda L; Carpenter, Kelly M

    2016-05-01

    This study seeks to determine whether comprehensive quitline services combined with text messaging improve smoking cessation rates beyond those achieved by offering comprehensive quitline services alone. The study sample consisted of callers to the Alere Wellbeing, Inc, commercial quitline in 2012. A quasi-experimental design was implemented using propensity score matching to create the intervention and control groups. The intervention group consisted of those who were offered and accepted a text message intervention in addition to usual quitline services, while the control group consisted of those who were not offered the text message intervention. Analyses utilized baseline data collected at intake, program use data (eg, call history and text message use), and reports of smoking behaviors and program satisfaction collected 6 months after intake. Similar rates of 7-day abstinence were reported regardless of whether participants received combined multi-call quitline services plus text messaging (25.3%) or multi-call quitline services in isolation (25.5%), though those who received combined services reported higher treatment satisfaction (P research should investigate whether text messaging programs improve quit rates when combined with less intensive services such as single-call phone counseling. While the impact of quitline and text messaging services for smoking cessation have been examined in isolation, no study has explored the impact of combined services on smoking outcomes. This study examines the role of text messaging in combination with comprehensive quitline services including multi-call phone counseling, access to an interactive website and nicotine replacement therapy. © The Author 2015. Published by Oxford University Press on behalf of the Society for Research on Nicotine and Tobacco. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  20. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1989-01-01

    A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

  1. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    McWilliams, R.

    1988-01-01

    The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  2. Internal disruption in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1990-01-01

    A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described

  3. Internal disruptions in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1990-01-01

    Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described

  4. A fast-time-response extreme ultraviolet spectrometer for measurement of impurity line emissions in the Experimental Advanced Superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Ling; Xu, Zong; Wu, Zhenwei; Zhang, Pengfei; Wu, Chengrui; Gao, Wei; Shen, Junsong; Chen, Yingjie; Liu, Xiang; Wang, Yumin; Gong, Xianzu; Hu, Liqun; Chen, Junlin; Zhang, Xiaodong; Wan, Baonian; Li, Jiangang [Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230026, Anhui (China); Morita, Shigeru; Ohishi, Tetsutarou; Goto, Motoshi [National Institute for Fusion Science, Toki 509-5292, Gifu (Japan); Department of Fusion Science, Graduate University for Advanced Studies, Toki 509-5292, Gifu (Japan); Dong, Chunfeng [Southwestern Institute of Physics, Chengdu 610041, Sichuan (China); and others

    2015-12-15

    A flat-field extreme ultraviolet (EUV) spectrometer working in the 20-500 Å wavelength range with fast time response has been newly developed to measure line emissions from highly ionized tungsten in the Experimental Advanced Superconducting Tokamak (EAST) with a tungsten divertor, while the monitoring of light and medium impurities is also an aim in the present development. A flat-field focal plane for spectral image detection is made by a laminar-type varied-line-spacing concave holographic grating with an angle of incidence of 87°. A back-illuminated charge-coupled device (CCD) with a total size of 26.6 × 6.6 mm{sup 2} and pixel numbers of 1024 × 255 (26 × 26 μm{sup 2}/pixel) is used for recording the focal image of spectral lines. An excellent spectral resolution of Δλ{sub 0} = 3-4 pixels, where Δλ{sub 0} is defined as full width at the foot position of a spectral line, is obtained at the 80-400 Å wavelength range after careful adjustment of the grating and CCD positions. The high signal readout rate of the CCD can improve the temporal resolution of time-resolved spectra when the CCD is operated in the full vertical binning mode. It is usually operated at 5 ms per frame. If the vertical size of the CCD is reduced with a narrow slit, the time response becomes faster. The high-time response in the spectral measurement therefore makes possible a variety of spectroscopic studies, e.g., impurity behavior in long pulse discharges with edge-localized mode bursts. An absolute intensity calibration of the EUV spectrometer is also carried out with a technique using the EUV bremsstrahlung continuum at 20-150 Å for quantitative data analysis. Thus, the high-time resolution tungsten spectra have been successfully observed with good spectral resolution using the present EUV spectrometer system. Typical tungsten spectra in the EUV wavelength range observed from EAST discharges are presented with absolute intensity and spectral identification.

  5. The Texas Experimental Tokamak: A plasma research facility. A proposal submitted to the Department of Energy in response to Program Notice 95-10: Innovations in toroidal magnetic confinement systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-12

    The Fusion Research Center (FRC) at the University Texas will operate the tokamak TEXT-U and its associated systems for experimental research in basic plasma physics. While the tokamak is not innovative, the research program, diagnostics and planned experiments are. The fusion community will reap the benefits of the success in completing the upgrades (auxiliary heating, divertor, diagnostics, wall conditioning), developing diverted discharges in both double and single null configurations, exploring improved confinement regimes including a limiter H-mode, and developing unique, critical turbulence diagnostics. With these new regimes, the authors are poised to perform the sort of turbulence and transport studies for which the TEXT group has distinguished itself and for which the upgrade was intended. TEXT-U is also a facility for collaborators to perform innovative experiments and develop diagnostics before transferring them to larger machines. The general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The emphasis of the proposed research is to provide well-documented plasmas which will be used to suggest and evaluate theories, to explore control techniques, to develop advanced diagnostics and analysis techniques, and to extend current drive techniques. Up to 1 MW of electron cyclotron heating (ECH) will be used not only for heating but as a localized, perturbative tool. Areas of proposed research are: (1) core turbulence and transport; (2) edge turbulence and transport; (3) turbulence analysis; (4) improved confinement; (5) ECH physics; (6) Alfven wave current drive; and (7) diagnostic development.

  6. The Texas Experimental Tokamak: A plasma research facility. A proposal submitted to the Department of Energy in response to Program Notice 95-10: Innovations in toroidal magnetic confinement systems

    International Nuclear Information System (INIS)

    1995-01-01

    The Fusion Research Center (FRC) at the University Texas will operate the tokamak TEXT-U and its associated systems for experimental research in basic plasma physics. While the tokamak is not innovative, the research program, diagnostics and planned experiments are. The fusion community will reap the benefits of the success in completing the upgrades (auxiliary heating, divertor, diagnostics, wall conditioning), developing diverted discharges in both double and single null configurations, exploring improved confinement regimes including a limiter H-mode, and developing unique, critical turbulence diagnostics. With these new regimes, the authors are poised to perform the sort of turbulence and transport studies for which the TEXT group has distinguished itself and for which the upgrade was intended. TEXT-U is also a facility for collaborators to perform innovative experiments and develop diagnostics before transferring them to larger machines. The general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The emphasis of the proposed research is to provide well-documented plasmas which will be used to suggest and evaluate theories, to explore control techniques, to develop advanced diagnostics and analysis techniques, and to extend current drive techniques. Up to 1 MW of electron cyclotron heating (ECH) will be used not only for heating but as a localized, perturbative tool. Areas of proposed research are: (1) core turbulence and transport; (2) edge turbulence and transport; (3) turbulence analysis; (4) improved confinement; (5) ECH physics; (6) Alfven wave current drive; and (7) diagnostic development

  7. 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

  8. Tokamak experimental section

    International Nuclear Information System (INIS)

    Berry, L.A.; Dunlap, J.L.; Arakawa, E.T.

    1977-01-01

    Descriptions of research during this period are given for the following topics: (1) ion and electron heating, (2) high-beta and gas puff experiments, (3) beam trapping by impurities, (4) counterinjection studies, (5) impurity measurements, (6) Balmer alpha line profiles, (7) internal mode structure, (8) sawtooth oscillations and plasma transport, (9) Ormak plasma modeling, (10) charge exchange measurements, (11) wall power measurements, (12) neutron time behavior due to deuterium neutral beam injection into a hydrogen plasma, (13) wall impurities in Ormak, (14) relativistic electron studies, (15) fast x-ray energy analyzer for the 1 to 10 keV range, and (16) CTR related atomic physics

  9. Identifying Understudied Nuclear Reactions by Text-mining the EXFOR Experimental Nuclear Reaction Library

    Energy Technology Data Exchange (ETDEWEB)

    Hirdt, J.A. [Department of Mathematics and Computer Science, St. Joseph' s College, Patchogue, NY 11772 (United States); Brown, D.A., E-mail: dbrown@bnl.gov [National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY 11973-5000 (United States)

    2016-01-15

    The EXFOR library contains the largest collection of experimental nuclear reaction data available as well as the data's bibliographic information and experimental details. We text-mined the REACTION and MONITOR fields of the ENTRYs in the EXFOR library in order to identify understudied reactions and quantities. Using the results of the text-mining, we created an undirected graph from the EXFOR datasets with each graph node representing a single reaction and quantity and graph links representing the various types of connections between these reactions and quantities. This graph is an abstract representation of the connections in EXFOR, similar to graphs of social networks, authorship networks, etc. We use various graph theoretical tools to identify important yet understudied reactions and quantities in EXFOR. Although we identified a few cross sections relevant for shielding applications and isotope production, mostly we identified charged particle fluence monitor cross sections. As a side effect of this work, we learn that our abstract graph is typical of other real-world graphs.

  10. Identifying Understudied Nuclear Reactions by Text-mining the EXFOR Experimental Nuclear Reaction Library

    International Nuclear Information System (INIS)

    Hirdt, J.A.; Brown, D.A.

    2016-01-01

    The EXFOR library contains the largest collection of experimental nuclear reaction data available as well as the data's bibliographic information and experimental details. We text-mined the REACTION and MONITOR fields of the ENTRYs in the EXFOR library in order to identify understudied reactions and quantities. Using the results of the text-mining, we created an undirected graph from the EXFOR datasets with each graph node representing a single reaction and quantity and graph links representing the various types of connections between these reactions and quantities. This graph is an abstract representation of the connections in EXFOR, similar to graphs of social networks, authorship networks, etc. We use various graph theoretical tools to identify important yet understudied reactions and quantities in EXFOR. Although we identified a few cross sections relevant for shielding applications and isotope production, mostly we identified charged particle fluence monitor cross sections. As a side effect of this work, we learn that our abstract graph is typical of other real-world graphs.

  11. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

    This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

  12. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

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

  13. 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

  14. Experimental measurement of magnetic field null in the vacuum chamber of KTM tokamak based on matrix of 2D Hall sensors

    Energy Technology Data Exchange (ETDEWEB)

    Shapovalov, G.; Chektybayev, B., E-mail: chektybaev@nnc.kz; Sadykov, A.; Skakov, M.; Kupishev, E.

    2016-11-15

    Experimental technique of measurement of magnetic field null region inside of the KTM tokamak vacuum chamber has been developed. Square matrix of 36 2D Hall sensors, which used in the technique, allows carrying out direct measurements of poloidal magnetic field dynamics in the vacuum chamber. To better measuring accuracy, Hall sensor’s matrix was calibrated with commercial Helmholtz coils and in situ measurement of defined magnetic field from poloidal and toroidal coils. Standard KTM Data-Acquisition System has been used to collect data from Hall sensors. Experimental results of measurement of magnetic field null in the vacuum chamber of KTM are shown in the paper. Additionally results of the magnetic field null reconstruction from signals of inductive total flux loops are shown in the paper.

  15. Experimental identification of nonlinear coupling between (intermediate, small)-scale microturbulence and an MHD mode in the core of a superconducting tokamak

    Science.gov (United States)

    Sun, P. J.; Li, Y. D.; Ren, Y.; Zhang, X. D.; Wu, G. J.; Xu, L. Q.; Chen, R.; Li, Q.; Zhao, H. L.; Zhang, J. Z.; Shi, T. H.; Wang, Y. M.; Lyu, B.; Hu, L. Q.; Li, J.; The EAST Team

    2018-01-01

    In this paper, we present clear experimental evidence of core region nonlinear coupling between (intermediate, small)-scale microturbulence and an magnetohydrodynamics (MHD) mode during the current ramp-down phase in a set of L-mode plasma discharges in the experimental advanced superconducting tokamak (EAST, Wan et al (2006 Plasma Sci. Technol. 8 253)). Density fluctuations of broadband microturbulence (k\\perpρi˜2{-}5.2 ) and the MHD mode (toroidal mode number m = -1 , poloidal mode number n = 1 ) are measured simultaneously, using a four-channel tangential CO2 laser collective scattering diagnostic in core plasmas. The nonlinear coupling between the broadband microturbulence and the MHD mode is directly demonstrated by showing a statistically significant bicoherence and modulation of turbulent density fluctuation amplitude by the MHD mode.

  16. Experimental determination of some equilibrium parameter of Damavand tokamak by magnetic probe measurements for representing a physical model for plasma vertical movement.

    Science.gov (United States)

    Farahani, N Darestani; Davani, F Abbasi

    2015-10-01

    This investigation is about plasma modeling for the control of vertical instabilities in Damavand tokamak. This model is based on online magnetic measurement. The algebraic equation defining the vertical position in this model is based on instantaneous force-balance. Two parameters in this equation, including decay index, n, and lambda, Λ, have been considered as functions of time-varying poloidal field coil currents and plasma current. Then these functions have been used in a code generated for modeling the open loop response of plasma. The main restriction of the suitability analysis of the model is that the experiments always have to be performed in the presence of a control loop for stabilizing vertical position. As a result, open loop response of the system has been identified from closed loop experimental data by nonlinear neural network identification method. The results of comparison of physical model with identified open loop response from closed loop experiments show root mean square error percentage less than 10%. The results are satisfying that the physical model is useful as a Damavand tokamak vertical movement simulator.

  17. Experimental evidence of significant temperature fluctuations in the plasma EDGE region of the TJ-I Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hidalgo, C; Balbin, R; Pedrosa, M A; Garcia-Cortes, I; Ochando, M A

    1993-07-01

    Density and temperature fluctuations have been measured in the plasma bulk side of the velocity shear location of the TJ-I tokamak using a feast swept Langmuir probe technique. Evidence of substantial temperature fluctuations which are in phase close to opposition with the corresponding density fluctuations has been found. This result suggests the possible role of radiation in determining edge fluctuation levels and call into question the determination of the density and potential fluctuations from the Langmuir current-probe and floating potential fluctuations. (Author) 16 refs.

  18. Experimental evidence of significant temperature fluctuations in the plasma edge region of the TJ-I Tokamak

    International Nuclear Information System (INIS)

    Hidalgo, C.; Balbin, R.; Pedrosa, M.A.; Garcia-Cortes, I.; Ochando, M.A.

    1993-01-01

    Density and temperature fluctuations have been measured in the plasma bulk side of the velocity shear location of the TJ-I tokamak using a foast swept Langmuir probe technique. Evidence of sustantial temperature fluctuations which are in phase close to opposition with the corresponding density fluctuations has been found. This result suggests the possible role of radiation in determining edge fluctuation levels and call into question the determination of the density and potential fluctuations from the Langmuir current-probe and floating potential fluctuations. (Author)

  19. Experimental evidence of significant temperature fluctuations in the plasma EDGE region of the TJ-I Tokamak

    International Nuclear Information System (INIS)

    Hidalgo, C.; Balbin, R.; Pedrosa, M. A.; Garcia-Cortes, I.; Ochando, M. A.

    1993-01-01

    Density and temperature fluctuations have been measured in the plasma bulk side of the velocity shear location of the TJ-I tokamak using a feast swept Langmuir probe technique. Evidence of substantial temperature fluctuations which are in phase close to opposition with the corresponding density fluctuations has been found. This result suggests the possible role of radiation in determining edge fluctuation levels and call into question the determination of the density and potential fluctuations from the Langmuir current-probe and floating potential fluctuations. (Author) 16 refs

  20. Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak; Etude experimentale des aspects topologiques du divertor ergodique de Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Costanzo, L

    2001-10-01

    The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor {gamma} was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that {gamma}=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a

  1. Neural net prediction of tokamak plasma disruptions

    International Nuclear Information System (INIS)

    Hernandez, J.V.; Lin, Z.; Horton, W.; McCool, S.C.

    1994-10-01

    The computation based on neural net algorithms in predicting minor and major disruptions in TEXT tokamak discharges has been performed. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal, measured by a single Mirnov coil. The time step used (= 0.04ms) corresponds to the experimental data sampling rate. Two kinds of approaches are adopted for the task, the contiguous future prediction and the multi-timescale prediction. Results are shown for comparison. Both networks are trained through the back-propagation algorithm with inertial terms. The degree of this success indicates that the magnetic fluctuations associated with tokamak disruptions may be characterized by a relatively low-dimensional dynamical system

  2. Experimental study on the practicability of a self-created spherical tokamak in the coil less STPC-EX machine

    International Nuclear Information System (INIS)

    Sinman, S.; Sinman, A.

    2003-01-01

    The aim of this study is to identify the physical basis of the alternative self-organization mechanism that exists on the STPC-EX machine and to determine complementary features with respect to present compact toroid concepts. In the STPC-EX machine, there exist two solenoids placed inside the central passive floating conductive hollow rod and externally onto flux conserver. These are in a passive state and they do not have an important role in the self-created spherical tokamak plasma (SCSTP) in the STPC-EX machine. In this study, conventional diagnostic tools are used and for photographic recording, the method of open shutter integrated post-fogging is chosen. Two annular coaxial plasma current sheets, one within the other in the same direction, are created and flow on the surface of the central conductive hollow rod. Consequently, the spherical tokamak is configured by a new creation mechanism called the dual-axial z-pinch. High betas of 0.4-0.6 and aspect ratios of up to 1.25 can be obtained. (author)

  3. 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

  4. Modification of tokamak edge turbulence using feedback

    International Nuclear Information System (INIS)

    Richards, B.; Uckan, T.; Wootton, A.J.; Carreras, B.A.; Bengtson, R.D.; Hurwitz, P.; Li, G.X.; Lin, H.; Rowan, W.L.; Tsui, H.Y.W.; Sen, A.K.; Uglum, J.

    1994-01-01

    Using active feedback, the turbulent fluctuation levels have been reduced by as much as a factor of 2 in the edge of the Texas Experimental Tokamak (TEXT) [K. W. Gentle, Nucl. Fusion Technol. 1, 479 (1981)]. A probe system was used to drive a suppressor wave in the TEXT limiter shadow. A decrease in the local turbulence-induced particle flux has been seen, but a global change in the particle transport at the present time has not been observed. By changing the phase shift and gain of the feedback network, the amplitude of the turbulence was increased by a factor of 10

  5. Edge turbulence and transport: Text and ATF modeling

    International Nuclear Information System (INIS)

    Ritz, C.P.; Rhodes, T.L.; Lin, H.; Rowan, W.L.; Bengtson, R.; Wootton, A.J.; Diamond, P.H.; Ware, A.S.; Thayer, D.R.

    1990-01-01

    We present experimental results on edge turbulence and transport from the tokamak TEXT and the torsatron ATF. The measured electrostatic fluctuations can explain the edge transport of particles and energy. Certain drive (radiation) and stabilizing (velocity shear) terms are suggested by the results. The experimental fluctuation levels and spectral widths can be reproduced by considering the nonlinear evolution of the reduced MHD equations, incorporating a thermal drive from line radiation. In the tokamak limit (with toroidal electric field) the model corresponds to the resistivity gradient mode, while in the currentless torsatron or stellarator limit it corresponds to a thermally driven drift wave

  6. Modeling of thermal effects on TIBER II [Tokamak Ignition/Burn Experimental Reactor] divertor during plasma disruption

    International Nuclear Information System (INIS)

    Bruhn, M.L.; Perkins, L.J.

    1987-01-01

    Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs. 14 refs

  7. Experimental Study of Reversed Shear Alfven Eigenmodes During The Current Ramp In The Alcator C-Mod Tokamak

    International Nuclear Information System (INIS)

    Edlund, E.M.; Porkolab, M.; Kramer, G.J.; Lin, L.; Lin, Y.; Tsuji, N.; Wukitch, S.J.

    2010-01-01

    Experiments conducted in the Alcator C-Mod tokamak at MIT have explored the physics of reversed shear Alfven eigenmodes (RSAEs) during the current ramp. The frequency evolution of the RSAEs throughout the current ramp provides a constraint on the evolution of q min , a result which is important in transport modeling and for comparison with other diagnostics which directly measure the magnetic field line structure. Additionally, a scaling of the RSAE minimum frequency with the sound speed is used to derive a measure of the adiabatic index, a measure of the plasma compressibility. This scaling bounds the adiabatic index at 1.40 ± 0.15 used in MHD models and supports the kinetic calculation of separate electron and ion compressibilities with an ion adiabatic index close to 7/4.

  8. Experimental validation of a Lyapunov-based controller for the plasma safety factor and plasma pressure in the TCV tokamak

    Science.gov (United States)

    Mavkov, B.; Witrant, E.; Prieur, C.; Maljaars, E.; Felici, F.; Sauter, O.; the TCV-Team

    2018-05-01

    In this paper, model-based closed-loop algorithms are derived for distributed control of the inverse of the safety factor profile and the plasma pressure parameter β of the TCV tokamak. The simultaneous control of the two plasma quantities is performed by combining two different control methods. The control design of the plasma safety factor is based on an infinite-dimensional setting using Lyapunov analysis for partial differential equations, while the control of the plasma pressure parameter is designed using control techniques for single-input and single-output systems. The performance and robustness of the proposed controller is analyzed in simulations using the fast plasma transport simulator RAPTOR. The control is then implemented and tested in experiments in TCV L-mode discharges using the RAPTOR model predicted estimates for the q-profile. The distributed control in TCV is performed using one co-current and one counter-current electron cyclotron heating actuation.

  9. Print versus digital texts: understanding the experimental research and challenging the dichotomies

    Directory of Open Access Journals (Sweden)

    Bella Ross

    2017-11-01

    Full Text Available This article presents the results of a systematic critical review of interdisciplinary literature concerned with digital text (or e-text uses in education and proposes recommendations for how e-texts can be implemented for impactful learning. A variety of e-texts can be found in the repertoire of educational resources accessible to students, and in the constantly changing terrain of educational technologies, they are rapidly evolving, presenting new opportunities and affordances for student learning. We highlight some of the ways in which academic studies have examined e-texts as part of teaching and learning practices, placing a particular emphasis on aspects of learning such as recall, comprehension, retention of information and feedback. We also review diverse practices associated with uses of e-text tools such as note-taking, annotation, bookmarking, hypertexts and highlighting. We argue that evidence-based studies into e-texts are overwhelmingly structured around reinforcing the existing dichotomy pitting print-based (‘traditional’ texts against e-texts. In this article, we query this approach and instead propose to focus on factors such as students’ level of awareness of their options in accessing learning materials and whether they are instructed and trained in how to take full advantage of the capabilities of e-texts, both of which have been found to affect learning performance.

  10. Resonant island divertor experiments on text

    International Nuclear Information System (INIS)

    deGrassie, J.S.; Evans, T.E.; Jackson, G.L.

    1988-09-01

    The first experimental tests of the resonant island divertor (RID) concept have been carried out on the Texas Experimental Tokamak (TEXT). Modular perturbation coils produce static resonant magnetic fields at the tokamak boundary. The resulting magnetic islands are used to guide heat and particle fluxes around a small scoop limiter head. An enhancement in the limiter collection efficiency over the nonisland operation, as evidenced by enhanced neutral density within the limiter head, of up to a factor of 4 is obtained. This enhancement is larger than one would expect given the measured magnitude of the cross-field particle transport in TEXT. It is proposed that electrostatic perturbations occur which enhance the ion convection rate around the islands. Preliminary experiments utilizing electron cyclotron heating (ECH) in conjunction with RID operation have also have been performed. 6 refs., 3 figs

  11. Biased limiter experiments on text

    International Nuclear Information System (INIS)

    Phillips, P.E.; Wootton, A.J.; Rowan, W.L.; Ritz, C.P.; Rhodes, T.L.; Bengtson, R.D.; Hodge, W.L.; Durst, R.D.; McCool, S.C.; Richards, B.; Gentle, K.W.; Schoch, P.; Forster, J.C.; Hickok, R.L.; Evans, T.E.

    1987-01-01

    Experiments using an electrically biased limiter have been performed on the Texas Experimental Tokamak (TEXT). A small movable limiter is inserted past the main poloidal ring limiter (which is electrically connected to the vacuum vessel) and biased at V Lim with respect to it. The floating potential, plasma potential and shear layer position can be controlled. With vertical strokeV Lim vertical stroke ≥ 50 V the plasma density increases. For V Lim Lim > 0 the results obtained are inconclusive. Variation of V Lim changes the electrostatic turbulence which may explain the observed total flux changes. (orig.)

  12. Design of Fire/Gas Penetration Seals and fire exposure tests for Tokamak Fusion Test Reactor experimental areas

    International Nuclear Information System (INIS)

    Cavalluzzo, S.

    1983-01-01

    A Fire/Gas Penetration Seal is required in every penetration through the walls and ceilings into the Test Cell housing the Tokamak Fusion Test Reactor (TFTR), as well as other adjacent areas to protect the TFTR from fire damage. The penetrations are used for field coil lead stems, diagnostics systems, utilities, cables, trays, mechanical devices, electrical conduits, vacuum liner, air conditioning ducts, water pipes, and gas pipes. The function of the Fire/Gas Penetration Seals is to prevent the passage of fire and products of combustion through penetrations for a period of time up to three hours and remain structurally intact during fire exposure. The Penetration Seal must withstand, without rupture, a fire hose water stream directed at the hot surface. There are over 3000 penetrations ranging in size from several square inches to 100 square feet, and classified into 90 different types. The material used to construct the Fire/Gas Penetration Seals consist of a single and a two-component room temperature vulcanizing (RTV) silicone rubber compound. Miscellaneous materials such as alumina silica refractory fibers in board, blanket and fiber forms are also used in the construction and assembly of the Seals. This paper describes some of the penetration seals and the test procedures used to perform the three-hour fire exposure tests to demonstrate the adequacy of the seals

  13. An Experimental Text in Transformational Geometry; Teachers' Guide; Cambridge Conference on School Mathematics Feasibility Study No. 43b.

    Science.gov (United States)

    Cambridge Conference on School Mathematics, Newton, MA.

    This teachers' guide was written to be used in conjunction with the student text, An Experimental Text in Transformational Geometry. The guide is intended to help teachers who have responsibility for teaching the topics Motions and Transformations in the Plane. Each section commences with a general discussion concerning the major ideas which are…

  14. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (n e and T e ) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  15. Application of Ferulic Acid for Alzheimer's Disease: Combination of Text Mining and Experimental Validation.

    Science.gov (United States)

    Meng, Guilin; Meng, Xiulin; Ma, Xiaoye; Zhang, Gengping; Hu, Xiaolin; Jin, Aiping; Zhao, Yanxin; Liu, Xueyuan

    2018-01-01

    Alzheimer's disease (AD) is an increasing concern in human health. Despite significant research, highly effective drugs to treat AD are lacking. The present study describes the text mining process to identify drug candidates from a traditional Chinese medicine (TCM) database, along with associated protein target mechanisms. We carried out text mining to identify literatures that referenced both AD and TCM and focused on identifying compounds and protein targets of interest. After targeting one potential TCM candidate, corresponding protein-protein interaction (PPI) networks were assembled in STRING to decipher the most possible mechanism of action. This was followed by validation using Western blot and co-immunoprecipitation in an AD cell model. The text mining strategy using a vast amount of AD-related literature and the TCM database identified curcumin, whose major component was ferulic acid (FA). This was used as a key candidate compound for further study. Using the top calculated interaction score in STRING, BACE1 and MMP2 were implicated in the activity of FA in AD. Exposure of SHSY5Y-APP cells to FA resulted in the decrease in expression levels of BACE-1 and APP, while the expression of MMP-2 and MMP-9 increased in a dose-dependent manner. This suggests that FA induced BACE1 and MMP2 pathways maybe novel potential mechanisms involved in AD. The text mining of literature and TCM database related to AD suggested FA as a promising TCM ingredient for the treatment of AD. Potential mechanisms interconnected and integrated with Aβ aggregation inhibition and extracellular matrix remodeling underlying the activity of FA were identified using in vitro studies.

  16. Liminality in J.M. Coetzee’s later experimental texts

    Directory of Open Access Journals (Sweden)

    Annemie Grobler

    2015-07-01

    Full Text Available Postcoloniality, which is essentially concerned with the transition and transgression of boundaries and borders, contextualises and defines liminality as an ephemeral concept that eludes pinning down. Liminality is continually involved in a dual process of evolving and subverting: dynamic in the sense of promoting the centre, but subversive in its destabilisation of the previous status quo. In the more recent novels by Coetzee (Slow man, Diary of a bad year, Elizabeth Costello, themes that are especially acute in highlighting the subversive nature of liminality emerge repeatedly: the threshold, death, proliferation and imitation. The problem investigated is: how is the concept of liminality developed in these novels? An examination of these novels in terms of the above-mentioned themes and various conceptual and theoretical elements shows that Coetzee encourages the reader to assume a liminal status, not only as reader of the texts but also in relation to contemporary reality. It is concluded that Coetzee uses certain themes that promote liminality, often in a subversive and deconstructive manner, to inform the reader and, thus, influence him or her to effect change in the contemporary world.

  17. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    Caldas, I.L.; Vannucci, A.

    1985-01-01

    A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.) [pt

  18. Design and experimental results of feedback control of Ohmic-heating transformer magnetic flux by LHCD power in HT-7 Tokamak

    International Nuclear Information System (INIS)

    Yiyun Huang

    2006-01-01

    In order to make a research on long pulse or even steady state operation with non-inductive drive in plasma discharge, a new feedback control scheme instead of the previous one has been designed and operated in HT-7 [HT-7 team presented by J. Li, et al., Plasma Phys. Control. Fusion 42 (2) (2000) 135-146] Tokamak experiment, 2004. Consumption of iron-core transformer magnetic flux (MFT) is feedback controlled for the first time by power of lower hybrid current drive (LHCD) P LH , when the Ohmic-heating circuit current can maintain the plasma current I P constant with another feedback control loop, which make MFT evolve at alternating-change state to avoid flux saturation. Plasma current I P can be maintained steadily up to 120s in this operation mode at reduced plasma parameters (I P ∼50-100KA, average density n-bar e =0.4-0.5x10 19 m -3 , P LH =100-200KW). Design and experimental results are presented in the paper, which including control model analysis, configurations of control system and MFT feedback control experiments in HT-7. The high voltage power supply (HVPS) of LHCD is the main controller that regulates the LHCD power into the plasma to control the MFT

  19. Experimental and theoretical investigation of Fe-catalysis phenomenon in hydrogen thermal desorption form hydrocarbon plasma-discharge films from T-10 tokamak

    International Nuclear Information System (INIS)

    Stankevich, V.G.; Svechnikov, N.Y.; Lebedev, A.M.; Menshikov, K.A.; Kolbasov, B.N.; Sukhanov, L.P.

    2017-01-01

    A comprehensive study of hydrocarbon films obtained in the plasma discharge of large fusion facilities will allow the minimization of parasitic capture. The investigation of the effect of Fe impurities on D 2 thermal desorption (TD) from homogeneous CD x films (x ∼ 0.5) formed in the D-plasma discharge of the T-10 tokamak were carried out. The experimental TD spectra of the films showed 2 groups of peaks at 650-850 K and 900-1000 K for 2 adsorption states. The main result of the iron catalysis effect consists in the shift of the high-temperature peak by -24 K and in the increase in the fraction of the weakly bonded adsorption states. To describe the effect of iron impurities on TD of hydrogen isotopes, a structural cluster model based on the interaction of the Fe + ion with the 1,3-C 6 H 8 molecule was proposed. The potential energy surfaces of chemical reactions with the H 2 elimination were calculated using ab initio methods of quantum chemistry. It was established that the activation barrier of hydrogen TD is reduced by about 1 eV due to the interaction of the Fe + ion with the π-subsystem of the 1,3-C 6 H 8 molecule leading to a redistribution of the double bonds along the carbon system

  20. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    Zacek, F.; Badalec, J.; Jakubka, J.; Kryska, L.; Preinhaelter, J.; Stoeckel, J.; Valovic, M.; Nanobashvili, S.; Weixelbaum, L.; Wenzel, U.; Spineanu, F.; Vlad, M.

    1990-10-01

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

  1. The Effect of Feedback by SMS-text messages and email on Household Electricity Consumption: Experimental Evidence

    DEFF Research Database (Denmark)

    Larsen, Anders

    2010-01-01

    This paper analyzes the effect of supplying online feedback by SMS-text messages and email about electricity consumption on the level of total household electricity consumption. An experiment was conducted in which 1,452 households were randomly allocated to three experimental groups and two...

  2. Analytical model spectrum for electrostatic turbulence in tokamaks

    International Nuclear Information System (INIS)

    Fiedler-Ferrari, N.; Misguich, J.H.

    1990-04-01

    In this work we present an analytical model spectrum, for three-dimensional electrostatic turbulence (homogeneous, stationary and locally isotropic in the plane perpendicular to the magnetic field), constructed by using experimental results from TFR and TEXT Tokamaks, and satisfying basic symmetry and parity conditions. The proposed spectrum seems to be tractable for explicit analytical calculations of transport processes, and consistent with experimental data. Additional experimental measurements in the bulk plasma remain however necessary in order to determine some unknown spectral properties of parallel propagation

  3. Reading to learn experimental practice: The role of text and firsthand experience in the acquisition of an abstract science principle

    Science.gov (United States)

    Richmond, Erica Kesin

    2008-10-01

    From the onset of schooling, texts are used as important educational tools. In the primary years, they are integral to learning how to decode and develop fluency. In the later elementary years, they are often essential to the acquisition of academic content. Unfortunately, many children experience difficulties with this process, which is due in large part to their unfamiliarity with the genre of academic texts. The articles presented in this dissertation share an underlying theme of how to develop children's ability to comprehend and learn from academic, and specifically, non-narrative texts. The first article reviews research on the development of non-narrative discourse to elucidate the linguistic precursors to non-narrative text comprehension. The second and third articles draw from an empirical study that investigated the best way to integrate text, manipulation, and first-hand experience for children's acquisition and application of an abstract scientific principle. The scientific principle introduced in the study was the Control of Variables Strategy (CVS), a fundamental idea underlying scientific reasoning and a strategy for designing unconfounded experiments. Eight grade 4 classes participated in the study (N = 129), in one of three conditions: (a) read procedural text and manipulate experimental materials, (b) listen to procedural text and manipulate experimental materials, or (c) read procedural text with no opportunity to manipulate experimental materials. Findings from the study indicate that children who had the opportunity to read and manipulate materials were most effective at applying the strategy to designing and justifying unconfounded experiments, and evaluating written and physical experimental designs; however, there was no effect of instructional condition on a written assessment of evaluating familiar and unfamiliar experimental designs one week after the intervention. These results suggest that the acquisition and application of an abstract

  4. Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

    International Nuclear Information System (INIS)

    Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P

    2004-01-01

    Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values

  5. Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

    2004-09-01

    Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

  6. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

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

  7. 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

  8. Collisional-radiative models for hydrogen-like and helium-like carbon and oxygen ions and applications to experimental data from the TS Tokamak and the reversed field pinch RFX

    Energy Technology Data Exchange (ETDEWEB)

    Carraro, L.; Sattin, F.; Puiatty, M.E.; Scarin, P.; Valisa, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Mattioli, M.; Demichelis, C.; Mandl, W. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Hogan, J.T. [Oak Ridge National Lab., TN (United States)

    1996-07-01

    Collisional radiative models (CRM) are needed to simulate experimental line brightnesses and emissivities from fusion devices. CRM are built for H-like and He-like carbon and oxygen ions. The impurity ion radial distribution is obtained using a transport code with two radius dependent transport parameters: a diffusion coefficient D and an inward convection velocity V. Examples are given of the quantitative interpretation of experimental spectroscopic data from two fusion devices: the Tore Supra Tokamak and the Reversed Field Pinch RFX. (K.A.). 60 refs.

  9. Collisional-radiative models for hydrogen-like and helium-like carbon and oxygen ions and applications to experimental data from the TS Tokamak and the reversed field pinch RFX

    International Nuclear Information System (INIS)

    Carraro, L.; Sattin, F.; Puiatty, M.E.; Scarin, P.; Valisa, M.; Mattioli, M.; Demichelis, C.; Mandl, W.

    1996-07-01

    Collisional radiative models (CRM) are needed to simulate experimental line brightnesses and emissivities from fusion devices. CRM are built for H-like and He-like carbon and oxygen ions. The impurity ion radial distribution is obtained using a transport code with two radius dependent transport parameters: a diffusion coefficient D and an inward convection velocity V. Examples are given of the quantitative interpretation of experimental spectroscopic data from two fusion devices: the Tore Supra Tokamak and the Reversed Field Pinch RFX. (K.A.)

  10. Baking system for ports of experimental advanced super-conducting tokamak vacuum vessel and thermal stress analysis

    International Nuclear Information System (INIS)

    Cheng Yali; Bao Liman; Song Yuntao; Yao Damao

    2006-01-01

    The baking system of Experimental Advanced Super-Conducting Toakamk (EAST) vacuum vessel is necessary to obtain the baking temperature of 150 degree C. In order to define suitable alloy heaters and achieve their reasonable layouts, thermal analysis was carried out with ANSYS code. The analysis results indicate that the temperature distribution and thermal stress of most parts of EAST vacuum vessel ports are uniform, satisfied for the requirement, and are safe based on ASME criterion. Feasible idea on reducing the stress focus is also considered. (authors)

  11. Summary of experimental core turbulence characteristics in ohmic and electron cyclotron resonance heated discharges in T-10 tokamak plasmas

    International Nuclear Information System (INIS)

    Vershkov, V.A.; Shelukhin, D.A.; Soldatov, S.V.; Urazbaev, A.O.; Grashin, S.A.; Eliseev, L.G.; Melnikov, A.V.

    2005-01-01

    This report summarizes the results of experimental turbulence investigations carried out at T-10 for more than 10 years. The turbulence characteristics were investigated using correlation reflectometry, multipin Langmuir probe (MLP) and heavy ion beam probe diagnostics. The reflectometry capabilities were analysed using 2D full-wave simulations and verified by direct comparison using a MLP. The ohmic and electron cyclotron resonance heated discharges show the distinct transition from the core turbulence, having complex spectral structure, to the unstructured one in the scrape-off layer. The core turbulence includes 'broad band, quasi-coherent' features, arising due to the excitation of rational surfaces with high poloidal m-numbers, with a low frequency near zero and specific oscillations at 15-30 kHz. All experimentally measured properties of low frequency and high frequency quasi-coherent oscillations are in good agreement with predictions of linear theory for the ion temperature gradient/dissipative trapped electron mode instabilities. Significant local changes in the turbulence characteristics were observed at the edge velocity shear layer and in the core near q = 1 radius after switching off the electron cyclotron resonance heating (ECRH). The local decrease in the electron heat conductivity and decrease in the turbulence level could be evidence of the formation of an electron internal transport barrier. The dynamic behaviour of the core turbulence was also investigated for the case of fast edge cooling and the beginning phase of ECRH

  12. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

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

  13. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1985-01-01

    The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

  14. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1986-01-01

    Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

  15. Impact of Metadata on Full-text Information Retrieval Performance: An Experimental Research on a Small Scale Turkish Corpus

    Directory of Open Access Journals (Sweden)

    Çağdaş Çapkın

    2016-12-01

    Full Text Available Information institutions use text-based information retrieval systems to store, index and retrieve metadata, full-text, or both metadata and full-text (hybrid contents. The aim of this research was to evaluate impact of these contents on information retrieval performance. For this purpose, metadata (MIR, full-text (FIR and hybrid (HIR content information retrieval systems were developed with default Lucene information retrieval model for a small scale Turkish corpus. In order to evaluate performance of this three systems, “precision - recall” and “normalized recall” tests were conducted. Experimental findings showed that there were no significant differences between MIR and FIR in mean average precision (MAP performance. On the other hand, MAP performance of HIR was significantly higher in comparison to MIR and FIR. When information retrieval performance was evaluated as user-centered, the “normalized recall” performances of MIR and HIR were significantly higher than FIR. Additionally, there were no significant differences between the systems in retrieved relevant document means. Processing different types of contents such as metadata and full-text had some advantages and disadvantages for information retrieval systems in terms of term management. The advantages brought together in hybrid content processing (HIR and information retrieval performance improved.

  16. Disruptions in Tokamaks

    International Nuclear Information System (INIS)

    Bondeson, A.

    1987-01-01

    This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)

  17. ITER tokamak device

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  18. 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

  19. Integral torque balance in tokamaks

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2011-01-01

    The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.

  20. An enhanced tokamak startup model

    Science.gov (United States)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

    The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

  1. Role of impurity dynamics in resistivity-gradient-driven turbulence and tokamak edge plasma phenomena

    International Nuclear Information System (INIS)

    Hahm, T.S.; Diamond, P.H.; Terry, P.W.; Garcia, L.; Carreras, B.A.

    1986-03-01

    The role of impurity dynamics in resistivity gradient driven turbulence is investigated in the context of modeling tokamak edge plasma phenomena. The effects of impurity concentration fluctuations and gradients on the linear behavior of rippling instabilities and on the nonlinear evolution and saturation of resistivity gradient driven turbulence are studied both analytically and computationally. At saturation, fluctuation levels and particle and thermal diffusivities are calculated. In particular, the mean-square turbulent radial velocity is given by 2 > = (E 0 L/sub s/B/sub z/) 2 (L/sub/eta/ -1 + L/sub z -1 ) 2 . Thus, edged peaked impurity concentrations tend to enhance the turbulence, while axially peaked concentrations tend to quench it. The theoretical predictions are in semi-quantitative agreement with experimental results from the TEXT, Caltech, and Tosca tokamaks. Finally, a theory of the density clamp observed during CO-NBI on the ISX-B tokamak is proposed

  2. 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

  3. Empirical scaling for present Ohmically heated tokamaks

    International Nuclear Information System (INIS)

    Daughney, C.

    1975-01-01

    Experimental results from the Adiabatic Toroidal Compressor (ATC) tokamak are used to obtain empirical scaling laws for the average electron temperature and electron energy confinement time as functions of the average electron density, the effective ion charge, and the plasma current. These scaling laws are extended to include dependence upon minor and major plasma radius and toroidal field strength through a comparison of the various tokamaks described in the literature. Electron thermal conductivity is the dominant loss process for the ATC tokamak. The parametric dependences of the observed electron thermal conductivity are not explained by present theoretical considerations. The electron temperature obtained with Ohmic heating is shown to be a function of current density - which will not be increased in the next generation of large tokamaks. However, the temperature dependence of the electron energy confinement time suggests that significant improvement in confinement time will be obtained with supplementary electron heating. (author)

  4. Equilibrium Reconstruction in EAST Tokamak

    International Nuclear Information System (INIS)

    Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)

  5. Gyrosheath near the tokamak edge

    International Nuclear Information System (INIS)

    Hazeltine, R.D.; Xiao, H.; Valanju, P.M.

    1993-03-01

    A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results

  6. A Pilot Study of Biomedical Text Comprehension using an Attention-Based Deep Neural Reader: Design and Experimental Analysis.

    Science.gov (United States)

    Kim, Seongsoon; Park, Donghyeon; Choi, Yonghwa; Lee, Kyubum; Kim, Byounggun; Jeon, Minji; Kim, Jihye; Tan, Aik Choon; Kang, Jaewoo

    2018-01-05

    With the development of artificial intelligence (AI) technology centered on deep-learning, the computer has evolved to a point where it can read a given text and answer a question based on the context of the text. Such a specific task is known as the task of machine comprehension. Existing machine comprehension tasks mostly use datasets of general texts, such as news articles or elementary school-level storybooks. However, no attempt has been made to determine whether an up-to-date deep learning-based machine comprehension model can also process scientific literature containing expert-level knowledge, especially in the biomedical domain. This study aims to investigate whether a machine comprehension model can process biomedical articles as well as general texts. Since there is no dataset for the biomedical literature comprehension task, our work includes generating a large-scale question answering dataset using PubMed and manually evaluating the generated dataset. We present an attention-based deep neural model tailored to the biomedical domain. To further enhance the performance of our model, we used a pretrained word vector and biomedical entity type embedding. We also developed an ensemble method of combining the results of several independent models to reduce the variance of the answers from the models. The experimental results showed that our proposed deep neural network model outperformed the baseline model by more than 7% on the new dataset. We also evaluated human performance on the new dataset. The human evaluation result showed that our deep neural model outperformed humans in comprehension by 22% on average. In this work, we introduced a new task of machine comprehension in the biomedical domain using a deep neural model. Since there was no large-scale dataset for training deep neural models in the biomedical domain, we created the new cloze-style datasets Biomedical Knowledge Comprehension Title (BMKC_T) and Biomedical Knowledge Comprehension Last

  7. Anomalous transport in tokamaks

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1989-01-01

    A review is presented of what is known about anomalous transport in tokamaks. It is generally thought that this anomalous transport is the result of fluctuations in various plasma parameters. In the plasma edge detailed measurements of the quantities required to directly determine the fluctuation driven fluxes are available. The total flux of particles is well explained by the measured electrostatic fluctuation driven flux. However, a satisfactory model to explain the origin of the fluctuations has not been identified. The processes responsible for determining the edge energy flux are less clear, but electrostatic convection plays an important part. In the confinement region experimental observations are presently restricted to measurements of density and potential fluctuations and their correlations. The characteristics of the measured fluctuations are discussed and compared with the predictions of various models. Comparisons between measured particle, electron heat and ion heat fluxes, and those fluxes predicted to result from the measured fluctuations, are made. Magnetic fluctuations is discussed

  8. Electric conductivity and bootstrap current in tokamak

    International Nuclear Information System (INIS)

    Mao Jianshan; Wang Maoquan

    1996-12-01

    A modified Ohm's law for the electric conductivity calculation is presented, where the modified ohmic current can be compensated by the bootstrap current. A comparison of TEXT tokamak experiment with the theories shows that the modified Ohm's law is a more close approximation to the tokamak experiments than the classical and neoclassical theories and can not lead to the absurd result of Z eff <1, and the extended neoclassical theory would be not necessary. (3 figs.)

  9. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Baker, C.C.; Roberts, M.

    1976-01-01

    Three designs of an EPR are briefly described. The plasma physics involved in the EPR design is discussed. Brief discussions on the following topics are included: (1) MHD equilibrium, (2) impurity control, (3) transport scaling, (4) burn cycle dynamics, (5) first wall system, (6) blanket/shield systems, (7) plasma heating, (8) toroidal field coils, (9) poloidal field coils, (10) energy storage and transfer, (11) tritium fuel cycle, (12) radiation damage, and (13) costs

  10. Tokamak rotation and charge exchange

    International Nuclear Information System (INIS)

    Hazeltine, R.D.; Rowan, W.L.; Solano, E.R.; Valanju, P.M.

    1991-01-01

    In the absence of momentum input, tokamak toroidal rotation rates are typically small - no larger in particular than poloidal rotation - even when the radial electric field is strong, as near the plasma edge. This circumstance, contradicting conventional neoclassical theory, is commonly attributed to the rotation damping effect of charge exchange, although a detailed comparison between charge-exchange damping theory and experiment is apparently unavailable. Such a comparison is attempted here in the context of recent TEXT experiments, which compare rotation rates, both poloidal and toroidal, in helium and hydrogen discharges. The helium discharges provide useful data because they are nearly free of ion-neutral charge exchange; they have been found to rotate toroidally in reasonable agreement with neoclassical predictions. The hydrogen experiments show much smaller toroidal motion as usual. The theoretical calculation uses the full charge-exchange operator and assumes plateau collisionality, roughly consistent with the experimental conditions. The authors calculate the ion flow as a function of v cx /v c , where v cx is the charge exchange rate and v c the Coulomb collision frequency. The results are in reasonable accord with the observations. 1 ref

  11. Usability evaluation of an experimental text summarization system and three search engines: implications for the reengineering of health care interfaces.

    Science.gov (United States)

    Kushniruk, Andre W; Kan, Min-Yem; McKeown, Kathleen; Klavans, Judith; Jordan, Desmond; LaFlamme, Mark; Patel, Vimia L

    2002-01-01

    This paper describes the comparative evaluation of an experimental automated text summarization system, Centrifuser and three conventional search engines - Google, Yahoo and About.com. Centrifuser provides information to patients and families relevant to their questions about specific health conditions. It then produces a multidocument summary of articles retrieved by a standard search engine, tailored to the user's question. Subjects, consisting of friends or family of hospitalized patients, were asked to "think aloud" as they interacted with the four systems. The evaluation involved audio- and video recording of subject interactions with the interfaces in situ at a hospital. Results of the evaluation show that subjects found Centrifuser's summarization capability useful and easy to understand. In comparing Centrifuser to the three search engines, subjects' ratings varied; however, specific interface features were deemed useful across interfaces. We conclude with a discussion of the implications for engineering Web-based retrieval systems.

  12. 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.

  13. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    1978-08-01

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

  14. Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

    Directory of Open Access Journals (Sweden)

    Yutthapong Pinanroj

    2014-04-01

    Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

  15. Design and construction of electronic components for a ''Novillo'' Tokamak

    International Nuclear Information System (INIS)

    Lopez C, R.

    1986-07-01

    The goal of this effort was to design, construct and make functional the electronic components for a ''Novillo'' Tokamak currently being experimentally investigated at the National Institute of Nuclear Research in Mexico. The problem was to develop programmable electronic switches capable of discharging high voltage kilowatt energies stored in capacitator banks onto the coils of the Tokamak. (author)

  16. The physics of magnetic confinement configurations : Tokamak theory and experiment

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1982-01-01

    Several aspects, both theoretical and experimental, in plasma physics are discussed. The problem of magnetic confinement in Tokamak devices is treated. A discussion on the history of the development and on the future problems to be solved in Tokamaks is made. (L.C.) [pt

  17. 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

  18. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  19. 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)

  20. 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

  1. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-05-01

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

  2. Plasma turbulence in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)

  3. A midsize tokamak as a fast track to burning plasmas

    Directory of Open Access Journals (Sweden)

    E. Mazzucato

    2011-03-01

    Full Text Available This paper describes the conceptual design of a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥ 10 with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER. This can be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a different magnetic divertor from those currently employed in present experiments is discussed.

  4. Theoretical and experimental studies of a planar inductive coupled rf plasma source as the driver in simulator facility (ISTAPHM) of interactions of waves with the edge plasma on tokamaks

    Science.gov (United States)

    Ghanei, V.; Nasrabadi, M. N.; Chin, O.-H.; Jayapalan, K. K.

    2017-11-01

    This research aims to design and build a planar inductive coupled RF plasma source device which is the driver of the simulator project (ISTAPHM) of the interactions between ICRF Antenna and Plasma on tokamak by using the AMPICP model. For this purpose, a theoretical derivation of the distribution of the RF magnetic field in the plasma-filled reactor chamber is presented. An experimental investigation of the field distributions is described and Langmuir measurements are developed numerically. A comparison of theory and experiment provides an evaluation of plasma parameters in the planar ICP reactor. The objective of this study is to characterize the plasma produced by the source alone. We present the results of the first analysis of the plasma characteristics (plasma density, electron temperature, electron-ion collision frequency, particle fluxes and their velocities, stochastic frequency, skin depth and electron energy distribution functions) as function of the operating parameters (injected power, neutral pressure and magnetic field) as measured with fixed and movable Langmuir probes. The plasma is currently produced only by the planar ICP. The exact goal of these experiments is that the produced plasma by external source can exist as a plasma representative of the edge of tokamaks.

  5. 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

  6. Comparison of experimentally-inferred ion thermal diffusivities with neoclassical theory for neutral beam-heated discharges in the Doublet III tokamak

    International Nuclear Information System (INIS)

    Groebner, R.J.

    1986-04-01

    The study of ion transport in neutral beam-heated discharges in tokamaks is necessary to determine if neoclassical theory can reliably be used to predict the performance of future machines. Previous studies of ion tranport have generally been difficult due to the lack of information regarding the ion temperature profile. The standard procedure used to study ion transport has been to model the T/sub i/ profile with the assumption that the ion thermal diffusivity profile chi/sub i/(r) was equal to a multiplier times chi/sub i//sup neo/(r), the ion thermal diffusivity calculated from neoclassical theory. The multiplier was varied until the calculated T/sub i/ profile agreed with the available ion temperature data, usually T/sub i/(0) or the measured neutron rate. Values of the multiplier in the range of 1 to 10 have generally been obtained with few estimates of the uncertainties in these values. Furthermore, there have been few, if any, attempts to calculate chi/sub i/ by modeling the ion temperature profiles in other ways. As a result, the issue as to whether or not the ion transport in tokamaks is in agreement with neoclassical theory has not been definitively answered

  7. Design parameters of Tokamak-7 system

    International Nuclear Information System (INIS)

    Ivanov, D.P.; Keilin, V.E.; Klimenko, E.Yu.; Strelkov, V.S.

    Superconducting windings for the main magnetic field of Tokamak-7 are discussed. The parameters of this facility are based on the use of commercially available superconducting materials for fields up to 80 kOe. Experimental parameters are described. (U.S.)

  8. 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

  9. 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

  10. 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.

  11. Tokamak Engineering Technology Facility scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

    1976-03-01

    A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.

  12. Thermonuclear ignition in the next generation tokamaks

    International Nuclear Information System (INIS)

    Johner, J.

    1989-04-01

    The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aB t x of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition

  13. Tokamak Engineering Technology Facility scoping study

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

    1976-03-01

    A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR

  14. Can better modelling improve tokamak control?

    International Nuclear Information System (INIS)

    Lister, J.B.; Vyas, P.; Ward, D.J.; Albanese, R.; Ambrosino, G.; Ariola, M.; Villone, F.; Coutlis, A.; Limebeer, D.J.N.; Wainwright, J.P.

    1997-01-01

    The control of present day tokamaks usually relies upon primitive modelling and TCV is used to illustrate this. A counter example is provided by the successful implementation of high order SISO controllers on COMPASS-D. Suitable models of tokamaks are required to exploit the potential of modern control techniques. A physics based MIMO model of TCV is presented and validated with experimental closed loop responses. A system identified open loop model is also presented. An enhanced controller based on these models is designed and the performance improvements discussed. (author) 5 figs., 9 refs

  15. 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

  16. Effect of impurity radiation on tokamak equilibrium

    International Nuclear Information System (INIS)

    Rebut, P.H.; Green, B.J.

    1977-01-01

    The energy loss from a tokamak plasma due to the radiation from impurities is of great importance in the overall energy balance. Taking the temperature dependence of this loss for two impurities characteristic of those present in existing tokamak plasmas, the condition for radial power balance is derived. For the impurities considered (oxygen and iron) it is found that the radiation losses are concentrated in a thin outer layer of the plasma and the equilibrium condition places an upper limit on the plasma paraticle number density in this region. This limiting density scales with mean current density in the same manner as is experimentally observed for the peak number density of tokamak plasmas. The stability of such equilibria is also discussed. (author)

  17. Ripple induced trapped particle loss in tokamaks

    International Nuclear Information System (INIS)

    White, R.B.

    1996-05-01

    The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetric orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor and International Thermonuclear Experimental Reactor equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks

  18. 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]...

  19. Tokamak simulation code manual

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. Injection of pellets into the TCA tokamak

    International Nuclear Information System (INIS)

    Martin, Y.

    1993-05-01

    This thesis presents experimental results from the analysis of the ablation process of pellets injected into the TCA tokamak. The determination of scaling laws relating the pellet penetration to the pellet and plasma parameters preceding injection, were used to improve the understanding of the interaction of the pellet with the plasma since a) the pellet and plasma conditions preceding injection were varied over a large range, and b) the estimation of the penetration depth takes into account the influence of striations in the deposition profile. Over 400 pellets with a range of sizes and speeds were injected into a range of plasma parameters in order to create a database from which the scaling laws could be deduced. The ablation characteristics were principally measured with two CCD video cameras, which provided good spatial resolution, and two filtered photomultiplier tubes, which provided good temporal resolution of the light emitted from the pellet ablation cloud. In the text, the traditional methods of analysing these diagnostics are examined with special reference to the presumptions that a) the pellet velocity is constant in the plasma, and b) the light intensity determined from the ablation cloud is proportional to the ablation rate. After successive data reduction from the database, in order to separate the effects of varying different parameters, the main observations were that, a) the pellet penetration varies as the square root of the pellet velocity, b) the scaling laws for the other parameters strongly depend on whether the pellet has sufficient velocity to reach the q=1 rational magnetic surface in the tokamak. (author) 45 refs

  1. Axisymmetric instability in a noncircular tokamak

    International Nuclear Information System (INIS)

    Lipschultz, B.

    1979-10-01

    The stability of dee, inverse-dee and square cross section plasmas to axisymmetric modes has been investigated experimentally in Tokapole II, a tokamak with a four-null poloidal divertor. Experimental results are closely compared with predictions of two numerical stability codes - the PEST code (ideal MHD, linear stability) adapted to tokapole geometry and a code which follows the nonlinear evolution of shapes similar to tokapole equilibria

  2. Interlock System for the COMPASS Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Adamek, J.; Pisacka, J.; Panek, R. [Institute of Plasma Physics AS CR, Association EURATOM/IPP.CR, Prague (Czech Republic); Sova, J.; Siba, J.; Kovarj, J. [Department of Control Engineering, Faculty of Electrical Engineering, Prague (Czech Republic)

    2009-07-01

    Full text of publication follows: The COMPASS tokamak (R=0.56 m, a=0.18 - 0.23 m) is starting operation presently at Institute of Plasma Physics AS CR in Prague. An important issue of the operation is the safety of the personnel and machine protection against faults, presented in this contribution. The personnel protection is based on a restricted access into the experimental hall during the operation of potentially dangerous systems. A tokamak hall access system, based on Honeywell WIN-PAK (tm) 2005, allows to set the status of the experimental area (open/closed) and to control the in and out movement of persons using access cards. On top of this, a check of the whole experimental area by the operator is enforced before the hall enclosure. A hardware interlock then interprets the experimental hall status and controls the operation of key systems accordingly. The permit for operation is granted and the real status of the systems is reported by hard wired potential less contacts. The control procedure is based on a PLC MicroPEL M66. This PLC provides HW interface between Actuators (Relays and switches) and it is connected on PESNET bus. Its programming is done using language Simple v.2 in Winstudio IDE. Second site of personnel protection system is created on PC where runs a .NET application on MSWindows XP or 2000. This PC is connected with PLC via PESNET bus (on RS485 layer) and it generates all control signals to PLC from the operator. Simultaneously, the PC receives all warning and alarm signals from the PLC. This signals are displayed on a screen of the PC in real-time, this way the GUI provides visualization of the controlled process. Except for this fact the operator is informed about the status of the system and individual subsystems on a PC via an operator's panel. Further we will describe the machine protection which uses similar system for checking conditions for the start of a shot. Fast key processes which have to be checked during the shot are

  3. Digital control of plasma position in Damavand tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.

    2002-03-01

    Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)

  4. Nonneutralized charge effects on tokamak edge magnetohydrodynamic stability

    International Nuclear Information System (INIS)

    Zheng, Linjin; Horton, W.; Miura, H.; Shi, T.H.; Wang, H.Q.

    2016-01-01

    Owing to the large ion orbits, excessive electrons can accumulate at tokamak edge. We find that the nonneutralized electrons at tokamak edge can contribute an electric compressive stress in the direction parallel to magnetic field by their mutual repulsive force. By extending the Chew–Goldburger–Low theory (Chew et al., 1956 [13]), it is shown that this newly recognized compressive stress can significantly change the plasma average magnetic well, so that a stabilization of magnetohydrodynamic modes in the pedestal can result. This linear stability regime helps to explain why in certain parameter regimes the tokamak high confinement can be rather quiet as observed experimentally.

  5. 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

  6. Online self-expression and experimentation as 'reflectivism': Using text analytics to examine the participatory forum Hello Sunday Morning.

    Science.gov (United States)

    Carah, Nicholas; Meurk, Carla; Angus, Daniel

    2017-03-01

    Hello Sunday Morning is an online health promotion organisation that began in 2009. Hello Sunday Morning asks participants to stop consuming alcohol for a period of time, set a goal and document their progress on a personal blog. Hello Sunday Morning is a unique health intervention for three interrelated reasons: (1) it was generated outside a clinical setting, (2) it uses new media technologies to create structured forms of participation in an iterative and open-ended way and (3) participants generate a written record of their progress along with demographic, behavioural and engagement data. This article presents a text analysis of the blog posts of Hello Sunday Morning participants using the software program Leximancer. Analysis of blogs illustrates how participants' expressions change over time. In the first month, participants tended to set goals, describe their current drinking practices in individual and cultural terms, express hopes and anxieties and report on early efforts to change. After month 1, participants continued to report on efforts to change and associated challenges and reflect on their place as individuals in a drinking culture. In addition to this, participants evaluated their efforts to change and presented their 'findings' and 'theorised' them to provide advice for others. We contextualise this text analysis with respect to Hello Sunday Morning's development of more structured forms of online participation. We offer a critical appraisal of the value of text analytics in the development of online health interventions.

  7. 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)

  8. 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

  9. Turbulence in tokamak plasmas. Effect of a radial electric field shear; Turbulence dans les plasmas de tokamaks. Effet d`un cisaillement de champ electrique radial

    Energy Technology Data Exchange (ETDEWEB)

    Payan, J

    1994-05-01

    After a review of turbulence and transport phenomena in tokamak plasmas and the radial electric field shear effect in various tokamaks, experimental measurements obtained at Tore Supra by the means of the ALTAIR plasma diagnostic technique, are presented. Electronic drift waves destabilization mechanisms, which are the main features that could describe the experimentally observed microturbulence, are then examined. The effect of a radial electric field shear on electronic drift waves is then introduced, and results with ohmic heating are studied together with relations between turbulence and transport. The possible existence of ionic waves is rejected, and a spectral frequency modelization is presented, based on the existence of an electric field sheared radial profile. The position of the inversion point of this field is calculated for different values of the mean density and the plasma current, and the modelization is applied to the TEXT tokamak. The radial electric field at Tore Supra is then estimated. The effect of the ergodic divertor on turbulence and abnormal transport is then described and the density fluctuation radial profile in presence of the ergodic divertor is modelled. 80 figs., 120 refs.

  10. Standardised (plain) cigarette packaging increases attention to both text-based and graphical health warnings: experimental evidence.

    Science.gov (United States)

    Shankleman, M; Sykes, C; Mandeville, K L; Di Costa, S; Yarrow, K

    2015-01-01

    To investigate whether standardised cigarette packaging increases the time spent looking at health warnings, regardless of the format of those warnings. A factorial (two pack styles x three warning types) within-subject experiment, with participants randomised to different orders of conditions, completed at a university in London, UK. Mock-ups of cigarette packets were presented to participants with their branded portion in either standardised (plain) or manufacturer-designed (branded) format. Health warnings were present on all packets, representing all three types currently in use in the UK: black & white text, colour text, or colour images with accompanying text. Gaze position was recorded using a specialised eye tracker, providing the main outcome measure, which was the mean proportion of a five-second viewing period spent gazing at the warning-label region of the packet. An opportunity sample of 30 (six male, mean age = 23) young adults met the following inclusion criteria: 1) not currently a smoker; 2) 50% viewing time. These participants spent a greater proportion of the available time gazing at the warning-label region when the branded section of the pack was standardised (following current Australian guidelines) rather than containing the manufacturer's preferred design (mean difference in proportions = 0.078, 95% confidence interval 0.049 to 0.106, p interaction p = 0.295). During incidental viewing of cigarette packets, young adult never-smokers are likely to spend more time looking at health warnings if manufacturers are compelled to use standardised packaging, regardless of the warning design. Copyright © 2014 The Authors. Published by Elsevier Ltd.. All rights reserved.

  11. The scientific program of the Tokamak de Varennes

    International Nuclear Information System (INIS)

    Daughney, C.C.

    1989-01-01

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

  12. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

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

    1985-12-01

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

  13. 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

  14. 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

  15. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-05-01

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime

  16. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    International Nuclear Information System (INIS)

    Azizov, E A

    2012-01-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined. (conferences and symposia)

  17. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

  18. Overview of Tokamak Results

    International Nuclear Information System (INIS)

    Unterberg, Bernhard; Samm, Ulrich

    2004-01-01

    An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas

  19. Analytic description of tokamak equilibrium sustained by high fraction bootstrap current

    International Nuclear Information System (INIS)

    Shi Bingren

    2002-01-01

    Recently, to save the current drive power and to obtain more favorable confinement merit for tokamak reactor, large faction bootstrap current sustained equilibrium has attracted great interests both theoretically and experimentally. An powerful expanding technique and the tokamak ordering are used to expand the Grad-Shafranov equation to obtain a series of ordinary differential equations which allow for different sets of input parameters. The fully bootstrap current sustained tokamak equilibria are then solved analytically

  20. Scaling laws for TEXT plasma profiles

    International Nuclear Information System (INIS)

    McCool, S.C.; Bravenec, R.V.; Chen, J.Y.; Foster, M.S.; Li, W.L.; Ouroura, A.; Phillips, P.E.; Richards, B.; Wenzel, K.W.; Zhang, Z.M.

    1994-01-01

    Regression analysis has been performed on a number of measured profiles including temperature and density vs. nominal macroscopic operating parameters for TEXT tokamak (pre-upgrade) ohmic plasmas. The resulting simple empirical model has enabled the authors to quickly approximate profiles of electron temperature and density, ion temperature, and soft x-ray brightness, as well as the scalar quantities: total radiated power, q=1 radius, sawtooth period and amplitude, and energy confinement time as a power law of toroidal field, plasma current, chord average density, and fueling gas atomic weight. The model profiles are only applicable to the plasma interior, i.e. within the limiter radius. In most cases the predicted model profiles are within the experimental error bars of measured profiles and are more accurate at predicting profile variation for small operating parameter changes than the measured profiles

  1. High beta tokamaks

    International Nuclear Information System (INIS)

    Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.

    1978-01-01

    MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby

  2. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Pare, V.K.

    1983-01-01

    Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

  3. Research using small tokamaks

    International Nuclear Information System (INIS)

    1993-01-01

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

  4. 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

  5. The basics of spherical tokamaks and progress in European research

    International Nuclear Information System (INIS)

    Gusev, V K; Alladio, F; Morris, A W

    2003-01-01

    When the aspect ratio of a tokamak (A = R/a) decreases significantly, there is a transformation of the well studied tokamak toroidal magnetic configuration into the spherical tokamak (ST) configuration. This configuration has high natural plasma elongation and triangularity and other unique equilibrium and stability properties of ST configuration, which are discussed in this paper. European research into ST physics is well advanced in spite of the young age of this branch of fusion science. An overview of selected experimental and theoretical results obtained at Ioffe, Culham and Frascati is given with the emphasis on their complementarity and links to the main stream of tokamak research, such as ITER. An outline of the basic ST advantages and the potential of ST research for new insights into magnetic confinement is also given. More detailed descriptions of recent advances in ST theory and experiment may be found in the invited papers by Akers and Ono in the proceedings of this conference

  6. Turbulence in tokamak plasmas. Effect of a radial electric field shear

    International Nuclear Information System (INIS)

    Payan, J.

    1994-05-01

    After a review of turbulence and transport phenomena in tokamak plasmas and the radial electric field shear effect in various tokamaks, experimental measurements obtained at Tore Supra by the means of the ALTAIR plasma diagnostic technique, are presented. Electronic drift waves destabilization mechanisms, which are the main features that could describe the experimentally observed microturbulence, are then examined. The effect of a radial electric field shear on electronic drift waves is then introduced, and results with ohmic heating are studied together with relations between turbulence and transport. The possible existence of ionic waves is rejected, and a spectral frequency modelization is presented, based on the existence of an electric field sheared radial profile. The position of the inversion point of this field is calculated for different values of the mean density and the plasma current, and the modelization is applied to the TEXT tokamak. The radial electric field at Tore Supra is then estimated. The effect of the ergodic divertor on turbulence and abnormal transport is then described and the density fluctuation radial profile in presence of the ergodic divertor is modelled. 80 figs., 120 refs

  7. Electron cyclotron emission from the PLT tokamak

    International Nuclear Information System (INIS)

    Hosea, J.; Arunasalam, V.; Cano, R.

    1977-07-01

    Experimental measurements of electron cyclotron emission from the PLT tokamak plasma reveal that black-body emission occurs at the fundamental frequency. Such emission, not possible by direct thermal excitation of electromagnetic waves, is herein attributed to thermal excitation of electrostatic (Bernstein) waves which then mode convert into electromagnetic waves. The local feature of the electrostatic wave generation permits spatially and time resolved measurements of electron temperature as for the second harmonic emission

  8. Coherent structures in tokamak plasmas workshop: Proceedings

    International Nuclear Information System (INIS)

    Koniges, A.E.; Craddock, G.G.

    1992-08-01

    Coherent structures have the potential to impact a variety of theoretical and experimental aspects of tokamak plasma confinement. This includes the basic processes controlling plasma transport, propagation and efficiency of external mechanisms such as wave heating and the accuracy of plasma diagnostics. While the role of coherent structures in fluid dynamics is better understood, this is a new topic for consideration by plasma physicists. This informal workshop arose out of the need to identify the magnitude of structures in tokamaks and in doing so, to bring together for the first time the surprisingly large number of plasma researchers currently involved in work relating to coherent structures. The primary purpose of the workshop, in addition to the dissemination of information, was to develop formal and informal collaborations, set the stage for future formation of a coherent structures working group or focus area under the heading of the Tokamak Transport Task Force, and to evaluate the need for future workshops on coherent structures. The workshop was concentrated in four basic areas with a keynote talk in each area as well as 10 additional presentations. The issues of discussion in each of these areas was as follows: Theory - Develop a definition of structures and coherent as it applies to plasmas. Experiment - Review current experiments looking for structures in tokamaks, discuss experimental procedures for finding structures, discuss new experiments and techniques. Fluids - Determine how best to utilize the resource of information available from the fluids community both on the theoretical and experimental issues pertaining to coherent structures in plasmas. Computation - Discuss computational aspects of studying coherent structures in plasmas as they relate to both experimental detection and theoretical modeling

  9. Neutral-beam current drive in tokamaks

    International Nuclear Information System (INIS)

    Devoto, R.S.

    1986-01-01

    The theory of neutral-beam current drive in tokamaks is reviewed. Experiments are discussed where neutral beams have been used to drive current directly and also indirectly through neoclassical effects. Application of the theory to an experimental test reactor is described. It is shown that neutral beams formed from negative ions accelerated to 500 to 700 keV are needed for this device

  10. Neutral-beam current drive in tokamaks

    International Nuclear Information System (INIS)

    Devoto, R.S.

    1987-01-01

    The theory of neutral-beam current drive in tokamaks is reviewed. Experiments are discussed where neutral beams have been used to drive current directly and also indirectly through neoclassical effects. Application of the theory to an experimental test reactor is described. It is shown that neutral beams formed from negative ions accelerated to 500-700 keV are needed for this device

  11. Comparison of tokamak burn cycle options

    International Nuclear Information System (INIS)

    Ehst, D.A.; Brooks, J.N.; Cha, Y.; Evans, K. Jr.; Hassanein, A.M.; Kim, S.; Majumdar, S.; Misra, B.; Stevens, H.C.

    1985-01-01

    Experimental confirmation of noninductive current drive has spawned a number of suggestions as to how this technique can be used to extend the fusion burn period and improve the reactor prospects of tokamaks. Several distinct burn cycles, which employ various combinations of Ohmic and noninductive current generation, are possible, and we will study their relative costs and benefits for both a commerical reactor as well as an INTOR-class device. We begin with a review of the burn cycle options

  12. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

    Aymar, R.; Bareyt, B.; Bon Mardion, G.

    1980-10-01

    This document describes the basic design for the main components of the Tokamak system of Tora Supra. As such, it focuses on the engineering problems, and refers to last year report on Tora Supra (EUR-CEA-1021) for objectives and experimental programme of the apparatus on one hand, and for qualifying tests of the main technical solutions on the other hand. Superconducting toroidal field coil system, vacuum vessels and radiation shields, poloidal field system and cryogenic system are described

  13. Small-scale tearing mode in tokamaks

    International Nuclear Information System (INIS)

    Ivanov, N.V.

    1983-01-01

    Considerations are given on the possible effect of small-scale tearing mode with m >> 1 on the plasma electron thermal conductivity in a tokamak. The estimate of the electron thermal conductivity coefficient is obtained. Calculation results are compared with experimental data. The calculated dependence of radial distribution of electron temperature is shown to vary weakly with the tn(m 2 /m 1 ) alteration everywhere, except for the vicinity of point r approximately 0

  14. Tokamaks (Second Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Stott, Peter [JET, UK (United Kingdom)

    1998-10-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  15. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  16. Burn Control Mechanisms in Tokamaks

    Science.gov (United States)

    Hill, M. A.; Stacey, W. M.

    2015-11-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

  17. 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

  18. Electrical conductivity in tokamaks and extended neoclassical theory

    International Nuclear Information System (INIS)

    Segre, S.E.; Zanza, V.

    1992-01-01

    The electrical conductivity measurements reported from various tokamaks (D-III, PLT, TEXT, ASDEX, JT-60, TEXTOR, JET, TFTR) and compared with the usual neoclassical theory are here also compared with the extended neoclassical theory where the electron-electron collision rate is anomalous while the electron-ion collision rate remains Coulombian. It is found that, out of the 14 experiments considered, three are consistent with both the neoclassical and the extended neoclassical theories, four are consistent only with the extended neoclassical theory, and four are consistent with the neoclassical theory and also, within the experimental errors, not inconsistent with the extended neoclassical theory; the remaining three experiments appear to be incompatible with both theories. It is concluded that the extended neoclassical theory is in better agreement with conductivity experiments than the conventional neoclassical theory and, indeed, the extended theory is a serious candidate for explaining tokamak behaviour, since it accommodates naturally an anomalous electron thermal transport, which the conventional neoclassical theory is unable to do. (author). 31 refs, 1 fig

  19. A novel approach to linearization of the electromagnetic parameters of tokamaks with an iron core

    Energy Technology Data Exchange (ETDEWEB)

    Fu, P. E-mail: fupeng@mail.ipp.ac.cn; Liu, Z.Z.; Zou, J.H

    2002-05-01

    The equivalent model of an iron core tokamak is developed, in which the electromagnetic parameters of several pairs of coils in opposite series (PCOS) are not dependent on the saturation of the iron core during tokamak operation. With this the electromagnetic parameters of all the coils in an iron core tokamak can be linearized, As an example, the electromagnetic parameters of Hefei Super-conductive Tokamak with iron core (HT-7) are linearized, and it is in good agreement with the experimental results. The linearization approach can be applied in real time plasma control and electromagnetic analysis.

  20. 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

  1. The collaborative tokamak control room

    International Nuclear Information System (INIS)

    Schissel, D.P.

    2006-01-01

    Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in collaborations between experimental sites and laboratories worldwide. In the US, the National Fusion Collaboratory Project is developing a persistent infrastructure to enable scientific collaboration for all aspects of magnetic fusion energy research by creating a robust, user-friendly collaborative environment and deploying this to the more than 1000 US fusion scientists in 40 institutions who perform magnetic fusion research. This paper reports on one aspect of the project which is the development of the collaborative tokamak control room to enhance both collocated and remote scientific participation in experimental operations. This work includes secured computational services that can be scheduled as required, the ability to rapidly compare experimental data with simulation results, a means to easily share individual results with the group by moving application windows to a shared display, and the ability for remote scientists to be fully engaged in experimental operations through shared audio, video, and applications. The project is funded by the USDOE Office of Science, Scientific Discovery through Advanced Computing (SciDAC) Program and unites fusion and computer science researchers to directly address these challenges

  2. A transport model for Alcator scaling in tokamaks

    International Nuclear Information System (INIS)

    Ohkawa, T.

    1978-01-01

    A theoretical model is proposed to explain the tokamak energy confinement time. With no adjustable numerical coefficients, the model predicts experimentally observed values to within a level of uncertainty consistent with the intrinsic spread of the experimental data and the necessity of calculating the confinement time without precise knowledge of the temperature profile. (Auth.)

  3. Physics analysis of the Apollo D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Santarius, J.F.; Emmert, G.A.

    1990-01-01

    Recent developments in the analysis and conceptual design of Apollo, a D- 3 He Tokamak Reactor are presented. Encouraging experimental results on TEXT motivated a key change in the Apollo concept utilization of an ergodic magnetic limiter for impurity control instead of a divertor. Parameters for the updated Apollo design and an analysis of the ergoidc magnetic limiter are given. The Apollo reference case uses direct conversion of synchrotron radiation to electricity by rectifying antennas (rectennas) for its power conversion system. Previous analyses of this concept are expanded, including further details of the rectennas and of the loss of synchrotron power to the waveguides and walls. Although Apollo will burn D- 3 He fuel, a significant amount of unburned tritium will be generated by D4D reactions. The possibility of operating a short, dedicated, T+ 3 He burn phase to eliminate this tritium will be examined

  4. 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.

  5. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  6. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

    1981-08-01

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  7. Theory of tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.

    1989-01-01

    The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.

  8. Axisymmetric tokamak scapeoff transport

    International Nuclear Information System (INIS)

    Singer, C.E.; Langer, W.D.

    1982-08-01

    We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed

  9. The physics of tokamak start-up

    International Nuclear Information System (INIS)

    Mueller, D.

    2013-01-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current

  10. Microinstability theory in tokamaks: a review

    International Nuclear Information System (INIS)

    Tang, W.M.

    1977-06-01

    Significant investigations in the area of tokamak microinstability theory are reviewed. Emphasis is given to the work covering the period from 1970 through 1976. Special attention is focused on low-frequency electrostatic drift-type modes, which are generally believed to be the dominant tokamak microinstabilities under normal operating conditions. The basic linear formalism including electromagnetic (finite beta) modifications is presented along with a general survey of the numerous papers investigating specific linear and nonlinear effects on these modes. Estimates of the associated anomalous transport and confinement times are discussed, and a summary of relevant experimental results is given. Studies of the nonelectrostatic and high-frequency instabilities associated with the presence of high energy ions from neutral beam injection (or with the presence of alpha particles from fusion reactions) are also surveyed

  11. Comparison between stellarator and tokamak divertor transport

    International Nuclear Information System (INIS)

    Feng, Y.; Lunt, T.; Kobayashi, M.; Reiter, D.

    2010-11-01

    The paper compares the essential divertor transport features of the poloidal divertor, which is well-developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences for how the divertor functions. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off-layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modeling using EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations. (author)

  12. 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.

  13. 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)

  14. Microinstability theory in tokamaks: a review

    Energy Technology Data Exchange (ETDEWEB)

    Tang, W.M.

    1977-06-01

    Significant investigations in the area of tokamak microinstability theory are reviewed. Emphasis is given to the work covering the period from 1970 through 1976. Special attention is focused on low-frequency electrostatic drift-type modes, which are generally believed to be the dominant tokamak microinstabilities under normal operating conditions. The basic linear formalism including electromagnetic (finite beta) modifications is presented along with a general survey of the numerous papers investigating specific linear and nonlinear effects on these modes. Estimates of the associated anomalous transport and confinement times are discussed, and a summary of relevant experimental results is given. Studies of the nonelectrostatic and high-frequency instabilities associated with the presence of high energy ions from neutral beam injection (or with the presence of alpha particles from fusion reactions) are also surveyed.

  15. Models for impurity effects in tokamaks

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1980-03-01

    Models for impurity effects in tokamaks are described with an emphasis on the relationship between attainment of high β and impurity problems. We briefly describe the status of attempts to employ neutral beam heating to achieve high β in tokamaks and propose a qualitative model for the mechanism by which heavy metal impurities may be produced in the startup phase of the discharge. We then describe paradoxes in impurity diffusion theory and discuss possible resolutions in terms of the effects of large-scale islands and sawtooth oscillations. Finally, we examine the prospects for the Zakharov-Shafranov catastrophe (long time scale disintegration of FCT equilibria) in the context of present and near-term experimental capability

  16. Turbulence and abnormal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    1988-06-01

    The objective of this thesis is the study of plasma microinstabilities in linear and nonlinear tokamak regime. After a brief review of experimental results the theoretical tools used in this study are presented. A variational method founded on the existence of angular variables system and on action for charged particles in tokamak configurations is detailed. The correspondent functional extreme with regard to fluctuating electromagnetic field, is calculated analytically with taking into account the toroidal geometry. A numerical code, TORRID, has been constructed on this principle and the main instabilities, particularly ionic instabilities and microtearing, has been linearly studied. The most simple non linear methods are rewieved and applied at the microtearing instabilities. The quasilinear transport coefficients are deducted of an entropy minimum production principle. The ionic thermic conductivity and the viscosity are calculated for an ionic turbulence [fr

  17. Runaway-ripple interaction in Tokamaks

    International Nuclear Information System (INIS)

    Laurent, L.; Rax, J.M.

    1989-08-01

    Two approaches of the interaction between runaway electrons and the ripple field, in tokamaks, are discussed. The first approach considers the resonance effect as an intense cyclotron heating of the electrons, by the ripple field, in the guiding center frame of the fast particles. In the second approach, an Hamiltonian formalism is used. A criterion for the onset of chaotic behavior and the results are given. A new universal instability of the runaway population in tokamak configuration is found. When combined with cyclotron losses one of its major consequence is to act as an effective slowing down mechanism preventing the free fall acceleration toward the synchrotron limit. This configuration allows the explanation of some experimental results of Tore Supra and Textor

  18. Decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Walton, G.R.

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory (PPPL) will complete its experimental lifetime with a series of deuterium-tritium pulses in 1994. As a result, the machine structures will become radioactive, and vacuum components will also be contaminated with tritium. Dose rate levels will range from less than 1 mr/h for external structures to hundreds of mr/h for the vacuum vessel. Hence, decommissioning operations will range from hands on activities to the use of remotely operated equipment. After 21 months of cool down, decontamination and decommissioning (D and D) operations will commence and continue for approximately 15 months. The primary objective is to render the test cell complex re-usable for the next machine, the Tokamak Physics Experiment (TPX). This paper presents an overview of decommissioning TFTR and discusses the D and D objectives

  19. Density limits in Tokamaks

    International Nuclear Information System (INIS)

    Tendler, M.

    1984-06-01

    The energy loss from a tokamak plasma due to neutral hydrogen radiation and recycling is of great importance for the energy balance at the periphery. It is shown that the requirement for thermal equilibrium implies a constraint on the maximum attainable edge density. The relation to other density limits is discussed. The average plasma density is shown to be a strong function of the refuelling deposition profile. (author)

  20. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.

    1984-05-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

  1. Modular tokamak magnetic system

    International Nuclear Information System (INIS)

    Yang, T.F.

    1988-01-01

    This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil

  2. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.; California Univ., Los Angeles

    1984-01-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)

  3. TPX tokamak construction management

    International Nuclear Information System (INIS)

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-01-01

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly

  4. 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.

  5. Simulation of MHD instability effects on burning plasma transport with ITB in tokamak and helical reactors

    International Nuclear Information System (INIS)

    Yamazaki, K.; Yamada, I.; Taniguchi, S.; Oishi, T.

    2009-01-01

    Full text: The high performance plasma behavior is required to realize economic and environmental-friendly fusion reactors compatible with conventional power plant systems. To improve plasma confinement, the formation of internal transport barrier (ITB) is anticipated, and its behavior is analyzed by the simulation code TOTAL (Toroidal Transport Linkage Analysis). This TOTAL code comprises a 2- or 3-dimensional equilibrium and 1-dimensional predictive transport code for both tokamak and helical systems. In the tokamak code TOTAL-T, the external current drive, bootstrap current, sawtooth oscillation, ballooning mode and neoclassical tearing mode (NTM) analyses are included. The steady-state burning plasma operation is achieved by the feedback control of pellet injection fuelling and external heating power control. The impurity dynamics of iron and tungsten is also included in this code. The NTM effects are evaluated using the modified Rutherford Model with the stabilization of the ECCD current drive. The excitation of m=2/n=1 NTM leads to the 20 % reduction in the central temperature in ITER-like reactors. Recently, the external non-resonant helical field application is analyzed and its stabilization properties are evaluated. The pellet injection effects on ITB formation is also clarified in tokamak and helical plasmas. Relationship between sawtooth oscillation and impurity ejection is recently simulated in comparison with experimental data. In this conference, we will show above-stated new results on MHD instability effects on burning plasma transport. (author)

  6. Trajectory planning of tokamak flexible in-vessel inspection robot

    International Nuclear Information System (INIS)

    Wang, Hesheng; Chen, Weidong; Lai, Yinping; He, Tao

    2015-01-01

    Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

  7. Trajectory planning of tokamak flexible in-vessel inspection robot

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Hesheng [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Lai, Yinping; He, Tao [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China)

    2015-10-15

    Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

  8. 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.

  9. J-TEXT-EPICS: An EPICS toolkit attempted to improve productivity

    International Nuclear Information System (INIS)

    Zheng, Wei; Zhang, Ming; Zhang, Jing; Zhuang, Ge

    2013-01-01

    Highlights: • Tokamak control applications can be developed in very short period with J-TEXT-EPICS. • J-TEXT-EPICS enables users to build control applications with device-oriented functions. • J-TEXT-EPICS is fully compatible with EPICS Channel Access protocol. • J-TEXT-EPICS can be easily extended by plug-ins and drivers. -- Abstract: The Joint Texas Experimental Tokamak (J-TEXT) team has developed a new software toolkit for building Experimental Physics and Industrial Control System (EPICS) control applications called J-TEXT-EPICS. It aims to improve the development efficiency of control applications. With device-oriented features, it can be used to set or obtain the configuration or status of a device as well as invoke methods on a device. With its modularized design, its functions can be easily extended. J-TEXT-EPICS is completely compatible with the original EPICS Channel Access protocol and can be integrated into existing EPICS control systems smoothly. It is fully implemented in C number sign, thus it will benefit from abundant resources in.NET Framework. The J-TEXT control system is build with this toolkit. This paper presents the design and implementation of J-TEXT EPICS as well as its application in the J-TEXT control system

  10. A control approach for plasma density in tokamak machines

    Energy Technology Data Exchange (ETDEWEB)

    Boncagni, Luca, E-mail: luca.boncagni@enea.it [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Pucci, Daniele; Piesco, F.; Zarfati, Emanuele [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy); Mazzitelli, G. [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Monaco, S. [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy)

    2013-10-15

    Highlights: •We show a control approach for line plasma density in tokamak. •We show a control approach for pressure in a tokamak chamber. •We show experimental results using one valve. -- Abstract: In tokamak machines, chamber pre-fill is crucial to attain plasma breakdown, while plasma density control is instrumental for several tasks such as machine protection and achievement of desired plasma performances. This paper sets the principles of a new control strategy for attaining both chamber pre-fill and plasma density regulation. Assuming that the actuation mean is a piezoelectric valve driven by a varying voltage, the proposed control laws ensure convergence to reference values of chamber pressure during pre-fill, and of plasma density during plasma discharge. Experimental results at FTU are presented to discuss weaknesses and strengths of the proposed control strategy. The whole system has been implemented by using the MARTe framework [1].

  11. Orbit effects on impurity transport in a rotating tokamak plasma

    International Nuclear Information System (INIS)

    Wong, K.L.; Cheng, C.Z.

    1988-05-01

    Particle orbits in a rotating tokamak plasma are calculated from the equation of motion in the frame that rotates with the plasma. It is found that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster with a higher bounce frequency, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle orbits near the surface of a rotating tokamak are also analyzed. Orbit effects indicate that more impurities can penetrate into a plasma rotating with counter-beam injection. Particle simulation is carried out with realistic experimental parameters and the results are in qualitative agreement with some experimental observations in the Tokamak Fusion Test Reactor (TFTR). 19 refs., 15 figs

  12. Total magnetic reconnection during a tokamak major disruption

    International Nuclear Information System (INIS)

    Goetz, J.A.

    1990-09-01

    Magnetic reconnection has long been considered to be the cause of sawtooth oscillations and major disruptions in tokamak experiments. Experimental confirmation of reconnection models has been hampered by the difficulty of direct measurement of reconnection, which would involve tracing field lines for many transits around the tokamak. Perhaps the most stringent test of reconnection in a tokamak involves measurement of the safety factor q. Reconnection arising from a single helical disturbance with mode numbers m and n should raise q to m/n everywhere inside of the original resonant surface. Total reconnection should also flatten the temperature and current density profiles inside of this surface. Disruptive instabilities have been studied in the Tokapole 2, a poloidal divertor tokamak. When Tokapole 2 is operated in the material limiter configuration, a major disruption results in current termination as in most tokamaks. However, when operated in the magnetic limiter configuration current termination is suppressed and major disruptions appear as giant sawtooth oscillations. The objective of this thesis is to determine if total reconnection is occurring during major disruptions. To accomplish this goal, the poloidal magnetic field has been directly measured in Tokapole 2 with internal magnetic coils. A full two-dimensional measurement over the central current channel has been done. From these measurements, the poloidal magnetic flux function is obtained and the magnetic surfaces are plotted. The flux-surface-averaged safety factor is obtained by integrating the local magnetic field line pitch over the experimentally obtained magnetic surface

  13. Relativistic runaway electrons in tokamak plasmas

    International Nuclear Information System (INIS)

    Jaspers, R.E.

    1995-01-01

    Runaway electrons are inherently present in a tokamak, in which an electric field is applied to drive a toroidal current. The experimental work is performed in the tokamak TEXTOR. Here runaway electrons can acquire energies of up to 30 MeV. The runaway electrons are studied by measuring their synchrotron radiation, which is emitted in the infrared wavelength range. The studies presented are unique in the sense that they are the first ones in tokamak research to employ this radiation. Hitherto, studies of runaway electrons revealed information about their loss in the edge of the discharge. The behaviour of confined runaways was still a terra incognita. The measurement of the synchrotron radiation allows a direct observation of the behaviour of runaway electrons in the hot core of the plasma. Information on the energy, the number and the momentum distribution of the runaway electrons is obtained. The production rate of the runaway electrons, their transport and the runaway interaction with plasma waves are studied. (orig./HP)

  14. MHD stability of an almost circular tokamak

    International Nuclear Information System (INIS)

    Roy, A.

    1990-10-01

    In a tokamak, the ratio β between the plasma pressure and that of the magnetic field is limited by the appearance of instabilities. The magnetic field in a tokamak reactor will always be limited by technological constraints. It is therefore crucial to know what factors have an effect on the β limit, since a zero resistivity plasma fluid model allows for theoretical reproduction of the β limits observed experimentally. Theoretical studies have shown that the distributions of pressure and current density may have a substantial effect on the β limit. The effect of the current density and pressure distributions on the β limit has been studied for tokamak with a circular core section. The best results are obtained when the current density is concentrated in the centre of the section and is nil at the periphery. But the second region of stability against ballooning modes cannot be obtained in a circular tokamak owing to the destabilisation of the universal modes. This study was then extended to the stability of plasmas the section of which is almost circular and has a point of reflection. Such configurations are vital for fusion since they allow systems in which the confinement time does not deteriorate with an increase in the additional heating power. The β limit was calculated for different positions of the reflection point. The results show that when it is displaced from the interior towards the exterior of the torus, the stability of the overall modes is progressively improved until it is vertical. But if the point of reflection is further displaced from this vertical position towards the exterior of the torus, localised modes close to the edge of the plasma are destabilised and bring about a drop in the β limit. (author) figs., tabs., 80 refs

  15. Mathematical modeling plasma transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Quiang, Ji [Univ. of Illinois, Urbana-Champaign, IL (United States)

    1997-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 1020/m3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.

  16. Mathematical modeling plasma transport in tokamaks

    International Nuclear Information System (INIS)

    Quiang, Ji

    1995-01-01

    In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%

  17. Tokamak instrumentation and controls

    International Nuclear Information System (INIS)

    Becraft, W.R.; Bettis, E.S.; Houlberg, W.A.; Onega, R.J.; Stone, R.S.

    1979-02-01

    The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine

  18. 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

  19. 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)

  20. Deposit of thin films for Tokamaks conditioning

    International Nuclear Information System (INIS)

    Valencia A, R.

    2006-01-01

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature ( -6 to 4.5 x 10 -6 Ω-m, thus taking the Z ef value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow

  1. Text Mining.

    Science.gov (United States)

    Trybula, Walter J.

    1999-01-01

    Reviews the state of research in text mining, focusing on newer developments. The intent is to describe the disparate investigations currently included under the term text mining and provide a cohesive structure for these efforts. A summary of research identifies key organizations responsible for pushing the development of text mining. A section…

  2. Contribution to the multi-machine pedestal scaling from the COMPASS tokamak

    Science.gov (United States)

    Komm, M.; Bílková, P.; Aftanas, M.; Berta, M.; Böhm, P.; Bogár, O.; Frassinetti, L.; Grover, O.; Háček, P.; Havlicek, J.; Hron, M.; Imríšek, M.; Krbec, J.; Mitošínková, K.; Naydenkova, D.; Pánek, R.; Peterka, M.; Snyder, P. B.; Stefanikova, E.; Stöckel, J.; Sos, M.; Urban, J.; Varju, J.; Vondráček, P.; Weinzettl, V.; the COMPASS Team

    2017-05-01

    First systematic measurements of pedestal structure during Ohmic and NBI-assisted Type I ELMy H-modes were performed on the COMPASS tokamak in two dedicated experimental campaigns during 2015 and 2016. By adjusting the NBI heating and a toroidal magnetic field, the electron pedestal temperature was increased from 200 eV up to 300 eV, which allowed reaching pedestal collisionality ν \\text{ped}\\ast   <  1 at q95 ~3. COMPASS has approached conditions for the Identity experiment done at JET & DIII-D, complementing the range of scanned ρ \\text{ped}\\ast . The pedestal pressure was successfully reproduced by the EPED model. The dependence of pedestal pressure width on ν \\text{ped}\\ast and β \\text{ped ~ }\\text{pol} is discussed.

  3. Topology of tokamak orbits

    International Nuclear Information System (INIS)

    Rome, J.A.; Peng, Y.K.M.

    1978-09-01

    Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well

  4. Dust Measurements in Tokamaks

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  5. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  6. 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)

  7. Measurements of plasma position in TJ-I Tokamak

    International Nuclear Information System (INIS)

    Qin, J.; Ascasibar, E.; Navarro, A.P.; Ochando, M.A.; Pastor, I.; Pedrosa, M.A.; Rodriguez, L.; Sanchez, J.; Team, TJ-I.

    1994-01-01

    This report presents the experimental measurements of plasma position in TJ-I tokamak by using small magnetic probes. The basis of method has been described in our previous work (1) in which the plasma current is considered as a filament current. The observed relations between the disruptive instabilities and plasma displacements are also show here. (Author) 7 refs

  8. 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.

  9. 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

  10. Modelling dust transport in tokamaks

    International Nuclear Information System (INIS)

    Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

    2008-01-01

    The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

  11. Diffusive heat transport across magnetic islands and stochastic layers in tokamaks

    International Nuclear Information System (INIS)

    Hoelzl, Matthias

    2010-01-01

    Heat transport in tokamak plasmas with magnetic islands and ergodic field lines was simulated at realistic plasma parameters in realistic tokamak geometries. This requires the treatment of anisotropic heat diffusion, which is more efficient along magnetic field lines by up to ten orders of magnitude than perpendicular to them. Comparisons with analytical predictions and experimental measurements allow to determine the stability properties of neoclassical tearing modes as well as the experimental heat diffusion anisotropy.

  12. A Midsize Tokamak As Fast Track To Burning Plasmas

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2010-01-01

    This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain ((ge) 10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.

  13. ASTRA - an automatic system for transport analysis in a tokamak

    International Nuclear Information System (INIS)

    Pereverzev, G.V.; Yushmanov, P.N.; Dnestrovskii, A.Yu.; Polevoi, A.R.; Tarasjan, K.N.; Zakharov, L.E.

    1991-08-01

    The set of codes described here - ASTRA (Automatic System of Transport Analysis) - is a flexible and effective tool for the study of transport mechanisms in reactor-oriented facilities of the tokamak type. Flexibility is provided within the ASTRA system by a wide choice of standard relationships, functions and subroutines representing various transport coefficients, methods of auxiliary heating and other physical processes in the tokamak plasma, as well as by the possibility of pre-setting transport equations and variables for data output in a simple and conseptually transparent form. The transport code produced by the ASTRA system provides an adequate representation of the discharges for present experimental conditions. (orig.)

  14. Simplified models for radiational losses calculating a tokamak plasma

    International Nuclear Information System (INIS)

    Arutiunov, A.B.; Krasheninnikov, S.I.; Prokhorov, D.Yu.

    1990-01-01

    To determine the magnitudes and profiles of radiational losses in a Tokamak plasma, particularly for high plasma densities, when formation of MARFE or detached-plasma takes place, it is necessary to know impurity distribution over the ionization states. Equations describing time evolution of this distribution are rather cumbersome, besides that, transport coefficients as well as rate constants of the processes involving complex ions are known nowadays with high degree of uncertainty, thus it is believed necessary to develop simplified, half-analytical models describing time evolution of the impurities analysis of physical processes taking place in a Tokamak plasma on the base of the experimental data. (author) 6 refs., 2 figs

  15. Development of a visualized software for tokamak experiment data processing

    International Nuclear Information System (INIS)

    Cao Jianyong; Ding Xuantong; Luo Cuiwen

    2004-01-01

    With the VBA programming in Microsoft Excel, the authors have developed a post-processing software of experimental data in tokamak. The standard formal data in the HL-1M and HL-2A tokamaks can be read, displayed in Excel, and transmitted directly into the MATLAB workspace, for displaying pictures in MATLAB with the software. The authors have also developed data post-processing software in MATLAB environment, which can read standard format data, display picture, supply visual graphical user interface and provide part of advanced signal processing ability

  16. A generic access to shot-based data for European Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Signoret, J.; Imbeaux, F. [Association EURATOM-CEA, CEA / DSM / Institut de Recherche sur la Fusion par confinement Magnetique, CEA-Cadarache, 13 - ST-Paul-Lez-Durance (France)

    2009-07-01

    The EFDA Integrated Tokamak Modeling Task Force has defined a data structure offering a generic representation of the properties of physics problems and tokamak subsystem characteristics. It gathers the hardware description, modeling results and data measured during experiments, structured in terms of Consistent Physical Objects (CPOs). A generic tool has been developed to retrieve shot-based data from the various European tokamak databases: Exp2ITM. A tokamak specific XML 'mapping file' is used to map the local data formats to the ITM (Integrated Tokamak Modeling) data format. Exp2ITM is then dynamically generated from the ITM data structure and uses generic procedures to import the shot-based data. Successful tests show we have managed to import into the ITM DB experimental data from Jet and Tore-Supra. This document is a poster. (authors)

  17. Particle and heat transport in Tokamaks

    International Nuclear Information System (INIS)

    Chatelier, M.

    1984-01-01

    A limitation to performances of tokamaks is heat transport through magnetic surfaces. Principles of ''classical'' or ''neoclassical'' transport -i.e. transport due to particle and heat fluxes due to Coulomb scattering of charged particle in a magnetic field- are exposed. It is shown that beside this classical effect, ''anomalous'' transport occurs; it is associated to the existence of fluctuating electric or magnetic fields which can appear in the plasma as a result of charge and current perturbations. Tearing modes and drift wave instabilities are taken as typical examples. Experimental features are presented which show that ions behave approximately in a classical way whereas electrons are strongly anomalous [fr

  18. 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

  19. 'Snowflake' H Mode in a Tokamak Plasma

    International Nuclear Information System (INIS)

    Piras, F.; Coda, S.; Duval, B. P.; Labit, B.; Marki, J.; Moret, J.-M.; Pitzschke, A.; Sauter, O.; Medvedev, S. Yu.

    2010-01-01

    An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a 'snowflake' (second-order null) divertor configuration for the first time in the TCV tokamak. This regime exhibits 2 to 3 times lower ELM frequency and 20%-30% increased normalized ELM energy (ΔW ELM /W p ) compared to an identically shaped, conventional single-null diverted H mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. The capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally.

  20. Study on assembly techniques and procedures for ITER tokamak device

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi; Ue, Koichi; Shimizu, Katsusuke; Onozuka, Masanori

    2006-06-01

    The International Thermonuclear Experimental Reactor (ITER) tokamak is mainly composed of a doughnut-shaped vacuum vessel (VV), four types of superconducting coils such as toroidal field coils (TF coils) arranged around the VV, and in-vessel components, such as blanket and divertor. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of the VV and the TF coil are required to be a high accuracy of ±3 mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements as well as the configuration of the tokamak with large size and heavy weight. Based on the above backgrounds, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The tokamak assembly operations are categorized into six work break down structures (WBS), i.e., (1) preparation for assembly operations, (2) sub-assembly of the 40deg sector composed of 40deg VV sector, two TF coils and thermal shield between VV and TF coil at the assembly hall, (3) completion of the doughnut-shaped tokamak assembly composed of nine 40deg sectors in the cryostat at the tokamak pit, (4) measurement of positioning and accuracy after the completion of the tokamak assembly, (5) installation of the ex-vessel components, and (6) installation of in-vessel components. In the present report, two assembly operations of (2) and (3) in the above six WBS, which are the most critical in the tokamak assembly, are mainly described. The report describes the following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology

  1. Tokamak building-design considerations for a large tokamak device

    International Nuclear Information System (INIS)

    Barrett, R.J.; Thomson, S.L.

    1981-01-01

    Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

  2. Characterization of the Tokamak Novillo in cleaning regime

    International Nuclear Information System (INIS)

    Lopez C, R.; Melendez L, L.; Valencia A, R.; Chavez A, E.; Colunga S, S.; Gaytan G, E.

    1992-02-01

    In this work the obtained results of the investigation about the experimental characterization of those low energy pulsed discharges of the Tokamak Novillo are reported. With this it is possible to fix the one operation point but appropriate of the Tokamak to condition the chamber in the smallest possible time for the cleaning discharges regime before beginning the main discharge. The characterization of the cleaning discharges in those Tokamaks is an unique process and characteristic of each device, since the good points of operation are consequence of those particularities of the design of the machine. In the case of the Tokamak Novillo, besides characterizing it a contribution is made to the cleaning discharges regime which consists on the one product of the current peak to peak of plasma by the duration of the discharge Ip t like reference parameter for the optimization of the operation of the device in the cleaning discharge regime. The maximum value of the parameter I (p) t, under different work conditions, allowed to find the good operation point to condition the discharges chamber of the Tokamak Novillo in short time and to arrive to a regime in which is not necessary the preionization for the obtaining of the cleaning discharges. (Author)

  3. Key features and progress of the KSTAR tokamak engineering

    International Nuclear Information System (INIS)

    Bak, J.S.; Choi, C.H.; Oh, Y.K.

    2003-01-01

    Substantial progress of the KSTAR tokamak engineering has been made on major tokamak structures, superconducting magnets, in-vessel components, diagnostic system, heating system, and power supplies. The engineering design has been elaborated to the extent necessary to allow a realistic assessment of its feasibility, performance, and cost. The prototype fabrication has been carried out to establish the reliable fabrication technologies and to confirm the validation of analyses employed for the KSTAR design. The completion of experimental building with beneficial occupancy for machine assembly was accomplished in Sep. 2002. The construction of special utility such as cryo-plant, de-ionized water-cooling system, and main power station will begin upon completion of building construction. The commissioning, construction, fabrication, and assembly of the whole facility will be going on by the end of 2005. This paper describes the main design features and engineering progress of the KSTAR tokamak, and elaborates the work currently underway. (author)

  4. Electron cyclotron current drive efficiency in an axisymmetric tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez-Tapia, C.; Beltran-Plata, M. [Instituto Nacional de Investigaciones Nucleares, Dept. de Fisica, Mexico D.F. (Mexico)

    2004-07-01

    The neoclassical transport theory is applied to calculate electron cyclotron current drive (ECCD) efficiency in an axisymmetric tokamak in the low-collisionality regime. The tokamak ordering is used to obtain a system of equations that describe the dynamics of the plasma where the nonlinear ponderomotive (PM) force due to high-power radio-frequency (RF) waves is included. The PM force is produced around an electron cyclotron resonant surface at a specific poloidal location. The ECCD efficiency is analyzed in the cases of first and second harmonics (for different impinging angles of the RF waves) and it is validated using experimental parameter values from TCV and T-10 tokamaks. The results are in agreement with those obtained by means of Green's function techniques. (authors)

  5. A model for plasma discharges simulation in Tokamak devices

    International Nuclear Information System (INIS)

    Fonseca, Antonio M.M.; Silva, Ruy P. da; Galvao, Ricardo M.O.; Kusnetzov, Yuri; Nascimento, I.C.; Cuevas, Nelson

    2001-01-01

    In this work, a 'zero-dimensional' model for simulation of discharges in Tokamak machine is presented. The model allows the calculation of the time profiles of important parameters of the discharge. The model was applied to the TCABR Tokamak to study the influence of parameters and physical processes during the discharges. Basically it is constituted of five differential equations: two related to the primary and secondary circuits of the ohmic heating transformer and the other three conservation equations of energy, charge and neutral particles. From the physical model, a computer program has been built with the objective of obtaining the time profiles of plasma current, the current in the primary of the ohmic heating transformer, the electronic temperature, the electronic density and the neutral particle density. It was also possible, with the model, to simulate the effects of gas puffing during the shot. The results of the simulation were compared with the experimental results obtained in the TCABR Tokamak, using hydrogen gas

  6. Natural current profiles in tokamaks

    International Nuclear Information System (INIS)

    Biskamp, D.

    1986-01-01

    It is proposed that a certain class of equilibrium, which follow from an elementary variational principle, are the natural current profiles in tokamaks, to which actual discharge profiles tend to relax. (orig.)

  7. Alcator C-Mod Tokamak

    Data.gov (United States)

    Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...

  8. JUST: Joint Upgraded Spherical Tokamak

    International Nuclear Information System (INIS)

    Azizov, E.A.; Dvorkin, N.Ya.; Filatov, O.G.

    1997-01-01

    The main goals, ideas and the programme of JUST, spherical tokamak (ST) for the plasma burn investigation, are presented. The place and prospects of JUST in thermonuclear investigations are discussed. (author)

  9. 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.

  10. New directions in tokamak reactors

    International Nuclear Information System (INIS)

    Baker, C.C.

    1985-01-01

    New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs

  11. The Tokamak IST-TOK

    International Nuclear Information System (INIS)

    Varandas, C.A.F.; Cabral, J.A.C.; Manso, M.E.

    1991-01-01

    A small tokamak is under construction at the Portuguese Technical Superior Institute. The main objective is to create a home based laboratory in which an independent scientific program might be developed. (L.C.J.A.). 14 refs, 6 figs

  12. Numerical Tokamak Project code comparison

    International Nuclear Information System (INIS)

    Waltz, R.E.; Cohen, B.I.; Beer, M.A.

    1994-01-01

    The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

  13. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group

    2003-01-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  14. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-01-01

    The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

  15. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

    2003-07-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  16. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

    2003-12-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  17. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes

    2003-01-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  18. High Beta Tokamak research

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  19. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-09-01

    A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  20. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-01-01

    A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  1. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  2. Measurement of inner wall limiter SOL widths in KSTAR tokamak

    Directory of Open Access Journals (Sweden)

    J.G. Bak

    2017-08-01

    Full Text Available Scrape-off layer (SOL widths λq are presented from the KSTAR tokamak using fast reciprocating Langmuir probe assembly (FRLPA measurements at the outboard mid-plane (OMP and the infra-Red (IR thermography at inboard limiter tiles in moderately elongated (κ = 1.45 – 1.55 L-mode inner wall-limited (IWL plasmas under experimental conditions such as BT = 2.0 T, PNBI = 1.4 – 1.5 MW, line averaged densities 2.5 – 5.1 × 1019 m−3 and plasma current Ip = 0.4 − 0.7 MA. There is clear evidence for a double exponential structure in q||(r from the FRLPA such that, for example at Ip = 0.6 MA, a narrow feature, λq,near (=3.5 mm is found close to the LFCS, followed by a broader width, λq,main (=57.0 mm. Double exponential profiles (λq,near = 1.5 – 2.8 mm, λq,main = 17.0 – 35.0 mm can be also observed in the IR heat flux mapped to the OMP throughout the range of Ip investigated. In addition, analysis of SOL turbulence statistics obtained with the FRLPA shows high relative fluctuation levels and positively skewed distributions in electron temperature and ion particle flux across the SOL, with both properties increasing for longer distance from the LCFS, as often previously observed in the tokamaks. Interestingly, the fluctuation character expressed in terms of spectral distributions remains unchanged in passing from the narrow to the broad SOL heat flux channel.

  3. 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

  4. Diagnostics Neutral Beam Injector at the TCV Tokamak

    International Nuclear Information System (INIS)

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

    2001-10-01

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

  5. 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.

  6. Data processing system for spectroscopy at Novillo Tokamak

    International Nuclear Information System (INIS)

    Ortega C, G.; Gaytan G, E.

    1998-01-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  7. ICRF heating experiments in JFT-2 tokamak

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi

    1986-01-01

    This is an experimental study of ICRF heating on JFT-2 Tokamak in Japan Atomic Energy Research Institute. In this study, we first clarified physical and engineering problems of ICRF heating of tokamak plasma. Next, we optimized the design of the ICRF heating system, and the plasma parameters for the heating. Finally, we could demonstrate a high efficiency of this additional heating method by launching RF power which is two or three times as large as an ohmic input power to a plasma. And we achieved following things. (1) We optimized a design of an antenna, and we improved a durability of the system for high voltage. With the result that we achieved the maximum power density on an antenna. (2) We demonstrated that electron heating regime and ion heating regime can be easily accessed by controlling plasma parameters. Also we found the optimum heating conditions in each heating regime. (3) We experimentally clarified the production mechanism of impurities during ICRF heating. We could reduce the influx of metal impurity ions to a plasma by employing low z materials for limiters and antenna shields. Consequently, we improved a heating efficiency of electrons. Next, we studied a power balance of plasma during ICRF heating, and we could compare heating characteristics of ICRF with other additional heatings on JFT-2. (author)

  8. Plasma transport in a compact ignition tokamak

    International Nuclear Information System (INIS)

    Singer, C.E.; Ku, L.P; Bateman, G.

    1987-02-01

    Nominal predicted plasma conditions in a compact ignition tokamak are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models which have given almost equally good fits to experimental data. Using a transport model which best fits the data, thermonuclear ignition occurs in a Compact Ignition Tokamak design with major radius 1.32 m, plasma half-width 0.43 m, elongation 2.0, and toroidal field and plasma current ramped in six seconds from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the 3 He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates, have a large effect on ignition and on the maximum beta that can be achieved

  9. Differential and Integral Models of TOKAMAK

    Directory of Open Access Journals (Sweden)

    Ivo Dolezel

    2004-01-01

    Full Text Available Modeling of 3D electromagnetic phenomena in TOKAMAK with typically distributed main and additional coils is not an easy business. Evaluated must be not only distribution of the magnetic field, but also forces acting in particular coils. Use of differential methods (such as FDM or FEM for this purpose may be complicated because of geometrical incommensurability of particular subregions in the investigated area or problems with the boundary conditions. That is why integral formulation of the problem may sometimes be an advantages. The theoretical analysis is illustrated on an example processed by both methods, whose results are compared and discussed.

  10. The critical temperature gradient model of plasma transport: applications to Jet and future tokamaks

    International Nuclear Information System (INIS)

    Rebut, P.H.; Lallia, P.P.; Watkins, M.L.

    1989-01-01

    The diversity and complexity of behaviour in tokamak plasmas place strong constraints on any model attempting a description in terms of a single underlying phenomenon. Assuming that turbulence in the magnetic topology is the underlying phenomenon, specific expressions for electron and ion heat flux are derived from heuristic and dimensional arguments. When used in plasma transport codes, rather satisfactory simulations of experimental results are achieved in different sized tokamaks in various regimes of operation. Predictions are given for the expected performance of JET at full planned power and implications for next step tokamaks are indicated

  11. 3D simulation studies of tokamak plasmas using MHD and extended-MHD models

    International Nuclear Information System (INIS)

    Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.

    1996-01-01

    The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data

  12. Theory-based scaling of the SOL width in circular limited tokamak plasmas

    International Nuclear Information System (INIS)

    Halpern, F.D.; Ricci, P.; Labit, B.; Furno, I.; Jolliet, S.; Loizu, J.; Mosetto, A.; Arnoux, G.; Silva, C.; Gunn, J.P.; Horacek, J.; Kočan, M.; LaBombard, B.

    2013-01-01

    A theory-based scaling for the characteristic length of a circular, limited tokamak scrape-off layer (SOL) is obtained by considering the balance between parallel losses and non-linearly saturated resistive ballooning mode turbulence driving anomalous perpendicular transport. The SOL size increases with plasma size, resistivity, and safety factor q. The scaling is verified against flux-driven non-linear turbulence simulations, which reveal good agreement within a wide range of dimensionless parameters, including parameters closely matching the TCV tokamak. An initial comparison of the theory against experimental data from several tokamaks also yields good agreement. (letter)

  13. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Hively, L.M.; Sigmar, D.J.

    1989-09-01

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

  14. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    Amiel, Stephane

    2014-01-01

    During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author) [fr

  15. Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  16. Tokamak engineering test reactor

    International Nuclear Information System (INIS)

    Conn, R.W.; Jassby, D.L.

    1975-07-01

    The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

  17. Design of the ITER Tokamak Assembly Tools

    International Nuclear Information System (INIS)

    Park, Hyunki; Her, Namil; Kim, Byungchul; Im, Kihak; Jung, Kijung; Lee, Jaehyuk; Im, Kisuk

    2006-01-01

    ITER (International Thermonuclear Experimental Reactor) Procurement allocation among the seven Parties, EU, JA, CN, IN , KO, RF and US had been decided in Dec. 2005. ITER Tokamak assembly tools is one of the nine components allocated to Korea for the construction of the ITER. Assembly tools except measurement and common tools are supplied to assemble the ITER Tokamak and classified into 9 groups according to components to be assembled. Among the 9 groups of assembly tools, large-sized Sector Sub-assembly Tools and Sector Assembly Tools are used at the first stage of ITER Tokamak construction and need to be designed faster than seven other assembly tools. ITER IT (International Team) proposed Korea to accomplish ITA (ITER Transitional Arrangements) Task on detailed design, manufacturing feasibility and contract specification of specific, large sized tools such as Upending Tool, Lifting Tool, Sector Sub-assembly Tool and Sector Assembly Tool in Oct. 2004. Based on the concept design by ITER IT, Korea carried out ITA Task on detailed design of large-sized and specific Sector Sub-assembly and Sector Assembly Tools until Mar. 2006. The Sector Sub-assembly Tools mainly consist of the Upending, Lifting, Vacuum Vessel Support and Bracing, and Sector Sub-assembly Tool, among which the design of three tools are herein. The Sector Assembly Tools mainly consist of the Toroidal Field (TF) Gravity Support Assembly, Sector In-pit Assembly, TF Coil Assembly, Vacuum Vessel (VV) Welding and Vacuum Vessel Thermal Shield (TS) Assembly Tool, among which the design of Sector In-pit Assembly Tool is described herein

  18. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  19. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  20. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  1. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  2. 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

  3. Resistive instabilities in tokamaks

    International Nuclear Information System (INIS)

    Rutherford, P.H.

    1985-10-01

    Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed

  4. Classical tokamak transport theory

    International Nuclear Information System (INIS)

    Nocentini, Aldo

    1982-01-01

    A qualitative treatment of the classical transport theory of a magnetically confined, toroidal, axisymmetric, two-species plasma is presented. The 'weakly collisional' ('banana' and 'plateau') and 'collision dominated' ('Pfirsch-Schlueter' and 'highly collisional') regimes, as well as the Ware effect are discussed. The method used to evaluate the diffusion coffieicnts of particles and heat in the weakly collisional regime is based on stochastic argument, that requires an analysis of the characteristic collision frequencies and lengths for particles moving in a tokamak-like magnetic field. The same method is used to evaluate the Ware effect. In the collision dominated regime on the other hand, the particle and heat fluxes across the magnetic field lines are dominated by macroscopic effects so that, although it is possible to present them as diffusion (in fact, the fluxes turn out to be proportional to the density and temperature gradients), a macroscopic treatment is more appropriate. Hence, fluid equations are used to inveatigate the collision dominated regime, to which particular attention is devoted, having been shown relatively recently that it is more complicated than the usual Pfirsch-Schlueter regime. The whole analysis presented here is qualitative, aiming to point out the relevant physical mechanisms involved in the various regimes more than to develop a rigorous mathematical derivation of the diffusion coefficients, for which appropriate references are given. (author)

  5. Equilibrium vertical field in the TBR Tokamak

    International Nuclear Information System (INIS)

    Ueta, A.Y.

    1985-01-01

    An experimental study on the influence of the vertical magnetic field of the TBR tokamak on the stability and equilibrium of plasma column, was done. Magnetic pick-up coils were built to measure plasma current and position, together with active networks, necessary fo the electronic processing of signals. Some measurements were on the space configuration of the vertical field, and on the influence due to the toroidal vessel. From the data obtained it was possible to discuss the influence of the currents induced on the vessel surface, on plasma equilibrium. Theoretical and experimental results of the vertica field, as a function of plasma current were compared, and allowed an evaluation of the plasma kinetic pressure and temperature. (Author) [pt

  6. Alfven wave coupling in large tokamaks

    International Nuclear Information System (INIS)

    Borg, G.G.; Knight, A.J.; Lister, J.B.; Appert, K.; Vaclavik, J.

    1988-01-01

    Supplementary plasma heating by Alfven waves (AWH) has been extensively studied both theoretically and experimentally for small, low temperature plasmas. However, only a few studies of AWH have been performed for fusion plasmas. In this paper the cylindrical kinetic code ISMENE is used to address problems af AWH in a large tokamak. The results of calculations are presented which show that the antenna loading scales with frequency and vessel dimensions according to ideal MHD theory. A sample scaling of the experimental antenna loading measured in TCA to the loading predicted for a fusion plasma is presented. We discuss whether this loading leads to a realistic antenna design. The choice of a suitable antenna configuration, mode number and operating frequency is presented for NET parameters with a typical operating scenario. (author) 6 figs., 8 refs

  7. [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

  8. Profile consistency, anomalous electron thermal conduction, and confinement analysis of tokamak devices

    International Nuclear Information System (INIS)

    Qu Wenxiao

    1992-01-01

    Assuming that there exists a position in the tokamak plasma where the energy transport is dominated by local anomalous electron thermal conduction and taking advantage of the basic experimental result usually referred to as profile consistency, the authors obtain a more convincing approach to the description of the confinement property of tokamak devices without touching upon the physical mechanism of global plasma energy transport. 8 refs

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

    International Nuclear Information System (INIS)

    Pachtman, A.

    1986-09-01

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

  10. The role of the spherical tokamak in clarifying tokamak physics

    International Nuclear Information System (INIS)

    Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.

    1999-01-01

    The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)

  11. Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Abiteboul, J.

    2012-10-01

    The goal of magnetic confinement devices such as tokamaks is to produce energy from nuclear fusion reactions in plasmas at low densities and high temperatures. Experimentally, toroidal flows have been found to significantly improve the energy confinement, and therefore the performance of the machine. As extrinsic momentum sources will be limited in future fusion devices such as ITER, an understanding of the physics of toroidal momentum transport and the generation of intrinsic toroidal rotation in tokamaks would be an important step in order to predict the rotation profile in experiments. Among the mechanisms expected to contribute to the generation of toroidal rotation is the transport of momentum by electrostatic turbulence, which governs heat transport in tokamaks. Due to the low collisionality of the plasma, kinetic modeling is mandatory for the study of tokamak turbulence. In principle, this implies the modeling of a six-dimensional distribution function representing the density of particles in position and velocity phase-space, which can be reduced to five dimensions when considering only frequencies below the particle cyclotron frequency. This approximation, relevant for the study of turbulence in tokamaks, leads to the so-called gyrokinetic model and brings the computational cost of the model within the presently available numerical resources. In this work, we study the transport of toroidal momentum in tokamaks in the framework of the gyrokinetic model. First, we show that this reduced model is indeed capable of accurately modeling momentum transport by deriving a local conservation equation of toroidal momentum, and verifying it numerically with the gyrokinetic code GYSELA. Secondly, we show how electrostatic turbulence can break the axisymmetry and generate toroidal rotation, while a strong link between turbulent heat and momentum transport is identified, as both exhibit the same large-scale avalanche-like events. The dynamics of turbulent transport are

  12. 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)

  13. Large aspect ratio tokamak study

    International Nuclear Information System (INIS)

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

    1979-01-01

    The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured

  14. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

    1981-02-01

    Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

  15. The stability margin on EAST tokamak

    International Nuclear Information System (INIS)

    Jin-Ping, Qian; Bao-Nian, Wan; Biao, Shen; Bing-Jia, Xiao; Walker, M.L.; Humphreys, D.A.

    2009-01-01

    The experimental advanced superconducting tokamak (EAST) is the first full superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. Its poloidal coils are relatively far from the plasma due to the necessary thermal isolation from the superconducting magnets, which leads to relatively weaker coupling between plasma and poloidal field. This may cause more difficulties in controlling the vertical instability by using the poloidal coils. The measured growth rates of vertical stability are compared with theoretical calculations, based on a rigid plasma model. Poloidal beta and internal inductance are varied to investigate their effects on the stability margin by changing the values of parameters α n and γ n (Howl et al 1992 Phys. Fluids B 4 1724), with plasma shape fixed to be a configuration with k = 1.9 and δ = 0.5. A number of ways of studying the stability margin are investigated. Among them, changing the values of parameters κ and l i is shown to be the most effective way to increase the stability margin. Finally, a guideline of stability margin M s (κ, l i , A) to a new discharge scenario showing whether plasmas can be stabilized is also presented in this paper

  16. Anomalous transport in the tokamak edge

    International Nuclear Information System (INIS)

    Vayakis, G.

    1991-04-01

    The tokamak edge has been studied with arrays of Langmuir and magnetic probes on the DITE and COMPASS-C devices. Measurements of plasma parameters such as density, temperature and radial magnetic field were taken in order to elucidate the character, effect on transport and origin of edge fluctuations. The tokamak edge is a strongly-turbulent environment, with large electrostatic fluctuation levels and broad spectra. The observations, including direct correlation measurements, are consistent with a picture in which the observed magnetic field fluctuations are driven by the perturbations in electrostatic parameters. The propagation characteristics of the turbulence, investigated using digital spectral techniques, appear to be dominated by the variation of the radial electric field, both in limiter and divertor plasmas. A shear layer is formed, associated in each case with the last closed flux surface. In the shear layer, the electrostatic wavenumber spectra are significantly broader. The predictions of a drift wave model (DDGDT) and of a family of models evolving from the rippling mode (RGDT group), are compared with experimental results. RGDT, augmented by impurity radiation effects, is shown to be the most reasonable candidate to explain the nature of the edge turbulence, only failing in its estimate of the wavenumber range. (Author)

  17. MDSplus integration at TCABR tokamak: Current status

    International Nuclear Information System (INIS)

    Sá, W.P. de; Ronchi, G.

    2016-01-01

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  18. Integrated plasma control for high performance tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.

    2005-01-01

    Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)

  19. MDSplus integration at TCABR tokamak: Current status

    Energy Technology Data Exchange (ETDEWEB)

    Sá, W.P. de, E-mail: pires@if.usp.br; Ronchi, G., E-mail: gronchi@if.usp.br

    2016-11-15

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  20. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1994-01-01

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  1. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  2. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  3. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...

  4. A flexible software architecture for tokamak discharge control systems

    International Nuclear Information System (INIS)

    Ferron, J.R.; Penaflor, B.; Walker, M.L.; Moller, J.; Butner, D.

    1995-01-01

    The software structure of the plasma control system in use on the DIII-D tokamak experiment is described. This system implements control functions through software executing in real time on one or more digital computers. The software is organized into a hierarchy that allows new control functions needed to support the DIII-D experimental program to be added easily without affecting previously implemented functions. This also allows the software to be portable in order to create control systems for other applications. The tokamak operator uses an X-windows based interface to specify the time evolution of a tokamak discharge. The interface provides a high level view for the operator that reduces the need for detailed knowledge of the control system operation. There is provision for an asynchronous change to an alternate discharge time evolution in response to an event that is detected in real time. Quality control is enhanced through off-line testing that can make use of software-based tokamak simulators

  5. 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.)

  6. Midplane Faraday rotation: A densitometer for large tokamaks

    International Nuclear Information System (INIS)

    Jobes, F.C.; Mansfield, D.K.

    1992-01-01

    The density in a large tokamak such as International Thermonuclear Experimental Reactor (ITER), or any of the proposed future US machines, can be determined by measuring the Faraday rotation of a 10.6 μm laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then n e (R) can be readily obtained with a simple Abel inversion about the center line of the tokamak. For a large machine, operated at a full field of 30 T m and a density of 2x10 20 /m 3 , the rotation angle would be quite large-about 60 degree for two passes. A layout in which a single laser beam is fanned out in the horizontal midplane of the tokamak, with a set of retroreflectors on the far side of the vacuum vessel, would provide good spatial resolution, depending only upon the number of reflectors. With this proposed layout, only one window would be needed. Because the rotation angle is never more than 1 ''fringe,'' the data is always good, and it is also a continuous measurement in time. Faraday rotation is dependent only upon the plasma itself, and thus is not sensitive to vibration of the optical components. Simulations of the expected results show that ITER, or any large tokamak, existing or proposed, would be well served even at low densities by a midplane Faraday rotation densitometer of ∼64 channels

  7. Study text and sets of questions for the training and examination of selected personnel of nuclear research facilities. Issue 2. Experimental teaching methods

    International Nuclear Information System (INIS)

    Fleischhans, J.; Hejzlar, R.; Hermansky, B.

    2004-01-01

    The VR-1 teaching reactor is described, 14 experimental exercises are given (e.g. Starting up and running the VR-1 reactor; Neutron detection and detectors; Measuring delayed neutrons ; Reactivity measurement; Control rod calibration; ...) and practical training at the existing Czech research reactors (LVR-15; LR-0; VR-1) is briefly highlighted. (P.A.)

  8. Energy losses on tokamak startup

    International Nuclear Information System (INIS)

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1983-01-01

    During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells

  9. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately

  10. 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

  11. Tokamak impurity-control techniques

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1980-01-01

    A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)

  12. Multimegawatt neutral beams for tokamaks

    International Nuclear Information System (INIS)

    Kunkel, W.B.

    1979-03-01

    Most of the large magnetic confinement experiments today and in the near future use high-power neutral-beam injectors to heat the plasma. This review briefly describes this remarkable technique and summarizes recent results as well as near term expectations. Progress has been so encouraging that it seems probable that tokamaks will achieve scientific breakeven before 1990

  13. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

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

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254 ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  14. ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

    International Nuclear Information System (INIS)

    HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

    2002-01-01

    A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

  15. Controlled thermonuclear fusion and the latest progress on China's HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Li Jiangang; Yang Yu

    2003-01-01

    After 50 years of research on controlled thermonuclear fusion, a new stage will be reached in 2003, when a site for the International Thermonuclear Experimental Reactor project will be chosen to start the construction. Scientists hope that this project could herald a new era in which the energy problem will be solved completely. The great progress made on the HT-7 superconducting tokamak in China has provided positive and powerful support for fusion research. The HT-7 is one of the only two superconducting tokamaks in the world that can carry out minute-scale high temperature plasma research, and has achieved a duration of 63.95s for the hot plasma discharge. This is a major step towards real steady-state operation of the tokamak configuration. We present an overview of the latest progress on the tokamak experiments in the Institute of Plasma Physics, Chinese Academy of Sciences

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  17. Application of avalanche photodiode for soft X-ray pulse-height analyses in the Ht-7 tokamak

    CERN Document Server

    Shi Yue Jiang; Hu Li Qun; Sun Yan Jun; LiuSheng; Ling Bil

    2002-01-01

    An avalanche photodiode (APD) has been used as soft X-ray energy pulse-height analysis system for the measurement of the electron temperature on the HT-7 tokamak. The experimental results obtained with the APD with its inferior energy resolution show a little difference compared to the conventional high energy-resolution Si (Li) detector. Both numerical analysis and experimental results prove that the APD is good enough for application of the electron temperature measurement in tokamaks.

  18. Thermally excited proton spin-flip laser emission in tokamaks

    International Nuclear Information System (INIS)

    Arunasalam, V.; Greene, G.J.

    1993-07-01

    Based on statistical thermodynamic fluctuation arguments, it is shown here for the first time that thermally excited spin-flip laser emission from the fusion product protons can occur in large tokamak devices that are entering the reactor regime of operation. Existing experimental data from TFTR supports this conjecture, in the sense that these measurements are in complete agreement with the predictions of the quasilinear theory of the spin-flip laser

  19. Review of tearing mode stabilization by RF power in tokamaks

    International Nuclear Information System (INIS)

    Giruzzi, G.; Zabiego, M.; Zohm, H.

    1999-01-01

    Control of tearing modes by means of heating and current drive inside the magnetic islands is one of the most important applications of RF power in tokamak reactors. The theoretical basis of this concept is reviewed, focusing on aspects related to RF-plasma interaction. Applications to the stabilization of neoclassical tearing modes in ITER by Electron Cyclotron Current Drive are presented to illustrate the basic physical dependences. The most significant experimental results and prospects for future applications are also discussed

  20. 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)

  1. 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)

  2. Characterization of the Novillo Tokamak in main discharge regime

    International Nuclear Information System (INIS)

    Lopez C, R.; Melendez L, L.; Chavez A, E.; Colunga S, S.; Valencia A, R.; Gaytan G, E.

    1992-07-01

    The analytical procedure to carry out the establishment of the discharge in a Tokamak including: a) Ionization, b) Diffusion losses, recombination, union, drift speed, spurious fields, and c) Electric field is presented. In an experimental way a procedure settles down by means of which it is characterized the plasma, specially a new characteristic discharge parameter is settled down and it is the plasma current by the duration of the (I p t) discharge. (Author)

  3. Generalized saddle point condition for ignition in a tokamak reactor with temperature and density profiles

    International Nuclear Information System (INIS)

    Mitari, O.; Hirose, A.; Skarsgard, H.M.

    1989-01-01

    In this paper, the concept of a generalized ignition contour map, is extended to the realistic case of a plasma with temperature and density profiles in order to study access to ignition in a tokamak reactor. The generalized saddle point is found to lie between the Lawson and ignition conditions. If the height of the operation path with Goldston L-mode scaling is higher than the generalized saddle point, a reactor can reach ignition with this scaling for the case with no confinement degradation effect due to alpha-particle heating. In this sense, the saddle point given in a general form is a new criterion for reaching ignition. Peaking the profiles for the plasma temperature and density can lower the height of the generalized saddle point and help a reactor to reach ignition. With this in mind, the authors can judge whether next-generation tokamaks, such as Compact Ignition Tokamak, Tokamak Ignition/Burn Experimental Reactor, Next European Torus, Fusion Experimental Reactor, International Tokamak Reactor, and AC Tokamak Reactor, can reach ignition with realistic profile parameters and an L-mode scaling law

  4. A modularized operator interface framework for Tokamak based on MVC design pattern

    International Nuclear Information System (INIS)

    Yin, Xuan; Zheng, Wei; Zhang, Ming; Zhang, Jing; Zhuang, G.; Ding, T.

    2014-01-01

    Highlights: • Our framework is based on MVC design pattern. • XML is used to cope with minor difference between different applications. • Functions dealing with EPICS and MDSplus have been modularized into reusable modules. • The modularized framework will shorten J-TEXT's software development cycle. - Abstract: Facing various and continually changing experimental needs, the J-TEXT Tokamak experiment requires home-made software applications developed for different sub-systems. Though dealing with different specific problems, these software applications usually share a lot of functionalities in common. With the goal of improving the productivity of research groups, J-TEXT has designed a C# desktop application framework which is mainly focused on operator interface development. Following the Model–View–Controller (MVC) design pattern, the main functionality dealing with Experimental Physics and Industrial Control System (EPICS) or MDSplus has been modularized into reusable modules. Minor difference among applications can be coped with XML configuration files. In this case, developers are able to implement various kinds of operator interface without knowing the implementation details of the bottom functions in Models, mainly focusing on Views and Controllers. This paper presents J-TEXT C# desktop application framework, introducing the technology of fast development of the modularized operator interface. Some experimental applications designed in this framework have been already deployed in J-TEXT, and will be introduced in this paper

  5. Computational studies in tokamak equilibrium and transport

    International Nuclear Information System (INIS)

    Braams, B.J.

    1986-01-01

    This thesis is concerned with some problems arising in the magnetic confinement approach to controlled thermonuclear fusion. The work address the numerical modelling of equilibrium and transport properties of a confined plasma and the interpretation of experimental data. The thesis is divided in two parts. Part 1 is devoted to some aspects of the MHD equilibrium problem, both in the 'direct' formulation (given an equation for the plasma current, the corresponding equilibrium is to be determined) and in the 'inverse' formulation (the interpretation of measurements at the plasma edge). Part 2 is devoted to numerical studies of the edge plasma. The appropriate Navier-Stokes system of fluid equations is solved in a two-dimensional geometry. The main interest of this work is to develop an understanding of particle and energy transport in the scrape-off layer and onto material boundaries, and also to contribute to the conceptual design of the NET/INTOR tokamak reactor experiment. (Auth.)

  6. Ignition probabilities for Compact Ignition Tokamak designs

    International Nuclear Information System (INIS)

    Stotler, D.P.; Goldston, R.J.

    1989-09-01

    A global power balance code employing Monte Carlo techniques had been developed to study the ''probability of ignition'' and has been applied to several different configurations of the Compact Ignition Tokamak (CIT). Probability distributions for the critical physics parameters in the code were estimated using existing experimental data. This included a statistical evaluation of the uncertainty in extrapolating the energy confinement time. A substantial probability of ignition is predicted for CIT if peaked density profiles can be achieved or if one of the two higher plasma current configurations is employed. In other cases, values of the energy multiplication factor Q of order 10 are generally obtained. The Ignitor-U and ARIES designs are also examined briefly. Comparisons of our empirically based confinement assumptions with two theory-based transport models yield conflicting results. 41 refs., 11 figs

  7. Transport Bifurcation in a Rotating Tokamak Plasma

    International Nuclear Information System (INIS)

    Highcock, E. G.; Barnes, M.; Schekochihin, A. A.; Parra, F. I.; Roach, C. M.; Cowley, S. C.

    2010-01-01

    The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.

  8. Tokamak fusion test reactor FELIX plate experiment

    International Nuclear Information System (INIS)

    Hua, T.O.; Nygren, R.E.; Turner, L.R.

    1986-01-01

    For a conducting material exposed to both a time-varying and a static magnetic field, such as a limiter blade in a tokamak, the induced eddy currents and the deflection arising from those eddy currents can be strongly coupled. The coupling effects reduce the currents and deflections markedly, sometimes an order of magnitude, from the values predicted if coupling is neglected. A series of experiments to study current-deflection coupling were performed using the Fusion Electromagnetic Inductance Experiment (FELIX) facility at Argonne National Laboratory. Magnetic damping and magnetic stiffness resulting from the coupling are discussed, and analytical expressions for induced eddy current and rigid body rotation in the FELIX plate experiment are compared with the experimental results. Predictions for the degree of coupling based on various parameters are made using the analytical model

  9. Tokamak Simulation Code modeling of NSTX

    International Nuclear Information System (INIS)

    Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.

    2000-01-01

    The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption

  10. Sliding Mode Control of a Tokamak Transformer

    Energy Technology Data Exchange (ETDEWEB)

    Romero, J. A.; Coda, S.; Felici, F.; Moret, J. M.; Paley, J.; Sevillano, G.; Garrido, I.; Le, H. B.

    2012-06-08

    A novel inductive control system for a tokamak transformer is described. The system uses the flux change provided by the transformer primary coil to control the electric current and the internal inductance of the secondary plasma circuit load. The internal inductance control is used to regulate the slow flux penetration in the highly conductive plasma due to the skin effect, providing first-order control over the shape of the plasma current density profile. Inferred loop voltages at specific locations inside the plasma are included in a state feedback structure to improve controller performance. Experimental tests have shown that the plasma internal inductance can be controlled inductively for a whole pulse starting just 30ms after plasma breakdown. The details of the control system design are presented, including the transformer model, observer algorithms and controller design. (Author) 67 refs.

  11. Thresholds of ion turbulence in tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Laurent, L.; Mourgues, F.; Roubin, J.P.; Samain, A.; Zou, X.L.

    1991-01-01

    The linear thresholds of ionic turbulence are numerically calculated for the Tokamaks JET and TORE SUPRA. It is proved that the stability domain at η i >0 is determined by trapped ion modes and is characterized by η i ≥1 and a threshold L Ti /R of order (0.2/0.3)/(1+T i /T e ). The latter value is significantly smaller than what has been previously predicted. Experimental temperature profiles in heated discharges are usually marginal with respect to this criterium. It is also shown that the eigenmodes are low frequency, low wavenumber ballooned modes, which may produce a very large transport once the threshold ion temperature gradient is reached

  12. Gas Fuelling System for SST-1Tokamak

    Science.gov (United States)

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

    2017-04-01

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

  13. Auxiliary radiofrequency heating of tokamaks, Task 3

    International Nuclear Information System (INIS)

    Scharer, J.E.

    1991-07-01

    The research performed under this grant during the past three years has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling and heating issues: efficient coupling during the L- to H-mode transition by analysis and computer simulation of ICRF antennas edge plasma profiles; analysis of both dielectric-filled waveguide and coil ICRF antenna coupling to plasma edge profiles; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results; ICRF full-wave field solutions, power conservation and heating analyses; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report. 15 refs

  14. Fractional variational problems and particle in cell gyrokinetic simulations with fuzzy logic approach for tokamaks

    Directory of Open Access Journals (Sweden)

    Rastović Danilo

    2009-01-01

    Full Text Available In earlier Rastovic's papers [1] and [2], the effort was given to analyze the stochastic control of tokamaks. In this paper, the deterministic control of tokamak turbulence is investigated via fractional variational calculus, particle in cell simulations, and fuzzy logic methods. Fractional integrals can be considered as approximations of integrals on fractals. The turbulent media could be of the fractal structure and the corresponding equations should be changed to include the fractal features of the media.

  15. Lithium capillary porous system behavior as PFM in FTU Tokamak experiments

    International Nuclear Information System (INIS)

    Apichela, M.L.; Mazzitelli, G.; Lyublinski, I.E.; Lazarev, V.; Mirnov, S.; Vertkov, A.

    2007-01-01

    Full text of publication follows: Liquid lithium use on the base of capillary porous systems (CPS) application as plasma facing material (PFM) of tokamaks is advanced way to solve the problems of plasma contamination with high Z impurity, PFM degradation and tritium retention. In frame of joint program between ENEA (Italy) and FSUE 'Red Star' and TRINITI (RF) started at the end of 2005 die test of liquid lithium limiter (LLL) with CPS in a high field, medium size, carbon free tokamak FTU have been performed successfully. The LLL has been inserted in ohmic plasma discharges and at additional heating with LH and ECR at power levels in the MW range without any particular problem (BT = 6 T, Ip = 0.5- 0.9 MA, n e = 0.2 -2.6x10 20 m -3 , t = 1.5 s, P∼ 2-5 MW/m 2 at a normal discharge). The behavior of lithium CPS based on stainless steel wire mesh and its surface modification in normal discharges and at disruptions has been studied. Results of microscopic analyses of CPS structure after experimental campaigns are presented. The possibility to withstand heat load exceeding 5 MW/m 2 without damage, lithium surface renewal, mechanical stabilization of liquid lithium against MHD forces have been confirmed. Application of W, Mo as the base material and possible structure types of CPS have been considered for operating parameters improvement of long-living plasma facing components. (authors)

  16. Models for Predicting Boundary Conditions in L-Mode Tokamak Plasma

    International Nuclear Information System (INIS)

    Siriwitpreecha, A.; Onjun, T.; Suwanna, S.; Poolyarat, N.; Picha, R.

    2009-07-01

    Full text: The models for predicting temperature and density of ions and electrons at boundary conditions in L-mode tokamak plasma are developed using an empirical approach and optimized against the experimental data obtained from the latest public version of the International Pedestal Database (version 3.2). It is assumed that the temperature and density at boundary of L-mode plasma are functions of engineering parameters such as plasma current, toroidal magnetic field, total heating power, line averaged density, hydrogenic particle mass (A H ), major radius, minor radius, and elongation at the separatrix. Multiple regression analysis is carried out for these parameters with 86 data points in L-mode from Aug (61) and JT60U (25). The RMSE of temperature and density at boundary of L-mode plasma are found to be 24.41% and 18.81%, respectively. These boundary models are implemented in BALDUR code, which will be used to simulate the L-mode plasma in the tokamak

  17. LIDAR Thomson scattering for advanced tokamaks. Final report

    International Nuclear Information System (INIS)

    Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

    1996-01-01

    The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured

  18. Characterization of the Novillo Tokamak in main discharge regime; Caracterizacion del Tokamak Novillo en regimen de descarga principal

    Energy Technology Data Exchange (ETDEWEB)

    Lopez C, R.; Melendez L, L.; Chavez A, E.; Colunga S, S.; Valencia A, R.; Gaytan G, E

    1992-07-15

    The analytical procedure to carry out the establishment of the discharge in a Tokamak including: a) Ionization, b) Diffusion losses, recombination, union, drift speed, spurious fields, and c) Electric field is presented. In an experimental way a procedure settles down by means of which it is characterized the plasma, specially a new characteristic discharge parameter is settled down and it is the plasma current by the duration of the (I{sub p}t) discharge. (Author)

  19. Edge Plasma Response to Non-Axisymmetric Fields in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Ferraro, N. M.; Lao, L. L.; Buttery, R. J.; Evans, T. E.; Snyder, P. B.; Wade, M.R., E-mail: ferraro@fusion.gat.com [General Atomics, San Diego (United States); Moyer, R. A.; Orlov, D. M. [University of California San Diego, La Jolla (United States); Lanctot, M. J. [Lawrence Livermore National Laboratory, Livermore (United States)

    2012-09-15

    Full text: The application of non-axisymmetric fields is found to have significant effects on the transport and stability of H-mode tokamak plasmas. These effects include dramatic changes in rotation and particle transport, and may lead to the partial or complete suppression of edge-localized modes (ELMs) under some circumstances. The physical mechanism underlying these effects is presently not well understood, in large part because the response of the plasma to non- axisymmetric fields is significant and complex. Here, recent advances in modeling the plasma response to non-axisymmetric fields are discussed. Calculations using a resistive two-fluid model in diverted toroidal geometry confirm the special role of the perpendicular electron velocity in suppressing the formation of islands in the plasma. The possibility that islands form near the top of the pedestal, where the zero-crossing of the perpendicular electron velocity may coincide with a mode-rational surface, is explored, and the implications for ELM suppression are discussed. Modeling results are compared with empirical data. It is shown that numerical modeling is successful in reproducing some experimentally observed effects of applied non-axisymmetric fields on the edge temperature and density profiles. The numerical model self-consistently includes the plasma, separatrix, and scrape-off layer. Rotation and diamagnetic effects are also included self-consistently. Solutions are calculated using the M3D-C1 extended-MHD code. (and others)

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

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-01-01

    This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies