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Sample records for ht-7 tokamak

  1. Experiments in the HT-7 Superconducting Tokamak

    International Nuclear Information System (INIS)

    Wan Baonian

    2002-01-01

    The HT-7 tokamak experiment research has made important progress. The main efforts have dealt with quasi-steady-state operation, lower-hybrid (LH) current drive (LHCD), plasma heating with ion cyclotron range of frequencies (ICRF), ion Bernstein waves (IBWs), fueling with pellets and supersonic molecular beams, first-wall conditioning techniques, and plasma and wall interaction. Plasma parameters in the experiments were much improved; for example, n e = 6.5 x 10 19 m -3 , and a plasma pulse length of >10 s was achieved. ICRF boronization and conditioning resulted in Z eff close to unity. Steady-state full LH wave current drive has been achieved for >3 s. LHCD rampup and recharge have also been demonstrated. The best [eta] CD exp of 10 19 m -2 A/W is achieved. Quasi-steady-state H-mode-like plasmas with a density close to the Greenwald limit were obtained by LHCD, where energy confinement time was nearly five times longer than in the ohmic case. The synergy between the IBW, pellet, and LHCD was investigated. New doped graphite as limiter material and ferritic steel used to reduce the ripples have been developed. Research on the mechanism of microturbulence has been extensively carried out experimentally

  2. The reconstruction of HT-7 superconducting tokamak and the present status of HT-7U project

    International Nuclear Information System (INIS)

    Weng, P.D.

    2000-01-01

    The first Chinese superconducting tokamak HT-7 was reconstructed from T-7. The main purposes of reconstruction are to improve the accessibility of the device and to provide a possibility of long pulse operation with high performance. The reconstruction has been done successfully. The HT-7U project has been approved and funded as a National Project, the engineering design and R and D are on the way. (author)

  3. Vacuum physics analysis of HT-7 superconducting tokamak pump limiter

    International Nuclear Information System (INIS)

    Hu Jiansheng; Li Chengfu; He Yexi

    1998-10-01

    The pump limiter is analysed with HT-7 superconducting tokamak parameter and the pump limiter construction. The particle exhaust of the pump limiter can be to achieve about 7.7%. So the pump limiter can be applied in the HT-7 device and will make good affection in plasma discharge

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  5. Design of the Cryostat for HT-7U Superconducting Tokamak

    Science.gov (United States)

    Yu, Jie; Wu, Song-tao; Song, Yun-tao; Weng, Pei-de

    2002-06-01

    The cryostat of HT-7U tokamak is a large vacuum vessel surrounding the entire basic machine with a cylindrical shell, a dished top and a flat bottom. The main function of HT-7U cryostat is to provide a thermal barrier between an ambient temperature test hall and a liquid helium-cooled superconducting magnet. The loads applied to the cryostat are from sources of vacuum pressure, dead weight, seismic events and electromagnetic forces originated by eddy currents. It also provides feed-through penetrations for all the connecting elements inside and outside the cryostat. The main material selected for the cryostat is stainless steel 304L. The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out by using a finite element code. Stress analysis results show that the maximum stress intensity was below the allowable value. In this paper, the structural analyses and design of HT-7U cryostat are emphasized.

  6. High density operation on the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Xiang Gao

    2000-01-01

    The structure of the operation region has been studied in the HT-7 superconducting tokamak, and progress on the extension of the HT-7 ohmic discharge operation region is reported. A density corresponding to 1.2 times the Greenwald limit was achieved by RF boronization. The density limit appears to be connected to the impurity content and the edge parameters, so the best results are obtained with very clean plasmas and peaked electron density profiles. The peaking factors of electron density profiles for different current and line averaged densities were observed. The density behaviour and the fuelling efficiency for gas puffing (20-30%), pellet injection (70-80%) and molecular beam injection (40-50%) were studied. The core crash sawteeth and MHD behaviour, which were induced by an injected pellet, were observed and the events correlated with the change of current profile and reversed magnetic shear. The MARFE phenomena on HT-7 are summarized. The best correlation has been found between the total input ohmic power and the product of the edge line averaged density and Z eff . HT-7 could be easily operated in the high density region MARFE-free using RF boronization. (author)

  7. Electron Heating of LHCD Plasma in HT-7 Tokamak

    International Nuclear Information System (INIS)

    Ding Yonghua; Wan Baonian; Lin Shiyao; Chen Zhongyong; Hu Xiwei; Shi Yuejiang; Hu Liqun; Kong Wei; Zhang Xiaoqing

    2006-01-01

    Electron heating via lower hybrid current drive (LHCD) has been investigated in HT-7 superconducting tokamak. Experiments show that the central electron temperature T e0 , the volume averaged electron temperature e > and the peaking factor of the electron temperature Q Te = T e0 / e > increase with the lower hybrid wave (LHW) power. Simultaneously the electron heating efficiency and the electron temperature as the function of the central line-averaged electron density (n e ) and the plasma current (I p ) have also been investigated. The experimental results are in a good agreement with those of the classical collision theory and the LHW power deposition theory

  8. Ion Bernstein wave heating experiments in HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Zhao Yanping

    2005-01-01

    Ion Bernstein Wave (IBW) experiments have been carried out in recent years in the HT-7 superconducting Tokamak. The electron heating experiment has been concentrated on deuterium plasma with an injecting RF power up to 350 kw. The globe heating and localized heating can be seen clearly by controlling the ICRF resonance layer's position. On-axis and off-axis electron heating have been realized by properly setting the target plasma parameters. Experimental results show that the maximum increment in electron temperature has been more than 1 keV, the electron temperature profile has been modified by IBW under different plasma conditions, and both energy and particle confinement improvements have been obtained. (author)

  9. Propagation of Blob in boundary of HT-7 tokamak

    International Nuclear Information System (INIS)

    Yan Ning; Zhang Wei; Chang Jiafeng; Ming Tingfeng; Ding Siye

    2011-01-01

    Intermittent characteristics of turbulence induced by coherent structures (Blob) are manifested clearly on the Langmuir probe signal of HT-7 tokamak. With conditional analysis, asymmetric characteristics of the intermittent bursts are demonstrated. The parameter of plasma inside the Blob is larger than the background plasma parameter. Due to the radial propagation of the coherent structures, the particle density and temperature profiles in the scrape-off layer (SOL) are non-exponential and flat away from the last closed flux surface (LCFS). Around the LCFS, large burst fluctuations are responsible for about 50% of the total transport. These experimental finds may imply that the coherent structures are distorted by the developed shear flow in the shear layer. In SOL region, the coherent structures propagate in the direction of ion diamagnetic drift. (authors)

  10. Enhanced lower hybrid current drive experiments on HT-7 tokamak

    International Nuclear Information System (INIS)

    Shen Weici; Kuang Guangli; Liu Yuexiu; Ding Bojiang; Shi Yaojiang

    2003-01-01

    Effective Lower Hybrid Current Driving (LHCD) and improved confinement experiments in higher plasma parameters (I p >200 kA, n e >2 x 10 13 cm -3 , T e ≥1 keV) have been curried out in optimized LH wave spectrum and plasma parameters in HT-7 superconducting tokamak. The dependence of current driving efficiency on LH power spectrum, plasma density (anti n e ) and toroidal magnetic field B T has been obtained under optimal conditions. A good CD efficiency was obtained at higher plasma current and higher electron density. The improvement of the energy confinement time is accompanied with the increase in line averaged electron density, and in ion and electron temperatures. The highest current driving efficiency reached η CD =I p (anti n e )R/P RF ≅1.05 x 10 19 Am -2 /W. Wave-plasma coupling was sustained in a good state and the reflective coefficient was less than 5%. The experiments have also demonstrated the ability of LH wave in the start-up and ramp-up of the plasma current. The measurement of the temporal distribution of plasma parameter shows that lower hybrid leads to a broader profile in plasma parameter. The LH power deposition profile and the plasma current density profile were modeled with a 2D Fokker-Planck code corresponding to the evolution process of the hard x-ray detector array

  11. Upgraded data service system for HT-7 tokamak

    International Nuclear Information System (INIS)

    Qu Lianzheng; Luo Jiarong; Wei Peijie; Li Guiming; Cheng Ting; Qi Na

    2005-01-01

    A data service system plays an indispensable role in HT-7 Tokamak experiment. Since the former system doesn't provide the function of timely data procession and analysis, and all client software are based on Windows, it can't fulfill virtual fusion laboratory for remote researchers. Therefore, a new system which is simplified by three kinds of data servers and one data analysis and visualization software tool has been developed. The data servers include a data acquisition server based on file system, an MDSplus server used as the central repository for analysis data, and a web server. Users who prefer the convenience of application that can be run in a Web Browser can easily access the experiment data without knowing X-Windows. In order to adjust instruments to control experiment the operators need to plot data duly as soon as they are gathered. To satisfy their requirement, an upgraded data analysis and visualization software GT-7 is developed. It not only makes 2D data visualization more efficient, but also it can be capable of processing, analyzing and displaying interactive 2D and 3D graph of raw, analyzed data by the format of ASCII, LZO and MDSplus. (authors)

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

    International Nuclear Information System (INIS)

    Wan Yuanxi

    2002-01-01

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

  13. Method of calculating the safety factor profile on the HT-7 tokamak

    International Nuclear Information System (INIS)

    Zhang Xianmei; Lu Yuancheng; Wan Baonian

    2001-01-01

    A method of calculating the safety factor profile on the HT-7 tokamak has been described. It is derived from Maxwell's equations, among which the authors mainly use two of them: one is the magnetic field diffusion equation, and the other is Ampere's Law. This method can be also used to evaluate the safety factor on other devices with a circular cross sections. It is helpful to the study of the plasma MHD behavior on the HT-7 tokamak

  14. Plasma recovery after various events in HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Hu, J.S.; Li, J.G.

    2008-01-01

    Normal plasma recoveries after various events, such as after shutdown, various boronization, oxidation and large air leak, were investigated in the 2007 campaign of HT-7. Plasma recoveries, including disruptive plasmas, would depend on the wall status, such as impurities content and hydrogen retention. After shutdown or air leak, impurities made plasma recovery very difficult. After boronization, plasma recoveries would depend on the procedures of the boronization (C 2 B 10 H 12 ). After oxidation, boronization would effectively suppress impurities and would be beneficial for plasma recovery. ICRF cleanings in various working gases, such as He and D 2 , would be useful for impurities and hydrogen removal. This research is important for effective operation of HT-7 and would be useful for EAST and ITER operations.

  15. Mitigation of current quench by runaway electrons in LHCD discharges in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Lu, H.W.; Hu, L.Q.; Lin, S.Y.; Zhong, G.Q.

    2009-01-01

    Production of runaway electrons during a major disruption has been observed in HT-7 Tokamak. The runaway current plateaus, which can carry part of the pre-disruptive current, are observed in lower-hybrid current drive (LHCD) limiter discharges. It is found that the runaway current can mitigate the disruptions effectively. Detailed observations are presented on the runaway electrons generated following disruptions in the HT-7 tokamak with carbon limited discharges. The results indicate that the magnetic oscillations play an important role in the activity of runaway electrons in disruption. (author)

  16. The compression algorithm for the data acquisition system in HT-7 tokamak

    International Nuclear Information System (INIS)

    Zhu Lin; Luo Jiarong; Li Guiming; Yue Dongli

    2003-01-01

    HT-7 superconducting tokamak in the Institute of Plasma Physics of the Chinese Academy of Sciences is an experimental device for fusion research in China. The main task of the data acquisition system of HT-7 is to acquire, store, analyze and index the data. The volume of the data is nearly up to hundreds of million bytes. Besides the hardware and software support, a great capacity of data storage, process and transfer is a more important problem. To deal with this problem, the key technology is data compression algorithm. In the paper, the data format in HT-7 is introduced first, then the data compression algorithm, LZO, being a kind of portable lossless data compression algorithm with ANSIC, is analyzed. This compression algorithm, which fits well with the data acquisition and distribution in the nuclear fusion experiment, offers a pretty fast compression and extremely fast decompression. At last the performance evaluation of LZO application in HT-7 is given

  17. 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

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

    International Nuclear Information System (INIS)

    Zhang Jian; Liu Baohua; Ding Tonghai; Du Shaowu

    2006-01-01

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

  19. Investigation of the LH wave energy conversion and current drive efficiency in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Chen, Z.Y.; Wan, B.N.; Shi, Y.J.; Lin, S.Y.; Hu, L.Q.; Asif, M.

    2005-01-01

    Lower hybrid current drive (LHCD) plasmas in the presence of DC electric filed have been investigated based on Karney-Fisch theory in the HT-7 tokamak. The relatively small scatter in the experimental data with various values of waveguide phasing and lower hybrid power, when plotted in the Karney-Fisch diagram, confirms that a reasonable theoretical interpretation is possible for the HT-7 data. The full non-inductively current drive efficiencies are obtained by fitting the experimental data to the theoretical curve. The efficiency strongly depends on the lower hybrid wave phase velocity

  20. Density Modulation Experiments to Determine Particle Transport Coefficients on HT-7 Tokamak

    International Nuclear Information System (INIS)

    Jie Yinxian; Gao Xiang; Tanaka, K; Sakamoto, R; Toi, K; Liu Haiqing; Gao Li; Asif, M; Liu Jin; Xu Qiang; Tong Xingde; Cheng Yongfei

    2006-01-01

    The particle diffusion coefficient and the convection velocity were studied based on the density modulation using D 2 gas puffing on the HT-7 tokamak. The density was measured by a five-channel FIR interferometer. The density modulation amplitude was 10% of the central chord averaged background density and the modulation frequency was 10 Hz in the experiments. The particle diffusion coefficient (D) and the convection velocity (V) were obtained for different background plasmas with the central chord averaged density e > = 1.5x10 19 m -3 and 3.0x10 19 m -3 respectively. It was observed that the influence of density modulation on the main plasma parameters was very weak. This technology is expected to be useful for the analysis of LHW and IBW heated plasmas on HT-7 tokamak in the near future

  1. Fusion neutron yield and flux calculation on HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Fu Yanzhang; Zhu Yubao; Chen Juequan

    2006-01-01

    Neutron yield and flux have been numerically estimated on HT-7 tokamak. The total fusion neutron yield and neutron flux distribution on different positions and azimuth angles of the device are presented. Analyses on the errors induced by ion temperature and density distribution factors are given in detail. The results of the calculations provide a useful database for neutron diagnostics and neutron radiation protection. (authors)

  2. Preliminary observation on coordination of pellet injection and ion Bernstein wave on a HT-7 tokamak

    International Nuclear Information System (INIS)

    Yang Yu; Zhao Yanping; Li Jiangang; Wan Baonian; Luo Jiarong; Gu Xuemao

    2002-01-01

    A pellet injection (PI) experiment was performed during the application of the ion Bernstein wave on a HT-7 tokamak. A preliminary coordination effect was observed. With a lower wave power, shortly after PI, the coupling of the wave was enhanced, and the particle confinement was improved. With higher power, off-axis heating for 15% at about α/3 in the low field side was observed

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  4. Improved plasma confinement by modulated toroidal current on HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Mao Jianshan; Zhao Junyu; Shen Biao; Luo Jiarong

    2004-01-01

    The improved confinement phase was observed during modulating toroidal current on the Hefei superconducting Tokamak-7 (HT-7). This improved plasma confinement phase is characterized by suppressing magnetohydrodynamic (MHD) instabilities effectively, thus increased the central line averaged electron density and the central electron temperature about 33%, out-put steeper density profiles, and reduced hydrogen radiation from the edge as well. The global energy confinement time was increased by 27%-45%; The impurity radiation was reduced by modulation of plasma toroidal current; particle confinement time was increased about two times; a stronger radial negative electric field formed inside the limiter. The radial electric field during modulating current was calculated and disscused. (authors)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  6. Lower hybrid current drive experiments with graphite limiters in the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Liu, J.; Gao, X.; Hu, L.Q.; Asif, M.; Chen, Z.Y.; Ding, B.J.; Zhou, Q.; Liu, H.Q.; Jie, Y.X.; Kong, W.; Lin, S.Y.; Ding, Y.H.; Gao, L.; Xu, Q.

    2006-01-01

    Recent progress of lower hybrid (LH) experiments with new graphite limiters configuration in the HT-7 tokamak is presented. The lower hybrid current drive (LHCD) efficiency can be determined by fitting based on experimental data. Improved particle confinement was observed via LHCD (P LHW >300 kW) characterized by the particle confinement time τ p increased about 1.56 times. It is found that runaways are suppressed during loop voltage is decreasing at the flat-top phase of LH discharges. The main limitations of pulse length are presented in long-pulse experiments with new limiter configuration

  7. Investigation on synergy of IBW and LHCD for integrated high performance operation in HT-7 tokamak

    International Nuclear Information System (INIS)

    Wan Baonian

    2002-01-01

    Control of the current density profile has been realized with off-axis current drive by LHW in the HT-7 tokamak predicted by a 2D FP code simulation and supported by measurements of a vertical HX array. IBW is explored to improve performance through heating electrons in the selected region. Strong synergy effect on driven current profile and increased driven efficiency was observed. Electron temperature shows an ITB-like profile with a significantly improved performance. Operation of IBW and LHCD synergetic discharges was optimized through moving the IBW resonant layer to maximize the plasma performance and to avoid the MHD activities. A variety of high performance discharges indicated by β N *H89=1∼ 4 was produced for several tens energy confinement times. This operation mode utilizing synergy effect of IBW and LHCD provide a new way to obtain steady-state operation in advanced tokamak scenario. (author)

  8. The resonance between runaway electrons and magnetic ripple in HT-7 Tokamak

    International Nuclear Information System (INIS)

    Zhou Ruijie; Hu Liqun; Lu Hongwei; Lin Shiyao; Zhong Guoqiang; Xu Ping; Zhang Jizong

    2011-01-01

    For suppressing the energy of runaway electrons in tokamak plasma, we analyzed the X-ray energy spectra by runaway electrons in different discharges of the HT-7 tokamak experiment performed in the autumn of 2009. The resonant phenomenon between runaway electrons and magnetic ripple was found. Although, the energy of runaway electrons in the plasma core can be as high as several tens of MeV, but when they are transported to the edge, the electron energy are limited to a certain range by resonance with the magnetic ripple of different harmonic numbers. The runaway electrons under high loop voltage resonate with low step magnetic perturbations, with high energy gain; whereas the runaway electrons under low loop voltage resonate with high level magnetic perturbations, with low energy gain. Using this mechanism, the energy of runaway electrons can be restricted to a low level, and this will significantly mitigate the damage effect on the equipment caused by runaway electrons. (authors)

  9. Investigation of slide-away discharges in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Lu Hongwei; Hu Liqun; Lin Shiyao; Zhong Guoqian; Zhou Ruijie

    2010-01-01

    In tokamak plasmas, the discharge will go into 'runaway' discharges if the density decays to the critical ones. The discharge will go into 'slide-away' discharges if the density reaches a lower level. The slide-away discharge is characterized by high confinement and lots of superthermal electrons which constitute a large part of plasma current. In HT-7 Tokamak, the slide-away discharges have been achieved by decreasing the plasma density. The relation ship between plasma current and the critical density of slide-away discharge was investigated. It was also found that the increase of density in slide-away discharge can make the confinement poor. And also, lots of superthermal electrons lost from the vacuum chamber. (authors)

  10. The development of joining doped graphite to copper for first wall application in HT-7 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Zhangjian, E-mail: zhouzhj@mater.ustb.edu.cn [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Zhong Zhihong [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Chen Junling [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Ge Changchun [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China)

    2010-12-15

    Two joining methods have been developed for joining carbon based plasma facing material to copper based heat sink material for the potential application in HT-7 and EAST tokamak. The first joining method is based on brazing technique by using a rapidly solidified foil-type Ti-Zr based amorphous filler with a melting temperature of 850 deg. C. The other joining method is direct active metal casting-casting the premixed powders of copper and active transition metals on the mechanical machined carbon surface directly. SEM observations demonstrate high quality of joining surface for both joints. The brazing technique is more promising for fabrication joint with larger size compared with the direct active alloy casting method. High heat flux test using an e-beam device was performed on the actively cooled C/Cu joint fabricated by brazing method. There has no damage occurred on the joint after heat loading at 6 MW/m{sup 2}.

  11. Modification of boundary fluctuations by LHCD in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Song Mei; Wan Baonian; Xu Guosheng; Ling Bili

    2003-01-01

    Measurements of boundary fluctuations and fluctuation driven electron fluxes have been performed in ohmic and lower hybrid current drive enhanced confinement plasma using a graphite Langmuir probe array on HT-7 tokamak. The fluctuations are significantly suppressed and the turbulent fluxes are remarkably depressed in the enhanced plasma. We characterized the statistical properties of fluctuations and the particle flux and found a non-Gaussian character in the whole scrape-off layer with minimum deviations from Gaussian in the proximity of the velocity shear layer in ohmic plasma. In the enhanced plasma the deviations in the boundary region are all reduces obviously. The fluctuations and induced electron fluxes show sporadic bursts asymmetric in time and the asymmetry is remarkably weakened in the lower hybrid current driving (LHCD) phase. The results suggest a coupling between the statistical behaviour of fluctuations and the turbulent flow

  12. Removal of particles by ICRF cleaning in HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Hu Jiansheng; Li Jiangang; Zhang Shouyin; Gu Xuemao; Zhang Xiaodong; Zhao Yanping; Gong Xianzu; Kuang Guangli; Li Chengfu; Luo Jiarong; Wang Xiaoming; Gao Xiang; Wan Baonian; Xie Jikang; Wan Yuanxi

    2001-01-01

    The ICRF (Ion Cyclotron Range Frequency) cleaning technique has been used as a routine wall cleaning method in the HT-7 superconducting tokamak. In a wide range of toroidal field, the removal rate of residual gas by ICRF cleaning was about twenty times higher than that of glow discharge cleaning (GDC). At different gas pressure and RF power levels, the ICRF cleaning is studied carefully. A good impurity cleaning effect and a very high hydrogen removal rate were obtained. The removal rate of hydrogen by 5 kW ICRF cleaning achieved was 1.6 x 10 -5 Torr.l/s. And the relationships among pressure P, outgassing rate Q, atomic layers L absorbed on surface and the cleaning mode were discussed briefly

  13. 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.

  14. Optimization of the protective energy removal parameters for tokamak HT7-U superconducting magnets

    Energy Technology Data Exchange (ETDEWEB)

    Khvostenko, P.P.; Chudnovsky, A.N.; Posadsky, I.A. [RRC ' Kurchatov Inst.' , Nuclear Fusion Inst., Moscow (Russian Federation); Bi, Y.F.; Cheng, S.M.; He, Y.X. [Academia Sinica, Hefei, Anhui (China). Inst. of Plasma Physics

    1998-07-01

    The design of the HT-7U superconducting tokamak is in progress now. The design incorporates superconducting magnets of the toroidal field and poloidal field systems. Toroidal field system consists of 16 D-shape coils and poloidal field system consists of 12 coils. All coils will be use NbTi/Cu cable-in-conduit conductor cooled with forced-flow supercritical helium at 4.5 K, 4 Bar. Quench in the superconducting magnets is accompanied byconversion of the stored magnetic field energy into a thermal one which is spent on heating of both the coil part which made transition into a normal state and dump resistors. A non-uniform heating of the coil part results in the emergence of thermomechanical stresses which can cause its destruction. The protective removal of a current is realized to prevent the coil destruction at the emergence of the quench. In that case, the faster the current removal occurs, the less the coil heating is. On the other hand, the current removal rate should not be too high in order to avoid an electric breakdown by the excited inductive voltage. Optimization of the protective energy removal parameters both for TF and PF superconducting magnets is presented. (author)

  15. Optimization of the protective energy removal parameters for tokamak HT7-U superconducting magnets

    International Nuclear Information System (INIS)

    Khvostenko, P.P.; Chudnovsky, A.N.; Posadsky, I.A.; Bi, Y.F.; Cheng, S.M.; He, Y.X.

    1998-01-01

    The design of the HT-7U superconducting tokamak is in progress now. The design incorporates superconducting magnets of the toroidal field and poloidal field systems. Toroidal field system consists of 16 D-shape coils and poloidal field system consists of 12 coils. All coils will be use NbTi/Cu cable-in-conduit conductor cooled with forced-flow supercritical helium at 4.5 K, 4 Bar. Quench in the superconducting magnets is accompanied by conversion of the stored magnetic field energy into a thermal one which is spent on heating of both the coil part which made transition into a normal state and dump resistors. A non-uniform heating of the coil part results in the emergence of thermomechanical stresses which can cause its destruction. The protective removal of a current is realized to prevent the coil destruction at the emergence of the quench. In that case, the faster the current removal occurs, the less the coil heating is. On the other hand, the current removal rate should not be too high in order to avoid an electric breakdown by the excited inductive voltage. Optimization of the protective energy removal parameters both for TF and PF superconducting magnets is presented. (author)

  16. Study of LHW and IBW synergy experiment on the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Gao, X.

    2001-01-01

    A successful experiment on lower hybrid wave (LHW) and ion Bernstein wave (IBW) synergy has been carried out in the HT-7 superconducting tokamak. With 500 kW of LHW heating power and 200 kW of injected IBW power, it is observed that the ion temperature increases from 500 eV to about 850 eV, the electron temperature increases from 800 eV to 1.2 keV, and the averaged electron density increases from 0.9x10 19 m -3 to 2.6x10 19 m -3 . The plasma parameters were obviously enhanced by means of the LHW and IBW heating and their synergy. The charge-exchange spectra of the neutral particle analysis (NPA) diagnostics data clearly showed that the high-energy ion tail which was produced by the LHW was decreased by the synergy with the IBW, and the bulk ion temperature was increased. The mechanism of the LHW and IBW synergy effect is discussed. (author)

  17. Improved confinement mode induced by the MARFE on the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Gao, X.; Zhao, Y.P.; Luo, J.R.; Jie, Y.X.; Gong, X.Z.; Wan, B.N.; Li, J.G.; Yin, F.X.; Kuang, G.L.; Zhang, X.D.; Zhang, S.Y.; Qiu, N.; Liu, X.N.; Zhao, J.Y.; Yang, Y.; Bao, Y.; Lin, B.L.; Wu, Z.W.; Li, Y.D.; Xu, Y.H.; Yang, K.; Wang, G.X.; Ye, W.W.; Chen, L.; Shi, Y.J.; Song, M.; Zhang, X.M.; Qin, P.J.; Gu, X.M.; Cui, N.Z.; Fan, H.Y.; Liu, S.X.; Chen, Y.F.; Hu, L.Q.; Hu, J.S.; Xia, C.Y.; Ruan, H.L.; Tong, X.D.; Mao, J.S.; Xie, J.K.; Wan, Y.X.

    1999-01-01

    In the HT-7 superconducting tokamak, the onset of a multifaceted asymmetric radiation from the edge (MARFE) usually occurs in the early ohmic discharges of each experimental campaign before wall conditioning. The occurrence and location of a MARFE is identified by different diagnostic systems. An improved confinement mode plasma which was induced by the MARFE is observed, and the global particle confinement time increases 1.9 times. The relaxation time between the MARFE event trigger and the L-H transition is about 1.4 ms, the following L-H transition time is 1.9 ms, and the improved confinement mode phase is maintained for about 40 ms. The MARFE cools the plasma edge, and the electron density profile is observed to become more narrow and peaked. The occurrence of a MARFE is strongly correlated with Z eff but not with the density, and it always occurs at Z eff = 3-8 ohmic discharges. In the case of a good wall condition (Z eff = 1-2), the onset of MARFEs has not been observed before reaching the Greenwald density limit. (author)

  18. Electron temperature fluctuation in the HT-7 tokamak plasma observed by electron cyclotron emission imaging

    International Nuclear Information System (INIS)

    Xiao-Yuan, Xu; Jun, Wang; Yi, Yu; Yi-Zhi, Wen; Chang-Xuan, Yu; Wan-Dong, Liu; Bao-Nian, Wan; Xiang, Gao; Luhmann, N. C.; Domier, C. W.; Wang, Jian; Xia, Z. G.; Shen, Zuowei

    2009-01-01

    The fluctuation of the electron temperature has been measured by using the electron cyclotron emission imaging in the Hefei Tokamak-7 (HT-7) plasma. The electron temperature fluctuation with a broadband spectrum shows that it propagates in the electron diamagnetic drift direction, and the mean poloidal wave-number k-bar θ is calculated to be about 1.58 cm −1 , or k-bar θρ s thickapprox 0.34. It indicates that the fluctuation should come from the electron drift wave turbulence. The linear global scaling of the electron temperature fluctuation with the gradient of electron temperature is consistent with the mixing length scale qualitatively. Evolution of spectrum of the fluctuation during the sawtooth oscillation phases is investigated, and the fluctuation is found to increase with the gradient of electron temperature increasing during most phases of the sawtooth oscillation. The results indicate that the electron temperature gradient is probably the driver of the fluctuation enhancement. The steady heat flux driven by electron temperature fluctuation is estimated and compared with the results from power balance estimation. (fluids, plasmas and electric discharges)

  19. Hard X-Ray PHA System on the HT-7 Tokamak

    International Nuclear Information System (INIS)

    Lin Shiyao; Shi Yuejiang; Wan Baonian; Chen Zhongyong; Hu Liqun

    2006-01-01

    A new hard X-ray pulse-height analysis (PHA) system has been established on HT-7 tokamak for long pulse steady-state operation. This PHA system consists of hard X-ray diagnostics and multi-channel analysers (MCA). The hard X-ray diagnostics consists of a vertical X-ray detector array (CdTe) and a horizontal X-ray detector array (NaI). The hard X-ray diagnostics can provide the profile of power deposition and the distribution function of fast electron during radio frequency (RF) current drive. The MCA system is the electronic part of the PHA system, which has been modularized and linked to PC through LAN. Each module of MCA can connect with 8 X-ray detectors. The embedded Ethernet adapter in the MCA module makes the data communication between PC and MCA very convenient. A computer can control several modules of MCA through certain software and a hub. The RAM in MCA can store 1024 or more spectra for each detector and therefore the PHA system can be applied in the long pulse discharge of several minutes

  20. Modification of boundary plasma behavior by Ion Bernstein Wave heating on HT-7 tokamak

    International Nuclear Information System (INIS)

    Xu Guoshen

    2002-01-01

    Cooperated with Fusion Research Center, the University of Texas at Austin, U.S.A. The boundary plasma behavior during Ion Bernstein Wave (IBW) heating was investigated using Langmuir probe arrays on HT-7 tokamak. The particle confinement improvement of over a factor of 2 was observed in 30 MHz IBW heated plasma with RF power > 120 kW. The strong de-correlation effect of fluctuations resulted in that the turbulent particle flux dropped more than an order of magnitude. In IBW heated plasma, an additional inward E r and associated poloidal ExB flows were produced, which could account for the additional poloidal velocity in the electron diamagnetic direction in the scrape-of layer (SOL). Three-wave nonlinear phase coupling increased evidently and low frequency fluctuations (about 5 kHz) were generated, which dominated the boundary turbulence during IBW heating. The 5/2-D resonant layer was located in the plasma edge region, which is found to be the mechanism underlying these phenomena. (author)

  1. Recycling behaviour during long pulse discharges after ICRF boronization in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Huang, J.; Wan, B.N.; Li, J.G.; Gong, X.Z.; Zhang, X.D.; Wu, Z.W.; Zhou, Q.

    2006-01-01

    The evolution of recycling behaviour has been investigated during long pulse discharges in the HT-7 tokamak after ICRF boronization (C 2 B 10 H 12 ) using the H/(H+D) ratio and the edge recycling coefficient R. After boronization, impurity reduction is observed, attributed to the fresh boron film, but the recycling coefficient can exceed unity due to a large amount of hydrogen absorbed in the film, leading to an uncontrollable density rise and discharge termination. When the H/(H+D) ratio was reduced to less than 25%, the electron density was easily controlled. The longest discharge, up to 240 s with central electron temperature T e (0) of about 1.0 keV and central electron density n e (0) of 0.8 x 10 19 m -3 , was achieved following boronization. After many discharges the effectiveness of boron film was weakened, and the density rise was correlated with an increase in both carbon and oxygen radiation which limited the duration of long pulse discharges

  2. Liquid lithium surface control and its effect on plasma performance in the HT-7 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, G.Z.; Ren, J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, J.S., E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, Z.; Yang, Q.X.; Li, J.G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zakharov, L.E. [Princeton University Plasma Physics Laboratory Princeton, NJ 08543 (United States); Ruzic, David N. [University of Illinois, Urbana, IL 61801 (United States)

    2014-12-15

    Highlights: • Strong interaction between plasma and Li would cause strong Li emission and lead to disruptive plasmas, and probable reasons were analyzed. • Serious Li would be emitted from the free statics surface mainly due to J × B force leading to plasma instable and disruptions. • CPS surface would partially suppress the emission and be beneficial for plasma operation. • Li emission from flowing LLLs on free surfaces on SS trenches and on SS plate were compared. - Abstract: Experiments with liquid lithium limiters (LLLs) have been successfully performed in HT-7 since 2009 and the effects of different limiter surface structures on the ejection of Li droplets have been studied and compared. The experiments have demonstrated that strong interaction between the plasma and the liquid surface can cause intense Li efflux in the form of ejected Li droplets – which can, in turn, lead to plasma disruptions. The details of the LLL plasma-facing surface were observed to be extremely important in determining performance. Five different LLLs were evaluated in this work: two types of static free-surface limiters and three types of flowing liquid Li (FLLL) structures. It has been demonstrated that a FLLL with a slowly flowing thin liquid Li film on vertical flow plate which was pre-treated with evaporated Li was much less susceptible to Li droplet ejection than any of the other structures tested in this work. It was further observed that the plasmas run against this type of limiter were reproducibly well-behaved. These results provide technical references for the design of FLLLs in future tokamaks so as to avoid strong Li ejection and to decrease disruptive plasmas.

  3. Modification of boundary plasma behavior by Ion Bernstein Wave heating on the HT-7 tokamak

    International Nuclear Information System (INIS)

    Xu, G.S.; Wan, B.N.; Song, M.; Ling, B.L.; Li, C.F.; Li, J.

    2003-01-01

    The boundary plasma behavior during Ion Bernstein Wave heating was investigated using Langmuir probe arrays on the HT-7 tokamak. A distinct weak turbulence regime was reproducibly observed in the 30 MHz IBW heated plasmas with RF power larger than 120 kW, which resulted in a particle confinement improvement of a factor of 2. The strong suppression and decorrelation effect of fluctuations resulted in the turbulent particle flux dropping by more than an order of magnitude in the plasma boundary region. An additional inward radial electric field and associated poloidal ExB flows were produced, which could account for the additional poloidal velocity in the electron diamagnetic direction at some radial locations of the boundary plasma. The electrostatic fluctuations were nearly completely decorrelated in the high frequency region and only low frequency fluctuations remained. The poloidal correlation was considerably reduced in the high poloidal wave number region and only the fluctuations with long poloidal wavelength remained. Three-wave nonlinear phase coupling between the whole frequency domain and the very low frequency region increased significantly in both the plasma edge and the SOL. Quite low frequency fluctuations (about 5 kHz) were generated, which dominated the boundary turbulence during IBW heating. Detailed analyses suggested that, when an IBW with a frequency of 30 MHz was launched into a plasma with the toroidal magnetic field between 1.75 T and 2.0 T, the ion cyclotron resonant layer of 5/2.D was located in the plasma edge region. The poloidal ExB sheared flows generated by IBW near this layer due to a ponderomotive interaction were found to be the mechanism underlying these phenomena. (author)

  4. Analysis of Electron Thermal Diffusivity and Bootstrap Current in Ohmically Heated Discharges after Boronization in the HT-7 Tokamak

    International Nuclear Information System (INIS)

    Zhang, X.M.; Wan, B.N.

    2005-01-01

    Significant improvements of plasma performance after ICRF boronization have been achieved in the full range of HT-7 operation parameters. Electron power balance is analyzed in the steady state ohmic discharges of the HT-7 tokamak. The ratio of the total radiation power to ohmic input power increases with increasing the central line-averaged electron density, but decreases with plasma current. It is obviously decreased after wall conditioning. Electron heat diffusivity χ e deduced from the power balance analysis is reduced throughout the main plasma after boronization. χ e decreases with increasing central line-averaged electron density in the parameter range of our study. After boronization, the plasma current profile is broadened and a higher current can be easily obtained on the HT-7 tokamak experiment. It is expected that the fact that the bootstrap current increases after boronization will explain these phenomena. After boronization, the plasma pressure gradient and the electron temperature near the boundary are larger than before, these factors influencing that the ratio of bootstrap current to total plasma current increases from several percent to above 10%

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. Continuous and real-time data acquisition system for superconducting tokamaks HT-7 and TRIAM-1M

    International Nuclear Information System (INIS)

    Wang, F.; Luo, J.R.; Nakamura, K.; Sato, K.N.; Hanada, K.; Sakamoto, M.; Idei, H.; Kawasaki, S.; Nakashima, H.

    2006-01-01

    Conventional data acquisition systems cannot deal with data acquisition for a long-time discharge of a nuclear fusion reactor. Thus, continuous data acquisition with a real-time data presentation during discharge must be developed. Two data acquisition systems, which include alternating CAMAC data acquisition and long-time PCI data acquisition, are designed for the long-time operation of HT-7 tokamak. Since an effective alternating mode is adopted, the alternating CAMAC data acquisition can accurately and continuously acquire data at a rate of 10 kHz. The acquired data is immediately transmitted to a data server and real-time results can be presented during the plasma discharge. As for the long-time PCI data acquisition, a special kind of PCI A/D card, which has a hard disk on board, is designed to collect data at a max speed of 200 kHz. Thus, the total sampling duration is only related to the capacity of the hard disk on board. These two types of data acquisitions were applied to HT-7 tokamak and a 250 s discharge was acquired. These data acquisition systems were also successfully demonstrated on a 2500 s plasma discharge on TRIAM-1M. This paper describes the two data acquisitions in detail

  7. Transient snakes in an ohmic plasma associated with a minor disruption in the HT-7 Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Mao, Songtao; Xu, Liqing; Hu, Liqun; Chen, Kaiyun [Chinese Academy of Sciences, Hefei (China)

    2014-05-15

    A transient burst (∼2 ms, an order of the fast-particle slowdown timescale) of a spontaneous snake is observed for the first time in a HT-7 heavy impurity ohmic plasma. The features of the low-Z impurity snake are presented. The flatten electron profile due to the heavy impurity reveals the formation of a large magnetic island. The foot of the impurity accumulation is consistent with the location of the transient snake. The strong frequency-chirping behaviors and the spatial structures of the snake are also presented.

  8. Heating and active control of profiles and transport by IBW in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Zhao Yanping; Wan Baonian; Li Jiangang

    2003-01-01

    Significant progress on Ion Bernstein Wave (IBW) heating and control of profiles has been obtained in HT-7. Both on-axis and off-axis electron heating with global peaked and local steep electron pressure profiles were realized if the position of the resonant layer was selected to be plasma far from the plasma edge region. Reduction of electron heat transport has been observed from sawtooth heat pulse propagation. Improvement of both particle and energy confinement was slight in the on-axis and considerable in the off-axis heating cases. The improved confinement in off-axis heating mode may be due to the extension of the high performance plasma volume caused by IBW. These studies demonstrate that IBWs are potentially a tool for active control of plasma profiles and transport. (author)

  9. Observation of magnetohydrodynamics instabilities in ion Bernstein wave and lower-hybrid-current driving synergetic discharges on HT-7 tokamak

    International Nuclear Information System (INIS)

    Mao Jianshan; Luo Jiarong; Shen Biao; Zhao Junyu; Hu Liqun; Zhu Yubao; Xu Guosheng; Asif, M.; Gao Xiang; Wan Baonian

    2004-01-01

    The normalized performance indicated by the product of β N H 89 >2 was achieved by a combination of the lower hybrid current driving (LHCD) and the ion Bernstein wave (IBW) heating in the HT-7 tokamak. More than 80% of the plasma current was sustained by the LHCD and the bootstrap current. Large edge pressure gradients were observed. The magnetohydrodynamic (MHD) instabilities were often driven to terminate the discharge or reduce the discharge performance, when the IBW resonant layer was near the rational surface. The resonant layer of the safety factor q=2 is located at 0.6 a with a=27 cm being the minor radius. The width of magnetic island (the poloidal mode number m=2) was about 2 cm. The plasma energy was reduced quickly by 30% by MHD instabilities. The behaviour of MHD instabilities is reported. A large sawtooth activity (m=1) was observed before inducing MHD (m=2)

  10. Numerical simulation and analysis for the baking out system of the HT-7U super-conducting tokamak device

    International Nuclear Information System (INIS)

    Song Yuntao

    2004-01-01

    It can provide an ultrahigh vacuum location for the plasma operation. In order to improve its vacuum degree and attain a high quality operation environment for plasma, it is very important to proceed 250 degree C baking out to clear the wall before the plasma operation. The paper firstly gives two kinds of structures for the baking of the vacuum vessel, in which one is the baking by electricity and another is baking by the nitrogen gas. Secondly based on the numerical simulation and analysis, some results have been attained such as the baking power, temperature field distribution and thermal stress for the vacuum vessel, which can provide some valuable theory basis for the engineering design and optimization of the baking system of the HT-7U vacuum vessel or other similar super-conducting tokamak devices

  11. Generation of sheared poloidal flows by electrostatic and magnetic Reynolds stress in the boundary plasma of HT-7 tokamak

    International Nuclear Information System (INIS)

    Xu, G.S.; Wan, B.N.; Li, J.

    2005-01-01

    The radial profiles of electrostatic and magnetic Reynolds stress (Maxwell stress) have been measured in the plasma boundary region of HT-7 tokamak. Experimental results show that the radial gradient of electrostatic Reynolds stress (ERS) changes sign across the last closed flux surface, and the neoclassical flow damping and the damping due to charge exchange processes are balanced by the radial gradient of ERS, which sustains the equilibrium sheared flow structure in a steady state. The contribution of magnetic Reynolds stress was found unimportant in a low β plasma. Detailed analyses indicate that the propagation properties of turbulence in radial and poloidal directions and the profiles of potential fluctuation level are responsible for the radial structure of ERS. (author)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  13. Multichannel heterodyne radiometers with fast-scanning backward-wave oscillators for ECE measurement on HT-7 tokamak

    International Nuclear Information System (INIS)

    Zhang, S.Y.; Poznyak, V.I.; Ploskirev, G.; Kalupin, D.; Wan, Y.X.; Xie, J.K.; Luo, J.R.; Li, J.G.; Gao, X.; Wan, B.N.; Zhang, X.D.; Wang, K.J.; Kuang, G.L.

    2001-01-01

    Two sets of fast-scanning heterodyne radiometer receiver systems employing backward-wave oscillators (BWOs) in 78-118 and 118-178 GHz were developed and installed for electron cyclotron emission (ECE) measurements on HT-7 superconducting tokamak. The double sideband (DSB) radiometer in 78-118 GHz measures 16 ECE frequency points with a scanning time period of 0.65 ms. The other radiometer in 118-178 GHz consists of one independent channel of DSB heterodyne receiver with intermediate frequency (IF) of 100-500 MHz and two channels of single sideband (SSB) heterodyne receiver that are sensitive to upper sideband and lower sideband individually; the IF frequency of the SSB channels are 1.5 GHz around the local oscillator frequencies with 1 GHz bandwidth. By employing a novel design, this unique radiometer measures 3 ECE frequency points at each of the 16 local oscillator frequency points in 118-178 GHz, and the full band can be swept in 0.65 ms period, thus the radiometer measures 48 ECE frequency points in 0.65 ms in principle. Each of the local oscillators' frequency points can be preset by program to meet specific physics interests. Horizontal view of ECE was installed to measure electron temperature profiles; vertically viewing optics along a perpendicular chord was also installed to study nonthermal ECE spectra. Preliminary measurement results were presented during ohmic and pellet injection plasmas

  14. A novel fast-scanning microwave heterodyne radiometer system for electron cyclotron emission measurements in the HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Zhang, S.Y.; Wan, Y.X.; Xie, J.K.; Luo, J.R.; Li, J.G.; Kuang, G.L.; Gao, X.; Zhang, X.D.; Wan, B.N.; Wang, K.J.; Mao, J.S.; Gong, X.Z.; Qin, P.J.

    2000-01-01

    Two sets of fast-scanning microwave heterodyne radiometer receiver systems employing backward-wave oscillators in the 78-118 GHz and 118-178 GHz ranges were developed for electron cyclotron emission measurements (ECE) on the HT-7 superconducting tokamak. The double-sideband radiometer in the 78-118 GHz range measures 16 ECE frequency points with a scanning period of 0.65 ms. The novel design of the 2 mm fast-scanning heterodyne radiometer in the 118-178 GHz range enables the unique system to measure 48 ECE frequency points in 0.65 ms periodically. The plasma profile consistency in reproducible ohmic plasmas was used to relatively calibrate each channel by changing the toroidal magnetic field shot-by-shot. The absolute temperature value was obtained by a comparison with the results from the soft x-ray pulse height analysis measurements and Thomson scattering system. A preliminary temperature profile measurement result in pellet injection plasma is presented. (author)

  15. Effect of magnetic fluctuations on the confinement and dynamics of runaway electrons in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Zhou, R.J.; Hu, L.Q.; Li, E.Z.; Xu, M.; Zhong, G.Q.; Xu, L.Q.; Lin, S.Y.

    2013-01-01

    Experimental results in the HT-7 tokamak indicated significant losses of runaway electrons due to magnetic fluctuations, but the loss processes did not only rely on the fluctuation amplitude. Efficient radial runaway transport required that there were no more than small regions of the plasma volume in which there was very low transport of runaways. A radial runaway diffusion coefficient of D_r ≈ 10 m"2s"-"1 was derived for the loss processes, and diffusion coefficient near the resonant magnetic surfaces and shielding factor ϒ = 0.8 were deduced. Test particle equations were used to analyze the effect of magnetic fluctuations on runaway dynamics. It was found that the maximum energy that runaways can gain is very sensitive to the value of a_s. a_s = (0.28 - 0.33) was found for the loss processes in the experiment, and maximum runaway energy could be controlled in the range of E = (4 MeV - 6 MeV) in this case. Additionally, to control the maximum runaway energy below 5 MeV, the normalized electric field needed to be under a critical value D_a = 6.8, and the amplitude normalized magnetic fluctuations b tilde needed to be at least of the order of b tilde ≈ 3 x 10"-"5. (author)

  16. Experiment and operation of a LHCD-35 kV/2.8 MW/1000 s high-voltage power supply on HT-7 tokamak

    International Nuclear Information System (INIS)

    Huang Yiyun

    2002-01-01

    A-35 kV/2.8 MW/1000s high-voltage power supply (HVPS) for HT-7 superconducting tokamak has been built successfully. The HVPS is scheduled to run on a 2.45 GHz/1 MW lower hybrid current drive (LHCD) system of HT-7 superconducting tokamak before the set-up of HT-7 superconducting tokamak in 2003. The HVPS has a series of advantages such as good steady and dynamic response, logical computer program controlling the HVPS without any fault, operational panel and experimental board for data acquisition, which both are grounded distinctively in a normative way to protect the main body of HVPS along with its attached equipment from dangers. Electric power cables and other control cables are disposed reasonably, to prevent signals from magnetic interference and ensure the precision of signal transfer. The author introduced the experiment and operation of a 35 kV/2.8 MW/1000 s HVPS for 2.45 GHz/1 MW LHCD system. The reliability and feasibility of the HVPS has been demonstrated in comparison with experimental results of original design and simulation data

  17. 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

  18. Generation of runaway electrons during deterioration of lower hybrid power coupling in lower hybrid current drive plasmas in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Chen, Z Y; Ju, H J; Zhu, J X; Li, M; Cai, W D; Liang, H F; Wan, B N; Shi, Y J; Xu, H D

    2009-01-01

    Efficient coupling of lower hybrid (LH) power from the wave launcher to the plasma is a very important issue in lower hybrid current drive (LHCD) experiments. The large unbalanced reflections in the grill trigger the LH protection system, which will trip the power, resulting in the reduction of the coupled LH power. The generation of runaway electrons has been investigated in LHCD plasmas with deterioration of LH coupling in the HT-7 tokamak. The deterioration of LH coupling results in an increase of the loop voltage and a more energetic fast electron population. These two effects favor the generation of a runaway population. It is found that most of the fast electrons generated by LH waves through parallel electron Landau damping were converted into a runaway population through the acceleration from the toroidal electric field when significant deterioration of LH coupling occurs.

  19. Diagnostics of internal inductance in HT-7

    International Nuclear Information System (INIS)

    Zeng Li; Wan Baonian; Qian Jinping; Fan Hengyu

    2001-01-01

    Two arrays of Mirnov coils and a pair of concentric loops have been installed to superconducting tokamak HT-7. Software compensation and digital Fourier series expansion are the two techniques that have been applied successfully in measuring diamagnetic flux of concentric loops and internal inductance. The internal inductance of plasma l i , poloidal beta β p , Grad Shafranov parameter Λ, plasma minor radius α p and the center of the outermost magnetic flux surface Δ g are determined

  20. Insulating process for HT-7U central solenoid model coils

    International Nuclear Information System (INIS)

    Cui Yimin; Pan Wanjiang; Wu Songtao; Wan Yuanxi

    2003-01-01

    The HT-7U superconducting Tokamak is a whole superconducting magnetically confined fusion device. The insulating system of its central solenoid coils is critical to its properties. In this paper the forming of the insulating system and the vacuum-pressure-impregnating (VPI) are introduced, and the whole insulating process is verified under the super-conducting experiment condition

  1. Overview of the latest HT-7 experiments

    International Nuclear Information System (INIS)

    Wan Baonian; Luo Jiarong; Li Jiangang

    2005-01-01

    An overview of the HT-7 experimental progress during 2003-2004 is presented. The operational scenarios of H-mode, negative reversed shear (RS) and high l i were investigated for quasi-steady-state high performance plasma discharges. Stationary internal transport barriers (ITBs) with normalized performance β N *H 89 > 1-3 have been obtained with combined injection of lower hybrid (LH) and ion Bernstein (IB) waves for a duration of several hundred energy confinement times in weak negative reversed shear. The maximum fraction of non-inductive current was >90% I p . The increase of the total injected power up to 1 MW did not degrade the plasma confinement significantly in the RS operational scenario. Plasma performance and duration were mainly limited by two kinds of MHD instabilities and recycling. The high l i plasma was created by fast plasma current ramp-down and sustained by central LHCD and IBW heating for a duration of >1 s with a strongly peaked electron temperature profile. The highest central electron temperature obtained was as much as 4.5 keV. Stationary improved confinement has been observed in the high l i plasma. The longest plasma discharge, with a duration of 240 s, T e (0) ∼ 1 keV and a central electron density of >0.8 x 10 19 m -3 , was achieved in 2004. A fully LHW current driven plasma without using ohmic current in the central solenoid coils was sustained for 80 s. The main limitation for the pulse length was due to the recycling, which caused an uncontrollable rise in the electron density. The poloidal large-scale ExB time-varying flows, electrostatic and magnetic Reynolds stress were directly measured in the boundary plasma of the HT-7 tokamak. (author)

  2. Towards steady-state operational design for the data and PF control systems of the HT-7U

    International Nuclear Information System (INIS)

    Luo, J.R.; Zhu, L.; Wang, H.Z.; Ji, Z.S.; Wang, F.

    2003-01-01

    Fusion energy is an ultimate and inexhaustible source of energy for mankind and is expected to be obtained in controlled operation within this century. Among various possible candidates for fusion, the tokamak is presently the most qualified one, and since it uses superconducting magnetic coils, it will be adequate for steady-state operation. The HT-7U superconducting tokamak is a part of national project in China on fusion research, scheduled to become available on-line by the end of 2004 (Wan Y.X. and HT-7 and HT-7U Groups 2000 Overview of steady state operation of HT-7 and present status of the HT-7U project Nucl. Fusion 40 1057). The control system of the HT-7U is designed as a distributed control system (HT7UDCS), including many subsystems that provide the various functions of supervision, remote control, real-time monitoring, data acquisition and data handling. The major features of the HT-7U tokamak, which make long-pulse (∼1000 s) operation possible are the flexible poloidal field (PF) system, an auxiliary heating system, the current-driving system and a divertor system. In order to realize these features simultaneously, real-time data handling and analysis, along with a significant control capability is required. This paper discusses the design of the HT7UDCS. (author)

  3. Magnetic sensorless control experiment without drift problem on HT-7

    International Nuclear Information System (INIS)

    Nakamura, K.; Luo, J.R.; Wang, H.Z.; Ji, Z.S.; Wang, H.; Wang, F.; Qi, N.; Sato, K.N.; Hanada, K.; Sakamoto, M.; Idei, H.; Hasegawa, M.; Iyomasa, A.; Kawasaki, S.; Nakashima, H.; Higashijima, A.

    2006-01-01

    Magnetic sensorless control experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. Previously the horizontal position was calculated from the vertical field coil current and voltage without using signals of magnetic sensors like magnetic coils and flux loops placed near the plasma. The calculations are made focusing on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem of integrator of magnetic sensors. Two kinds of experiments were carried out, to keep the position constant and swing the position in a triangular waveform

  4. Design of a long pulse and low drift analog integrator in HT-7

    International Nuclear Information System (INIS)

    Liu Dongmei; Wan Baonian; Shen Biao

    2007-01-01

    Magnetic measurements are a fundamental diagnostic system for Tokamak. Inductive magnetic coils are used on HT-7. So the integrator is required to determine the magnetic field strength. This paper discusses the traditional analog integrator, and introduces a new integrator based on real-time drift compensation schemes. This new design can significantly reduce the integral error caused by input offset, temperature-induced drift, noise and so on. Operation in the HT-7 Tokamak shows that very low drift and noise characteristics compatible of the now integrators can meet requirement of long pulse discharges. (authors)

  5. High performance discharges near the operational limit in HT-7

    International Nuclear Information System (INIS)

    Li Jiangang; Wan Baonian; Luo Jiarong; Gao Xiang; Zhao Yanping; Kuang Guangli; Zhang Xiaodong; Yang Yu; Yi Bao; Bojiang Ding; Jikang Xie; Yuanxi Wan

    2001-01-01

    Efforts have been made on the HT-7 tokamak to extend the stable operation boundaries. Extensive RF boronization and siliconization have been used and a wider operational Hugill diagram has been obtained. The transit density reached 1.3 times the Greenwald density limit in ohmic discharges. A stationary high performance discharge with q a =2.1 has been obtained after siliconization. Confinement improvement was obtained as a result of the significant reduction of electron thermal diffusivity χ e in the outer region of the plasma. An improved confinement phase was also observed with LHCD in the density range of 70-120% of the Greenwald density limit. Off-axis LH wave power deposition was attributed to the weak hollow current density profile. Code simulations and measurements showed good agreement with the off-axis LH wave deposition. Supersonic molecular beam injection has been successfully used to achieve stable high density operation in the region of the Greenwald density limit. (author)

  6. Equilibrium optimization code OPEQ and results of applying it to HT-7U

    International Nuclear Information System (INIS)

    Zha Xuejun; Zhu Sizheng; Yu Qingquan

    2003-01-01

    The plasma equilibrium configuration has a strong impact on the confinement and MHD stability in tokamaks. For designing a tokamak device, it is an important issue to determine the sites and currents of poloidal coils which have some constraint conditions from physics and engineering with a prescribed equilibrium shape of the plasma. In this paper, an effective method based on multi-variables equilibrium optimization is given. The method can optimize poloidal coils when the previously prescribed plasma parameters are treated as an object function. We apply it to HT-7U equilibrium calculation, and obtain good results

  7. 5-HT7 Receptor Antagonists with an Unprecedented Selectivity Profile.

    Science.gov (United States)

    Ates, Ali; Burssens, Pierre; Lorthioir, Olivier; Lo Brutto, Patrick; Dehon, Gwenael; Keyaerts, Jean; Coloretti, Francis; Lallemand, Bénédicte; Verbois, Valérie; Gillard, Michel; Vermeiren, Céline

    2018-04-23

    Selective leads: In this study, we generated a new series of serotonin 5-HT 7 receptor antagonists. Their synthesis, structure-activity relationships, and selectivity profiles are reported. This series includes 5-HT 7 antagonists with unprecedented high selectivity for the 5-HT 7 receptor, setting the stage for lead optimization of drugs acting on a range of neurological targets. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  8. Experiments of full non-inductive current drive on HT-7

    International Nuclear Information System (INIS)

    Zhang, X.D.; Wu, Z.W.; Chen, Z.Y.; Gong, X.Z.; Wang, H.; Xu, D.; Huang, Y.; Luo, J.; Gao, X.; Hu, L.; Zhao, J.; Wan, B.N.; Li, J.

    2005-01-01

    Some experimental results of steady-state operation and full non-inductive current drive have been obtained on HT-7. Three types of experiment are used to study long pulse discharge, quasi-steady-state operation and full non-inductive current drive. The experiments show that the plasma current in the full non-inductive drive case is instable due to no adjusting effect of OH heating field, when the waveguide tube discharge lead to the LHW power injecting tokamak plasma decrease. This instability of plasma current will increase the interaction of plasma with limiter and first surface and bring impurity. All discharges of full non-inductive current drive are terminated because of impurity spurting. To adjust the LHW injection power for control the loop voltage during long pulse discharge is the most effective method for steady-state operation on HT-7. (author)

  9. Application of PLC timing control in the neutral beam injector of HT-7

    International Nuclear Information System (INIS)

    Song Shihua; Liu Zhimin; Liu Sheng; Hu Chundong

    2006-01-01

    HT-7 tokamak high power Neutral Beam Injector heating system runs in the mode of pulse timing-control of PLC. The thesis discusses the theory about the operation for the experiment of discharge, which is controlled by PLC logical connection and introduces excellent user-friendly operating interface and the development of the ladder application program and upper monitor program in the VB6.0 environment. Monitor the conditions of power and facility real time by the upper monitor interface. The application of PLC control system ensures the experiment facility running safely and convenient for modifying and setting the parameter simply during the course of experiment. (authors)

  10. 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

  11. Disruption mitigation experiment with massive gas injection of HT-7

    International Nuclear Information System (INIS)

    Zhuang Huidong; Zhang Xiaodong

    2013-01-01

    Massive gas injection (MGI) is a promising method on disruption mitigation. The working principle of the fast valve for disruption mitigation was introduced. The disruption mitigation experiments by MGI on HT-7 were described. The experiment shows that the impurities radiation is improved by injecting appropriate amount of gas, and the current quench rate is slow down, so the electromagnetic load on the device is mitigated. The experiments show that the fast valve can completely satisfy the requirement of disruption mitigation on HT-7. (authors)

  12. Plasma performance improvement with neon gas puffing in HT-7

    International Nuclear Information System (INIS)

    Gong, X.; Wan, B.; Li, J.; Shi, Y.; Zhang, X.; Zhu, Y.; Wu, Z.; Liu, H.; Qian, J.

    2005-01-01

    The neon gas puffing for the production of a radiative layer near the plasma edge with the improved energy and particle confinement has been investigated in HT-7 during the 2003 campaign. Plasma characteristics of these discharges in HT-7 are similar to the TEXTOR RI-mode discharges. The peaked electron temperature and the broadened density profiles were formed in these discharges with the combination of LHCD and IBW heating. The central electron temperature was increased by nearly 50%, compared those discharges with the same plasma parameters and injected power without the neon gas puffing. These discharges also exhibited relatively higher plasma inductance. (author)

  13. Quasi-steady-state operation around operational limit in HT-7

    International Nuclear Information System (INIS)

    Li, J.; Xie, J.K.; Wan, B.N.; Luo, J.R.; Gao, X.; Zhao, Y.; Yang, Y.; Kuang, G.L.; Bao, Y.; Ding, B.J.; Wan, Y.X.

    2001-01-01

    Efforts have been made on HT-7 tokamak for extending the stable operation boundaries. Extensive RF boronization and siliconization have been used and wider operational Hugill diagram was obtained. Transit density reached 1.3 time of Greenwald density limit in ohmic discharges. Stationary high performance discharge with q a =2.1 has been obtained after siliconization. Confinement improvement was obtained due to the significant reduction of electron thermal diffusivity χ e in the out region of the plasma. Improved confinement phase was also observed by LHCD under the density range 70%∼120% of Greenwald density limit. The weak hollow current density profile was attribute to off-axis LHW power deposition. Code simulations and measurements showed a good agreement of off-axis LHW deposition. Supersonic molecular beam injection has been successfully used to get stable high-density operation in the range of Greenwald density limit. (author)

  14. 5-HT7 receptor activation: procognitive and antiamnesic effects.

    Science.gov (United States)

    Meneses, A; Perez-Garcia, G; Liy-Salmeron, G; Ponce-López, T; Lacivita, E; Leopoldo, M

    2015-02-01

    The serotonin (5-hydroxytryptamine (5-HT)) 5-HT7 receptor is localized in brain areas mediating memory; however, the role of this receptor on memory remains little explored. First, demonstrating the associative nature of Pavlovian/instrumental autoshaping (P/I-A) task, rats were exposed (three sessions) to CS-US (Pavlovian autoshaping), truly random control, free operant, and presentations of US or CS, and they were compared with rats trained-tested for one session to the P/I-A procedure. Also, effects of the 5-HT7 receptor agonist LP-211 administered intraperitoneally after training was determined on short- (1.5 h) and long-term memory 24 and 48 h) and on scopolamine-induced memory impairment and cAMP production. Autoshaping and its behavioral controls were studied. Other animals were subjected to an autoshaping training session and immediately afterwards were given (intraperitoneal) vehicle or LP-211 (0.1-10 mg/kg) and/or scopolamine (0.2 mg/kg) and tested for short-term memory (STM) and long-term memory (LTM); their brains were extracted for the cAMP ELISA immunoassay. P/I-A group produced the higher %CR. LP-211 did not affect STM; nonetheless, at 0.5 and 1.0 mg/kg, it improved LTM. The 5-HT7 receptor antagonist SB-269970 (SB; 10.0 mg/kg) alone had no effect; nevertheless, the LP-211 (1.0 mg/kg) LTM facilitation was reversed by SB. The scopolamine (0.2 mg/kg) induced-decrement in CR was accompanied by significant increased cAMP production. The scopolamine-induced decrement in CR and increments in cAMP were significantly attenuated by LP-211. Autoshaping is a reliable associative learning task whose consolidation is facilitated by the 5-HT7 receptor agonist LP-211.

  15. Manufacture of the rectifier of the HT-7U PFPS

    International Nuclear Information System (INIS)

    Gao Ge; Fu Peng; Tang Lunjun; Wang Linsen

    2005-01-01

    The rectifiers of the HT-7U poloidal field power supply (PFPS) are introduced. A new control method, four quadrants converter, is brought forward, which overcomes the short-coming of both the circulating current mode and the non-circulating current mode. This control mode also resolves the problem of DC circulating current in the identical phase anti-parallel connection rectifiers when these rectifiers run in the circulating current mode. (authors)

  16. Design and simulation for the pulse high-voltage DC power supply (HVPS) of 1.2 MW/2.45 GHz HT-7U lower hybrid current drive system

    International Nuclear Information System (INIS)

    Huang Yiyun; Kuang Guangli; Xu Weihua; Liu Baohua; Lin Jianan; Wu Junshuan; Zheng Guanghua; Yang Chunshen

    2000-01-01

    The superconducting tokamak HT-7U has been designed by the Institute of Plasma Physics since 1998 and will be set up before 2003. The 1.2 MW/2.45 GHz HT-7U LHCD (Lower hybrid current drive) system which being the most efficient non-induction device can heat the plasma and drive the plasma current has been efficiently in operation now, and a particular design of the 2.8 MW/-35 kV high-voltage DC power supply has been already completed and will apply to the klystron of LHCD on HT-7 and the future HT-7U, and the project of the power supply has been examined and approved professionally by an authorized group of high-level specialist in the Institute of Plasma Physics. The detailed design of the power supply and the simulation results are referred

  17. Development of a Fast Valve for Disruption Mitigation and its Preliminary Application to EAST and HT-7

    International Nuclear Information System (INIS)

    Zhuang Huidong; Zhang Xiaodong

    2013-01-01

    In large tokamaks, disruption of high current plasma would damage plasma facing component surfaces (PFCs) or other inner components due to high heat load, electromagnetic force load and runaway electrons. It would also influence the subsequent plasma discharge due to production of impurities during disruptions. So the avoidance and mitigation of disruptions is essential for the next generation of tokamaks, such as ITER. Massive gas injection (MGI) is a promising method of disruption mitigation. A new fast valve has been developed successfully on EAST. The valve can be opened in 0.5 ms, and the duration of open state is largely dependent on the gas pressure and capacitor voltage. The throughput of the valve can be adjusted from 0 mbar·L to 700 mbar·L by changing the capacitor voltage and gas pressure. The response time and throughput of the fast valve can meet the requirement of disruption mitigation on EAST. In the last round campaign of EAST and HT-7 in 2010, the fast valve has operated successfully. He and Ar was used for the disruption mitigation on HT-7. By injecting the proper amount of gas, the current quench rate could be slowed down, and the impurities radiation would be greatly improved. In elongated plasmas of EAST discharges, the experimental data is opposite to that which is expected. (magnetically confined plasma)

  18. The renewed HT-7 plasma control system based on real-time Linux cluster

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Q.P., E-mail: qpyuan@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J.; Zhang, R.R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Walker, M.L.; Penaflor, B.G.; Piglowski, D.A.; Johnson, R.D. [General Atomics, DIII-D National Fusion Facility, San Diego, CA (United States)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer The hardware and software structure of the new HT-7 plasma control system (HT-7 PCS) is reported. Black-Right-Pointing-Pointer All original systems were integrated in the new HT-7 PCS. And the implementation details of the control algorithms are given in the paper. Black-Right-Pointing-Pointer Different from EAST PCS, the AC operation mode is realized in HT-7 PCS. Black-Right-Pointing-Pointer The experiment results are discussed. Good control performance has been obtained. - Abstract: In order to improve the synchronization, flexibility and expansibility of the plasma control on HT-7, a new plasma control system (HT-7 PCS) was constructed. The HT-7 PCS was based on a real-time Linux cluster with a well-defined, robust and flexible software infrastructure which was adapted from DIII-D PCS. In this paper, the hardware structure and system customization details for HT-7 PCS are reported. The plasma position and current control, plasma density control and off-normal event detection, which were realized in separated systems originally, have been integrated and implemented in such HT-7 PCS. All these control algorithms have been successfully validated in the last several HT-7 experiment campaigns. Good control performance has been achieved and the experiment results are discussed in the paper.

  19. Identification of anomalous Doppler resonance effect during current ramp down in HT-7 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Li Erzhong, E-mail: rzhonglee@ipp.ac.c [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Hu Liqun; Ling Bili; Liu Yong; Ti Ang; Zhou Reijie; Lu Hongwei; Gao Xiang [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China)

    2010-09-21

    The abrupt steep jump of electron cyclotron emission (ECE) signals during current ramp-down has been observed and explained by an anomalous Doppler resonance effect (ADR). The identifying process of ADR was presented based on the fast Fourier transform (FFT) technique. The threshold value for triggering a steep jump on ECE signals has been identified under different discharge conditions.

  20. Design of Amplifier Circuit for the HT-7 Tokamak Thomson Scattering System

    International Nuclear Information System (INIS)

    Shi Lingwei; Ling Bili; Zhao Junyu; Yang Li; Zang Qing; Hu Qingsheng; Jia Yanqing

    2008-01-01

    Thomson scattering diagnostic is important for measuring electron temperature and density profiles. To improve the signal-to-noise ratio, a silicon avalanche photodiode (APD) with high quantum efficiency, high sensitivity, and high gain up to 100 was adopted to measure the Thomson scattering spectrum. A preamplifier, which has low noise, high bandwidth, and high sensitivity, was designed with suitable transimpedance. Using AD8367 as the post-amplifier, good performance of the APD readout electronics have been obtained. A discussion is presented on the performance of the amplifier using a laser diode to simulate the Thomson scattering light. The test results indicate that the designed circuit has a high amplifying factor and fast rising edge. So reduction of the integral gate of the CAMAC ADC converter can improve the signal-to-noise ratio. (brief communication and research note)

  1. Heating and active control of profiles and transport by IBW in the HT-7 tokamak

    International Nuclear Information System (INIS)

    Zhao Yanping

    2002-01-01

    By a series of technical improvements and intensive RF boronization, significant progresses on the IBW heating and control of profiles and transport has been obtained since last IAEA meeting. Both on-axis and off-axis electron heating with global peaked and local steeped electron pressure profile was realized if the resonant layer is in plasma far from the edge region. Maximum increment of electron temperature was about 2 keV at power of 200 kW. The heating factor reached 9.4 eV x 10 13 cm -3 /kW. Reduction of local electron heat transport around resonant layer has been observed. Significant improvement of particle confinement by a factor of 2-4 with very peaked density profile was obtained if 5/2-deuterium resonant layer is located at the plasma edge. Global transport and edge poloidal velocity shear can been controlled by IBW. (author)

  2. Wall conditioning with a high magnetic field in HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Li Jiangang; Gu Xuemao; Gao Xiang; Zhang Souying; Jie Yingxian; Yang Xiaokang

    2000-01-01

    ICRF wall conditioning techniques, which includes the hydrogen removal, impurity cleaning, boronization and siliconization, were described in this paper. This new technique has been demonstrated to be very effective for wall conditioning, recycling, isotopic control and used daily during experiments. The RF plasma parameters were measured as T e =3-8 eV, T i =0.5-2 keV, n e =0.3-5 x 10 17 m -3 by different diagnostics. The nontoxic and nonexplosive solid carborane powder was used for the RF boronization. Energetic ions cracked the carborane molecule and the boron ions impacted and deposited onto first wall. Comparing with GDC boronization, the B/C coating film shows the higher adhesion, better uniformity and longer lifetime to the plasma discharges. Siliconization was carried out by using a high field side long RF antenna, which made the discharge more uniform. The ratio of SiH 4 to helium is about 5:95 at the pressure range of P v =0.8-8 x 10 -2 Pa. Compare with boronization, it showed quicker recovery from a bad wall condition due to leakage of air to good wall condition. Plasma density could be easily controlled after siliconization. But the lifetime is much shorter than that obtained by boronization. Plasma performance has been improved after RF boronization and siliconization. (author)

  3. Effects of the potential 5-HT7 receptor agonist AS 19 in an autoshaping learning task.

    Science.gov (United States)

    Perez-García, Georgina S; Meneses, A

    2005-08-30

    This work aimed to evaluate further the role of 5-HT7 receptors during memory formation in an autoshaping Pavlovian/instrumental learning task. Post-training administration of the potential 5-HT7 receptor agonist AS 19 or antagonist SB-269970 enhanced memory formation or had no effect, respectively. The AS 19 facilitatory effect was reversed by SB-269970, but not by the selective 5-HT1A antagonist WAY100635. Amnesia induced by scopolamine (cholinergic antagonist) or dizocilpine (NMDA antagonist) was also reversed by AS 19. Certainly, reservations regarding the selectivity of AS 19 for 5-HT7 and other 5-HT receptors in vivo are noteworthy and, therefore, its validity for use in animal models as a pharmacological tool. Having mentioned that, it should be noticed that together these data are providing further support to the notion of the 5-HT7 receptors role in memory formation. Importantly, this 5-HT7 receptor agonist AS 19 appears to represent a step forward respect to the notion that potent and selective 5-HT7 receptor agonists can be useful in the treatment of dysfunctional memory in aged-related decline and Alzheimer's disease.

  4. Improvement of ketamine-induced social withdrawal in rats: the role of 5-HT7 receptors.

    Science.gov (United States)

    Hołuj, Małgorzata; Popik, Piotr; Nikiforuk, Agnieszka

    2015-12-01

    Social withdrawal, one of the core negative symptoms of schizophrenia, can be modelled in the social interaction (SI) test in rats using N-methyl-D-aspartate receptor glutamate receptor antagonists. We have recently shown that amisulpride, an antipsychotic with a high affinity for serotonin 5-HT7 receptors, reversed ketamine-induced SI deficits in rats. The aim of the present study was to further elucidate the potential involvement of 5-HT7 receptors in the prosocial action of amisulpride. Acute administration of amisulpride (3 mg/kg) and SB-269970 (1 mg/kg), a 5-HT7 receptor antagonist, reversed ketamine-induced social withdrawal, whereas sulpiride (20 or 30 mg/kg) and haloperidol (0.2 mg/kg) were ineffective. The 5-HT7 receptor agonist AS19 (10 mg/kg) abolished the prosocial efficacy of amisulpride (3 mg/kg). The coadministration of an inactive dose of SB-269970 (0.2 mg/kg) showed the prosocial effects of inactive doses of amisulpride (1 mg/kg) and sulpiride (20 mg/kg). The anxiolytic chlordiazepoxide (2.5 mg/kg) and the antidepressant fluoxetine (2.5 mg/kg) were ineffective in reversing ketamine-induced SI deficits. The present study suggests that the antagonism of 5-HT7 receptors may contribute towards the mechanisms underlying the prosocial action of amisulpride. These results may have therapeutic implications for the treatment of negative symptoms in schizophrenia and other disorders characterized by social withdrawal.

  5. Spatial memory deficit across aging: current insights of the role of 5-HT7 receptors

    Directory of Open Access Journals (Sweden)

    Gregory eBeaudet

    2015-01-01

    Full Text Available Elderly persons often face biological, psychological or social changes over time that may cause discomfort or morbidity. While some cognitive domains remain stable over time, others undergo a decline. Spatial navigation is a complex cognitive function essential for independence, safety and quality of life. While egocentric (body-centered navigation is quite preserved during aging, allocentric (externally-centered navigation — based on a cognitive map using distant landmarks — declines with age. Recent preclinical studies showed that serotonergic 5-HT7 receptors are localized in brain regions associated with allocentric spatial navigation processing. Behavioral assessments with pharmacological or genetic tools have confirmed the role of 5-HT7 receptors in allocentric navigation. Moreover, few data suggested a selective age-related decrease in the expression of 5-HT7 receptors in pivotal brain structures implicated in allocentric navigation such as the hippocampal CA3 region. We aim to provide a short overview of the potential role of 5-HT7 receptors in spatial navigation, and to argue for their interests as therapeutic targets against age-related cognitive decline.

  6. Mechanical Stimulation of the HT7 Acupuncture Point to Reduce Ethanol Self-Administration in Rats

    Directory of Open Access Journals (Sweden)

    Suk-Yun Kang

    2017-01-01

    Full Text Available Background. Alcoholism, which is a disabling addiction disorder, is a major public health problem worldwide. The present study was designed to determine whether the application of acupuncture at the Shenmen (HT7 point suppresses voluntary alcohol consumption in addicted rats and whether this suppressive effect is potentiated by the administration of naltrexone. Methods. Rats were initially trained to self-administer a sucrose solution by operating a lever. A mechanical acupuncture instrument (MAI for objective mechanical stimulation was used on rats whose baseline response had been determined. In addition, the effect of HT7 acupuncture on beta-endorphin concentration and ethanol intake via naltrexone were investigated in different groups. Results. We found that ethanol intake and beta-endorphin level in rats being treated with the MAI at the HT7 point reduced significantly. The treatment of naltrexone at high doses reduced the ethanol intake and low-dose injection of naltrexone in conjunction with the MAI also suppressed ethanol intake. Conclusions. The results of the current study indicate that using the MAI at the HT7 point effectively reduces ethanol consumption in rats. Furthermore, the coadministration of the MAI and a low dose of naltrexone can produce some more potent reducing effect of ethanol intake than can acupuncture alone.

  7. Synthesis and In Vitro Evaluation of Oxindole Derivatives as Potential Radioligands for 5-HT7 Receptor Imaging with PET

    DEFF Research Database (Denmark)

    Herth, Matthias Manfred; Volk, Balázs; Pallagi, Katalin

    2012-01-01

    The most recently discovered serotonin (5-HT) receptor subtype, 5-HT(7), is considered to be associated with several CNS disorders. Noninvasive in vivo positron emission tomography (PET) studies of cerebral 5-HT(7) receptors could provide a significant advance in the understanding of the neurobio...

  8. Investigation of the Energy Confinement in Ohmic and LHCD Plasmas in HT-7

    International Nuclear Information System (INIS)

    Zhang Xiaoqing; Wan Baonian; Shen Biao; Hu Xiwei; Qian Jinping; Fan Hengyu; Ding Yonghua

    2006-01-01

    Investigation of the energy confinement in ohmic and lower hybrid current drive (LHCD) plasmas in HT-7 has been performed. In ohmic discharges at low densities the global energy confinement time τ E increases almost linearly with the density, saturates at a critical density (2.5 x 10 13 /cm 3 for HT-7) and is nearly constant at higher densities. The energy confinement time is in good agreement with the Neo-Alcator scaling law at different densities and currents. In the LHCD plasmas the global energy confinement time similar to that of the L-mode discharges has been observed to be in good agreement with the low confinement mode (L mode) scaling law of ITER89-P in higher electron density and plasma current

  9. Novel 2-aminotetralin and 3-aminochroman derivatives as selective serotonin 5-HT7 receptor agonists and antagonists.

    Science.gov (United States)

    Holmberg, Pär; Sohn, Daniel; Leideborg, Robert; Caldirola, Patrizia; Zlatoidsky, Pavel; Hanson, Sverker; Mohell, Nina; Rosqvist, Susanne; Nordvall, Gunnar; Johansson, Anette M; Johansson, Rolf

    2004-07-29

    The understanding of the physiological role of the G-protein coupled serotonin 5-HT(7) receptor is largely rudimentary. Therefore, selective and potent pharmacological tools will add to the understanding of serotonergic effects mediated through this receptor. In this report, we describe two compound classes, chromans and tetralins, encompassing compounds with nanomolar affinity for the 5-HT(7) receptor and with good selectivity. Within theses classes, we have discovered both agonists and antagonists that can be used for further understanding of the pharmacology of the 5-HT(7) receptor.

  10. Labeling and preliminary in vivo evaluation of the 5-HT7 receptor selective agonist [(11)C]E-55888

    DEFF Research Database (Denmark)

    Hansen, Hanne D; Andersen, Valdemar L; Lehel, Szabolcs

    2015-01-01

    E-55888 has been identified as a selective serotonin 7 (5-HT7) receptor agonist. In this study, we describe the synthesis, radiolabeling and in vivo evaluation of [(11)C]E-55888 as a radioligand for positron emission tomography (PET) imaging. [(11)C]E-55888 was obtained by N-methylation of an app...... neither be displaced by the structurally different 5-HT7 receptor ligand SB-269970 nor by self-block with unlabeled E-55888. Based on these data, [(11)C]E-55888 does not show promise as a PET radioligand for imaging the 5-HT7 receptor in vivo....

  11. Evaluation of 3-Ethyl-3-(phenylpiperazinylbutyl)oxindoles as PET Ligands for the Serotonin 5-HT7 Receptor

    DEFF Research Database (Denmark)

    Herth, Matthias M; Andersen, Valdemar L; Hansen, Hanne D

    2015-01-01

    We have investigated several oxindole derivatives in the pursuit of a 5-HT7 receptor PET ligand. Herein the synthesis, chiral separation, and pharmacological profiling of two possible PET candidates toward a wide selection of CNS-targets are detailed. Subsequent (11)C-labeling and in vivo evaluat...... evaluation in Danish landrace pigs showed that both ligands displayed high brain uptake. However, neither of the radioligands could be displaced by the 5-HT7 receptor selective inverse agonist SB-269970....

  12. Am5-HT7: molecular and pharmacological characterization of the first serotonin receptor of the honeybee (Apis mellifera).

    Science.gov (United States)

    Schlenstedt, Jana; Balfanz, Sabine; Baumann, Arnd; Blenau, Wolfgang

    2006-09-01

    The biogenic amine serotonin (5-HT) plays a key role in the regulation and modulation of many physiological and behavioural processes in both vertebrates and invertebrates. These functions are mediated through the binding of serotonin to its receptors, of which 13 subtypes have been characterized in vertebrates. We have isolated a cDNA from the honeybee Apis mellifera (Am5-ht7) sharing high similarity to members of the 5-HT(7) receptor family. Expression of the Am5-HT(7) receptor in HEK293 cells results in an increase in basal cAMP levels, suggesting that Am5-HT(7) is expressed as a constitutively active receptor. Serotonin application to Am5-ht7-transfected cells elevates cyclic adenosine 3',5'-monophosphate (cAMP) levels in a dose-dependent manner (EC(50) = 1.1-1.8 nm). The Am5-HT(7) receptor is also activated by 5-carboxamidotryptamine, whereas methiothepin acts as an inverse agonist. Receptor expression has been investigated by RT-PCR, in situ hybridization, and western blotting experiments. Receptor mRNA is expressed in the perikarya of various brain neuropils, including intrinsic mushroom body neurons, and in peripheral organs. This study marks the first comprehensive characterization of a serotonin receptor in the honeybee and should facilitate further analysis of the role(s) of the receptor in mediating the various central and peripheral effects of 5-HT.

  13. Peripheral 5-HT7 receptors as a new target for prevention of lung injury and mortality in septic rats.

    Science.gov (United States)

    Cadirci, Elif; Halici, Zekai; Bayir, Yasin; Albayrak, Abdulmecit; Karakus, Emre; Polat, Beyzagul; Unal, Deniz; Atamanalp, Sabri S; Aksak, Selina; Gundogdu, Cemal

    2013-10-01

    Sepsis is a complex pathophysiological event involving metabolic acidosis, systemic inflammatory response syndrome, tissue damage and multiple organ dysfunction syndrome. Although many new mechanisms are being investigated to enlighten the pathophysiology of sepsis, there is no effective treatment protocol yet. Presence of 5-HT7 receptors in immune tissues prompted us to hypothesize that these receptors have roles in inflammation and sepsis. We investigated the effects of 5-HT7 receptor agonists and antagonists on serum cytokine levels, lung oxidative stress, lung histopathology, nuclear factor κB (NF-κB) positivity and lung 5-HT7 receptor density in cecal ligation and puncture (CLP) induced sepsis model of rats. Agonist administration to septic rats increased survival time; decreased serum cytokine response against CLP; decreased oxidative stress and increased antioxidant system in lungs; decreased the tissue NF-κB immunopositivity, which is high in septic rats; and decreased the sepsis-induced lung injury. In septic rats, as a result of high inflammatory response, 5-HT7 receptor expression in lungs increased significantly and agonist administration, which decreased inflammatory response and related mortality, decreased the 5-HT7 receptor expression. In conclusion, all these data suggest that stimulation of 5-HT7 receptors may be a new therapeutic target for prevention of impaired inflammatory response related lung injury and mortality. Copyright © 2013 Elsevier GmbH. All rights reserved.

  14. Pellet injection experiments on tokamaks in ASIPP, China

    International Nuclear Information System (INIS)

    Yang, Y.; Bao, Y.; Li, J.; Gu, X.; He, Y.

    2001-01-01

    Pellet injection has been proved to be an effective method for deep fueling of fusion devices. Improvements of both the particle confinement and the energy confinement were observed in many experiments. In HT-6M and HT-7 tokamaks, single and multi-pellet experiments are tried, and attractive results are obtained. (author)

  15. Pellet injection experiments on tokamaks in ASIPP, China

    International Nuclear Information System (INIS)

    Yang, Y.; Bao, Y.; Li, J.; Gu, X.; He, Y.

    1999-01-01

    Pellet Injection has been proven to be an effective method for deep fuelling of fusion devices. Improvements of both the particle confinement and the energy confinement were observed in many experiments. In HT-6M and HT-7 tokamaks, single and multi-pellet experiments are tried, and attractive results are obtained

  16. Temperature field and thermal stress analysis of the HT-7U vacuum vessel

    International Nuclear Information System (INIS)

    Song Yuntao; Yao Damao; Wu Songtao; Weng Peide

    2000-01-01

    The HT-7U vacuum vessel is an all-metal-welded double-wall interconnected with toroidal and poloidal stiffening ribs. The channels formed between the ribs and walls are filled with boride water as a nuclear shielding. On the vessel surface facing the plasma are installed cable-based Ohmic heaters. Prior to plasma operation the vessel is to be baked out and discharge cleaned at about 250 degree C. During baking out the non-uniformity of temperature distribution on the vacuum vessel will bring about serious thermal stress that can damage the vessel. In order to determine and optimize the design of the HT-7U vacuum vessel, a three-dimensional finite element model was performed to analyse its temperature field and thermal stress. the maximal thermal stress appeared on the round of lower vertical port and maximal deformation located just on the region between the upper vertical port and the horizontal port. The results show that the reinforced structure has a good capability of withstanding the thermal loads

  17. Deqi Induction by HT7 Acupuncture Alters Theta and Alpha Band Coherence in Human Healthy Subjects

    Directory of Open Access Journals (Sweden)

    Go-Eun Lee

    2017-01-01

    Full Text Available The aim of this preliminary study is to investigate the changes in phase synchronization in the theta and alpha bands before and during the performance of classical acupuncture on the Sinmun (HT7. The electroencephalogram (EEG signals from nine healthy young subjects were recorded before and during acupuncture in the “closed-eye” state. The EEG signals were acquired from 19 surface scalp electrodes (FP1, FP2, F7, F3, Fz F4, F8, T3, C3, Cz, C4, T4, T5, P3, Pz, P4, T6, O1, and O2. Needles were inserted into the HT7 bilaterally and were then manipulated to induce deqi and retained for 15 minutes. Phase synchronization was measured by phase coherence. In the theta band, coherence significantly increased between the temporal (T5, T6 and occipital areas (O1, O2 during the acupuncture stimulation. In the alpha band, coherence significantly increased between the left temporal area (T5 and other areas (frontal, parietal, and occipital. Phase coherence in the theta and alpha bands tended to increase during the retention of the acupuncture needles after deqi. Therefore, it can be concluded that acupuncture stimulation with deqi is clinically effective via the central nervous system (CNS.

  18. Synthesis and pharmacological evaluation of a new series of radiolabeled ligands for 5-HT7 receptor PET neuroimaging

    International Nuclear Information System (INIS)

    Colomb, Julie; Becker, Guillaume; Forcellini, Elsa; Meyer, Sandra; Buisson, Lauriane; Zimmer, Luc; Billard, Thierry

    2014-01-01

    Introduction: The brain serotonin-7 receptor (5-HT 7 ) is the most recently discovered serotonin receptor. It is targeted by several drug-candidates in psychopharmacology and neuropharmacology. In these fields, positron emission tomography (PET) is a molecular imaging modality offering great promise for accelerating the development process from preclinical discovery to clinical phases. We recently described fluorinated 5-HT 7 radioligands, inspired by the structure of SB269970, the prototypical 5-HT 7 antagonist. Although these results were promising, it appeared that the radiotracer-candidates suffered, among other drawbacks, from too low a 5-HT 7 receptor affinity. Methods: In the present study, seven structural analogs of SB269970 were synthesized using design strategies aiming to improve their radiopharmacological properties. Their 5-HT 7 binding properties were investigated by cellular functional assay. The nitro-precursors of the analogs were radiolabeled by [ 18 F-]nucleophilic substitution, and in vitro autoradiography was performed in rat brain, followed by in vivo microPET. Result: The chemical and radiochemical purity of the fluorine radiotracers was > 99% with specific activity in the 40–129 GBq/μmol range. The seven derivatives presented heterogeneous binding affinities toward 5-HT 7 and 5-HT 1A receptors. While [ 18 F]2F3P3 had promising characteristics in vitro, it showed poor brain penetration in vivo, partially reversed after pharmacological inhibition of P-glycoprotein. Conclusions: These results indicated that, while chemical modification of these series improved several radiotracer-candidates in terms of 5-HT 7 receptor affinity and specificity toward 5-HT 1A receptors, other physicochemical modulations would be required in order to increase brain penetration

  19. Expression and role of 5-HT7 receptor in brain and intestine in rats with irritable bowel syndrome.

    Science.gov (United States)

    Zou, Bai-cang; Dong, Lei; Wang, Yan; Wang, Sheng-hao; Cao, Ming-bo

    2007-12-05

    The 5-hydroxytryptamine7 receptor (5-HT(7) receptor, 5-HT(7)R) plays an important role in the regulation of smooth muscle relaxation and visceral sensation and might be involved in the pathogenesis of the gastrointestinal dyskinesia, abdominal pain and visceral paresthesia in irritable bowel syndrome (IBS). The aim of this study was to investigate the role of the 5-HT(7) receptor in the pathogenesis of IBS. A rat model of irritable bowel syndrome with diarrhea (IBS-D) was established by colonic instillation of acetic acid and restraint stress. A rat model with irritable bowel syndrome with constipation (IBS-C) was established by stomach irrigated with 0 - 4 degrees C cool water daily for 14 days. The content and distribution of 5-HT in the brain and gut were examined by immunohistochemistry and the mRNA expression of the 5-HT(7) receptor was determined by fluorescent quantitative reverse transcription polymerase chain reaction. The accumulation of cyclic adenosine monophosphate (cAMP) in all the same tissues was measured by radioimmunity. The models of IBS were reliable by identification. The immunohistochemistry results showed that there were significantly more 5-HT positive cells in the IBS-D group than in the control group in the hippocampus, hypothalamus, jejunum, ileum, proximate colon and distal colon (P intestine is related to the IBS pathogenesis. The up-regulated expression of the 5-HT(7) receptor in the brain and colon might play an important role in the pathogenesis of IBS-C.

  20. Observation of Electron Energy Pinch in HT-7 ICRF Heated Plasmas

    International Nuclear Information System (INIS)

    Ding Siye; Wan Baonian; Ti Ang; Zhang Xinjun; Liu Zixi; Qian Jinping; Zhong Guoqiang; Duan Yanmin; Wang Lu

    2014-01-01

    Inward energy transport (pinch phenomenon) in the electron channel is observed in HT-7 plasmas using off-axis ion cyclotron resonance frequency (ICRF) heating. Experimental results and power balance transport analysis by TRANSP code are presented in this article. With the aids of GLF23 and Chang-Hinton transport models, which predict energy diffusivity in experimental conditions, the estimated electron pinch velocity is obtained by experimental data and is found reasonably comparable to the results in the previous study, such as Song on Tore Supra. Density scanning shows that the energy convective velocity in the electron channel has a close relation to density scale length, which is qualitatively in agreement with Wang's theoretical prediction. The parametric dependence of electron energy convective velocity on plasma current is still ambiguous and is worthy of future research on EAST. (magnetically confined plasma)

  1. Selective labelling of 5-HT7 receptor recognition sites in rat brain using [3H]5-carboxamidotryptamine

    International Nuclear Information System (INIS)

    Stowe, R.L.; Barnes, N.M.

    1998-01-01

    The aim of the present study was to establish a radioligand binding assay to selectively label the native 5-HT 7 receptor expressed in rat brain. In rat whole brain (minus cerebellum and striatum) homogenate, (±)-pindolol (10 μM)-insensitive [ 3 H]5-CT ([ 3 H]5-carboxamidotryptamine; 0.5 nM) specific binding (defined by 5-HT, 10 μM) displayed a pharmacological profile similar to the recombinant 5-HT 7 receptor, although the Hill coefficients for competition curves generated by methiothepin, ritanserin, sumatriptan, clozapine and pimozide were significantly less than unity. In homogenates of rat hypothalamus, (±)-pindolol (10 μM)-insensitive [ 3 H]5-CT recognition sites also resembled, pharmacologically, the 5-HT 7 receptor, although pimozide still generated Hill coefficients significantly less than unity. Subsequent studies were performed in the additional presence of WAY100635 (100 nM) to prevent [ 3 H]5-CT binding to residual, possibly, 5-HT 1A sites. Competition for this [ 3 H]5-CT binding indicated the labelling in whole rat brain homogenate of a homogenous population of sites with the pharmacological profile of the 5-HT 7 receptor. Saturation studies also indicated that (±)-pindolol (10 μM)/WAY 100635 (100 nM)-insensitive [ 3 H]5-CT binding to homogenates of whole rat brain was saturable and to an apparently homogenous population of sites which were labelled with nanomolar affinity (B max =33.2±0.7 fmol mg -1 protein, pK d =8.78±0.05, mean±S.E.M., n=3). The development of this 5-HT 7 receptor binding assay will aid investigation of the rat native 5-HT 7 receptor. (Copyright (c) 1998 Elsevier Science B.V., Amsterdam. All rights reserved.)

  2. 5-HT1A and 5-HT7 receptor crosstalk in the regulation of emotional memory: implications for effects of selective serotonin reuptake inhibitors.

    Science.gov (United States)

    Eriksson, Therese M; Holst, Sarah; Stan, Tiberiu L; Hager, Torben; Sjögren, Benita; Ogren, Sven Öve; Svenningsson, Per; Stiedl, Oliver

    2012-11-01

    This study utilized pharmacological manipulations to analyze the role of direct and indirect activation of 5-HT(7) receptors (5-HT(7)Rs) in passive avoidance learning by assessing emotional memory in male C57BL/6J mice. Additionally, 5-HT(7)R binding affinity and 5-HT(7)R-mediated protein phosphorylation of downstream signaling targets were determined. Elevation of 5-HT by the selective serotonin reuptake inhibitor (SSRI) fluoxetine had no effect by itself, but facilitated emotional memory performance when combined with the 5-HT(1A)R antagonist NAD-299. This facilitation was blocked by the selective 5-HT(7)R antagonist SB269970, revealing excitatory effects of the SSRI via 5-HT(7)Rs. The enhanced memory retention by NAD-299 was blocked by SB269970, indicating that reduced activation of 5-HT(1A)Rs results in enhanced 5-HT stimulation of 5-HT(7)Rs. The putative 5-HT(7)R agonists LP-44 when administered systemically and AS19 when administered both systemically and into the dorsal hippocampus failed to facilitate memory. This finding is consistent with the low efficacy of LP-44 and AS19 to stimulate protein phosphorylation of 5-HT(7)R-activated signaling cascades. In contrast, increasing doses of the dual 5-HT(1A)R/5-HT(7)R agonist 8-OH-DPAT impaired memory, while co-administration with NAD-299 facilitated of emotional memory in a dose-dependent manner. This facilitation was blocked by SB269970 indicating 5-HT(7)R activation by 8-OH-DPAT. Dorsohippocampal infusion of 8-OH-DPAT impaired passive avoidance retention through hippocampal 5-HT(1A)R activation, while 5-HT(7)Rs appear to facilitate memory processes in a broader cortico-limbic network and not the hippocampus alone. Copyright © 2012 Elsevier Ltd. All rights reserved.

  3. Effects of the 5-HT7 receptor antagonists SB-269970 and DR 4004 in autoshaping Pavlovian/instrumental learning task.

    Science.gov (United States)

    Meneses, Alfredo

    2004-12-06

    There is an important debate regarding the functional role of the 5-HT(1A) and 5-HT(7) receptor in memory systems. Hence, the objective of this paper is to investigate the function of serotonin (5-hydroxytryptamine, 5-HT) in memory consolidation, utilising an autoshaping Pavlovian/instrumental learning test. Specific antagonists at 5-HT(1A) (WAY 100635) and 5-HT(7) (SB-269970 or DR 4004) receptors administered i.p. or s.c.) after training, significantly decreased the improvement of performance produced by the 5-HT(1A/7) agonist 8-OH-DPAT to levels lower than controls'. These same antagonists attenuated the decreased level of performance produced by mCPP, although they decrease the performance levels after p-chloroamphetamine (PCA) lesion of the 5-HT system, which has no effect on its own on the conditioned response. Moreover, SB-269970 or DR 4004 reversed amnesia induced by scopolamine and dizocilpine. These data confirm a role for 5-HT(1A) and 5-HT(7) receptors in memory formation and support the hypothesis that serotonergic, cholinergic, and glutamatergic systems interact in cognitively impaired animals. These findings support a potential role for both 5-HT(1A) and 5-HT(7) receptors in the pathophysiology and/or treatment of schizophrenia, cognitive deficits and the mechanism of action of atypical antipsychotic drugs.

  4. Design and realization on function of pre-forming and continuous winding for HT-7U special winding machine

    International Nuclear Information System (INIS)

    Yu Jie; Gao Daming; Wen Jun; Zhu Wenhua; Cheng Leping; Tao Yuming

    2000-05-01

    The winding machine is one of the critical facilities for R and D of HT-7U construction. The machine mainly consists of five parts, CICC pay-off spool, a four-rollers straightening assembly, a four-roller forming/bending assembly, continuous winding structure and CNC control system with three-axis CNC control. The facility is needed for CICC magnet fabrication of HT-7U. The main requirements of the winding machine are: continuous winding to reduce number of joints inside the coils; pre-forming CICC conductor to avoid winding with tension; suitable for all TF and PF coils within the scope of various coil shape and dimension limit; improving the configuration tolerance, specially flatness of the CICC conductor. The author emphasizes on the design and realization on function of Pre-forming and Continuous Winding for HT-7U special winding machine. The winding machine with high accuracy has just been developed and applied to the construction of HT-7U model coils

  5. Sensitization of restraint-induced corticosterone secretion after chronic restraint in rats: Involvement of 5-HT7 receptors

    Science.gov (United States)

    García-Iglesias, Brenda B.; Mendoza-Garrido, María E.; Gutiérrez-Ospina, Gabriel; Rangel-Barajas, Claudia; Noyola-Díaz, Martha; Terrón, José A.

    2013-01-01

    Serotonin (5-HT) modulates the hypothalamic-pituitary-adrenal (HPA) axis response to stress. We examined the effect of chronic restraint stress (CRS; 20 min/day) as compared to control (CTRL) conditions for 14 days, on: 1) restraint-induced ACTH and corticosterone (CORT) secretion in rats pretreated with vehicle or SB-656104 (a 5-HT7 receptor antagonist); 2) 5-HT7 receptor-like immunoreactivity (5-HT7-LI) and protein in the hypothalamic paraventricular nucleus (PVN) and adrenal glands (AG); 3) baseline levels of 5-HT and 5-hydroxyindolacetic acid (5-HIAA), and 5-HIAA/5-HT ratio in PVN and AG; and 4) 5-HT-like immunoreactivity (5-HT-LI) in AG and tryptophan hydroxylase (TPH) protein in PVN and AG. On day 15, animals were subdivided into Treatment and No treatment groups. Treatment animals received an i.p. injection of vehicle or SB-656104; No Treatment animals received no injection. Sixty min later, Treatment animals were either decapitated with no further stress (0 min) or submitted to acute restraint (10, 30, 60 or 120 min); hormone serum levels were measured. No Treatment animals were employed for the rest of measurements. CRS decreased body weight gain and increased adrenal weight. In CTRL animals, acute restraint increased ACTH and CORT secretion in a time of restraint-dependent manner; both responses were inhibited by SB-656104. Exposure to CRS abolished ACTH but magnified CORT responses to restraint as compared to CTRL conditions; SB-656104 had no effect on ACTH levels but significantly inhibited sensitized CORT responses. In CTRL animals, 5-HT7-LI was detected in magnocellular and parvocellular subdivisions of PVN and sparsely in adrenal cortex. Exposure to CRS decreased 5-HT7-LI and protein in the PVN, but increased 5-HT7-LI in the adrenal cortex and protein in whole AG. Higher 5-HT and 5-HIAA levels were detected in PVN and AG from CRS animals but 5-HIAA/5-HT ratio increased in AG only. Finally, whereas 5-HT-LI was sparsely observed in the adrenal cortex

  6. Spinal 5-HT7 Receptors and Protein Kinase A Constrain Intermittent Hypoxia-Induced Phrenic Long-term Facilitation

    Science.gov (United States)

    Hoffman, M.S.; Mitchell, G.S.

    2013-01-01

    Phrenic long-term facilitation (pLTF) is a form of serotonin-dependent respiratory plasticity induced by acute intermittent hypoxia (AIH). pLTF requires spinal Gq protein-coupled serotonin-2 receptor (5-HT2) activation, new synthesis of brain-derived neurotrophic factor (BDNF) and activation of its high-affinity receptor, TrkB. Intrathecal injections of selective agonists for Gs protein-coupled receptors (adenosine 2A and serotonin-7; 5-HT7) also induce long-lasting phrenic motor facilitation via TrkB “trans-activation.” Since serotonin release near phrenic motor neurons may activate multiple serotonin receptor subtypes, we tested the hypothesis that 5-HT7 receptor activation contributes to AIH-induced pLTF. A selective 5-HT7 receptor antagonist (SB-269970, 5mM, 12μl) was administered intrathecally at C4 to anesthetized, vagotomized and ventilated rats prior to AIH (3, 5-min episodes, 11% O2). Contrary to predictions, pLTF was greater in SB-269970 treated versus control rats (80±11% vs 45±6% 60 min post-AIH; p<0.05). Hypoglossal LTF was unaffected by spinal 5-HT7 receptor inhibition, suggesting that drug effects were localized to the spinal cord. Since 5-HT7 receptors are coupled to protein kinase A (PKA), we tested the hypothesis that PKA inhibits AIH-induced pLTF. Similar to 5-HT7 receptor inhibition, spinal PKA inhibition (KT-5720, 100μM, 15μl) enhanced pLTF (99±15% 60 min post-AIH; p<0.05). Conversely, PKA activation (8-br-cAMP, 100μM, 15μl) blunted pLTF versus control rats (16±5% vs 45±6% 60 min post-AIH; p<0.05). These findings suggest a novel mechanism whereby spinal Gs protein-coupled 5-HT7 receptors constrain AIH-induced pLTF via PKA activity. PMID:23850591

  7. 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.

  8. Radiosynthesis and in vivo evaluation of novel radioligands for PET imaging of cerebral 5-HT7 receptors

    DEFF Research Database (Denmark)

    Hansen, Hanne D; Herth, Matthias M; Ettrup, Anders

    2014-01-01

    in the living brain. Here, we present the radiosynthesis and in vivo evaluation of Cimbi-712 (3-{4-[4-(4-methylphenyl)piperazine-1-yl]butyl}p-1,3-dihydro-2H-indol-2-one) and Cimbi-717 (3-{4-[4-(3-methoxyphenyl)piperazine-1-yl]butyl}-1,3-dihydro-2H-indol-2-one) as selective 5-HT7R PET radioligands in the pig...

  9. Translation of Novel Serotonin 5-HT7 Agonist Drug Candidates in Rodent Models of Fragile X Syndrome

    Science.gov (United States)

    2016-09-01

    HT1A partial agonist for autism . 6th Cisbio HTRF symposium (Brewster, MA), September 14-17, 2015. Acknowledged DOD funding. Teaching Lectures . 10...grant is to synthesize 5-PAT-type 5HT7 receptor agonists and assess their effectiveness to correct FXS phenotypes in Fmr1-KO mice and other mouse models...President of DELSIA (Delivering Science Innovation for Autism ) and Vice President, Innovative Technologies at Autism Speaks, Daniel Smith, who

  10. Downregulation of 5-HT7 Serotonin Receptors by the Atypical Antipsychotics Clozapine and Olanzapine. Role of Motifs in the C-Terminal Domain and Interaction with GASP-1

    DEFF Research Database (Denmark)

    Manfra, Ornella; Van Craenenbroeck, Kathleen; Skieterska, Kamila

    2015-01-01

    have previously found that the atypical antipsychotics clozapine and olanzapine inhibited G protein activation and, surprisingly, induced both internalization and lysosomal degradation of 5-HT7 receptors. Here, we aimed to determine the mechanism of clozapine- and olanzapine-mediated degradation of 5......-HT7 receptors. In the C-terminus of the 5-HT7 receptor, we identified two YXXΦ motifs, LR residues, and a palmitoylated cysteine anchor as potential sites involved in receptor trafficking to lysosomes followed by receptor degradation. Mutating either of these sites inhibited clozapine- and olanzapine...... of clozapine or olanzapine to the 5-HT7 receptor leads to antagonist-mediated lysosomal degradation by exposing key residues in the C-terminal tail that interact with GASP-1....

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

    International Nuclear Information System (INIS)

    2000-10-01

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

  12. Activation of 5-HT7 receptors reverses NMDA-R-dependent LTD by activating PKA in medial vestibular neurons.

    Science.gov (United States)

    Li, Yan-Hai; Han, Lei; Wu, Kenneth Lap Kei; Chan, Ying-Shing

    2017-09-01

    The medial vestibular nucleus (MVN) is a major output station for neurons that project to the vestibulo-spinal pathway. MVN neurons show capacity for long-term depression (LTD) during the juvenile period. We investigated LTD of MVN neurons using whole-cell patch-clamp recordings. High frequency stimulation (HFS) robustly induced LTD in 90% of type B neurons in the MVN, while only 10% of type A neurons were responsive, indicating that type B neurons are the major contributors to LTD in the MVN. The neuromodulator serotonin (5-HT) is known to modulate LTD in neural circuits of the cerebral cortex and the hippocampus. We therefore aim to determine the action of 5-HT on the LTD of type B MVN neurons and elucidate the relevant 5-HT receptor subtypes responsible for its action. Using specific agonists and antagonists of 5-HT receptors, we found that selective activation of 5-HT 7 receptor in type B neurons in the MVN of juvenile (P13-16) rats completely abolished NMDA-receptor-mediated LTD in a protein kinase A (PKA)-dependent manner. Our finding that 5-HT restricts plasticity of type B MVN neurons via 5-HT 7 receptors offers a mechanism whereby vestibular tuning contributes to the maturation of the vestibulo-spinal circuit and highlights the role of 5-HT in postural control. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. In the search for a lead structure among series of potent and selective hydantoin 5-HT7 R agents: The drug-likeness in vitro study.

    Science.gov (United States)

    Latacz, Gniewomir; Lubelska, Annamaria; Jastrzębska-Więsek, Magdalena; Partyka, Anna; Sobiło, Andrzej; Olejarz, Agnieszka; Kucwaj-Brysz, Katarzyna; Satała, Grzegorz; Bojarski, Andrzej J; Wesołowska, Anna; Kieć-Kononowicz, Katarzyna; Handzlik, Jadwiga

    2017-12-01

    Since the year 1993, when 5-HT 7 receptor (5-HT 7 R) was discovered, there is no selective 5-HT 7 R ligand introduced to the pharmaceutical market. One out of the main reasons disqualifying the 5-HT 7 R ligands is weak drugability properties, including metabolic instability or low permeability. This study is focused on the search of a lead compound by "drug-likeness" estimation of the first series of selective and potent 5-HT 7 R ligands among 5-(4-fluorophenyl)-3-(2-hydroxy-3-(4-aryl-piperazin-1-yl)propyl)-5-methylimidazolidine-2,4-dione derivatives (11-16). The most important drugability parameters, i.e., permeability, metabolic stability, and safety, have been evaluated. The main metabolic pathways were determined. The forced swim test (FST) in mice was performed as a primary in vivo assay for compound 13 and the reference 2. The experiments showed promising drug-like properties for all ligands, with special attention to the benzhydryl (diphenylmethyl) derivative 13. The studies have also indicated in vivo activity of the compound 13 that was observed as a significant and specific antidepressant-like activity in the FST. Taking into account the beneficial properties of 13, i.e., good drug-like parameters, the significant antagonistic action, high selectivity to 5-HT 7 R, and its in vivo antidepressant-like activity, the compound should be considered as a new lead in the search for drugs acting on CNS via 5-HT 7 receptor. © 2017 John Wiley & Sons A/S.

  14. Effects of the selective 5-HT7 receptor antagonist SB-269970 and amisulpride on ketamine-induced schizophrenia-like deficits in rats.

    Directory of Open Access Journals (Sweden)

    Agnieszka Nikiforuk

    Full Text Available A wide body of evidence suggests that 5-HT7 receptors are implicated in a variety of central nervous system functions, including control of learning and memory processes. According to recent preclinical data, the selective blockade of these receptors may be a potential target for cognitive improvement in schizophrenia. The first aim of the present study was to evaluate the effects of the selective 5-HT7 receptor antagonist, SB-269970, and the antipsychotic drug with a high affinity for 5-HT7 receptors, amisulpride, on ketamine-induced deficits in attentional set-shifting and novel object recognition tasks in rats. Because the role of 5-HT7 receptor blockade in ameliorating positive and negative symptoms of schizophrenia remains equivocal, the second aim of these experiments was to examine the effectiveness of SB-269970 and amisulpride in reversing ketamine-induced deficits in prepulse inhibition of the startle reflex and in social interaction test in rats. The study revealed that acute administration of SB-269970 (1 mg/kg or amisulpride (3 mg/kg ameliorated ketamine-induced cognitive inflexibility and novel object recognition deficit in rats. Both compounds were also effective in attenuating ketamine-evoked disruption of social interactions. In contrast, neither SB-269970 nor amisulpride affected ketamine-disrupted prepulse inhibition or 50 kHz USVs accompanying social behaviour. In conclusion, antagonism of 5-HT7 receptors may represent a useful pharmacological approach in the treatment of cognitive deficits and some negative symptoms of schizophrenia.

  15. Antinociception by systemically-administered acetaminophen (paracetamol) involves spinal serotonin 5-HT7 and adenosine A1 receptors, as well as peripheral adenosine A1 receptors.

    Science.gov (United States)

    Liu, Jean; Reid, Allison R; Sawynok, Jana

    2013-03-01

    Acetaminophen (paracetamol) is a widely used analgesic, but its sites and mechanisms of action remain incompletely understood. Recent studies have separately implicated spinal adenosine A(1) receptors (A(1)Rs) and serotonin 5-HT(7) receptors (5-HT(7)Rs) in the antinociceptive effects of systemically administered acetaminophen. In the present study, we determined whether these two actions are linked by delivering a selective 5-HT(7)R antagonist to the spinal cord of mice and examining nociception using the formalin 2% model. In normal and A(1)R wild type mice, antinociception by systemic (i.p.) acetaminophen 300mg/kg was reduced by intrathecal (i.t.) delivery of the selective 5-HT(7)R antagonist SB269970 3μg. In mice lacking A(1)Rs, i.t. SB269970 did not reverse antinociception by systemic acetaminophen, indicating a link between spinal 5-HT(7)R and A(1)R mechanisms. We also explored potential roles of peripheral A(1)Rs in antinociception by acetaminophen administered both locally and systemically. In normal mice, intraplantar (i.pl.) acetaminophen 200μg produced antinociception in the formalin test, and this was blocked by co-administration of the selective A(1)R antagonist DPCPX 4.5μg. Acetaminophen administered into the contralateral hindpaw had no effect, indicating a local peripheral action. When acetaminophen was administered systemically, its antinociceptive effect was reversed by i.pl. DPCPX in normal mice; this was also observed in A(1)R wild type mice, but not in those lacking A(1)Rs. In summary, we demonstrate a link between spinal 5-HT(7)Rs and A(1)Rs in the spinal cord relevant to antinociception by systemic acetaminophen. Furthermore, we implicate peripheral A(1)Rs in the antinociceptive effects of locally- and systemically-administered acetaminophen. Copyright © 2013 Elsevier Ireland Ltd. All rights reserved.

  16. Selective pharmacological blockade of the 5-HT7 receptor attenuates light and 8-OH-DPAT induced phase shifts of mouse circadian wheel running activity

    Directory of Open Access Journals (Sweden)

    Jonathan eShelton

    2015-01-01

    Full Text Available Recent reports have illustrated a reciprocal relationship between circadian rhythm disruption and mood disorders. The 5-HT7 receptor may provide a crucial link between the two sides of this equation since the receptor plays a critical role in sleep, depression, and circadian rhythm regulation. To further define the role of the 5-HT7 receptor as a potential pharmacotherapy to correct circadian rhythm disruptions, the current study utilized the selective 5-HT7 antagonist JNJ-18038683 (10 mg/kg in three different circadian paradigms. While JNJ-18038683 was ineffective at phase shifting the onset of wheel running activity in mice when administered at different circadian time (CT points across the circadian cycle, pretreatment with JNJ-18038683 blocked non-photic phase advance (CT6 induced by the 5-HT1A/7 receptor agonist 8-OH-DPAT (3 mg/kg. Since light induced phase shifts in mammals are partially mediated via the modulation of the serotonergic system, we determined if JNJ-18038683 altered phase shifts induced by a light pulse at times known to phase delay (CT15 or advance (CT22 wheel running activity in free running mice. Light exposure resulted in a robust shift in the onset of activity in vehicle treated animals at both times tested. Administration of JNJ-18038683 significantly attenuated the light-induced phase delay and completely blocked the phase advance. The current study demonstrates that pharmacological blockade of the 5-HT7 receptor by JNJ-18038683 blunts both non-photic and photic phase shifts of circadian wheel running activity in mice. These findings highlight the importance of the 5-HT7 receptor in modulating circadian rhythms. Due to the opposite modulating effects of light resetting between diurnal and nocturnal species, pharmacotherapy targeting the 5-HT7 receptor in conjunction with bright light therapy may prove therapeutically beneficial by correcting the desynchronization of internal rhythms observed in depressed individuals.

  17. Molecular and pharmacological characterization of serotonin 5-HT2α and 5-HT7 receptors in the salivary glands of the blowfly Calliphora vicina.

    Science.gov (United States)

    Röser, Claudia; Jordan, Nadine; Balfanz, Sabine; Baumann, Arnd; Walz, Bernd; Baumann, Otto; Blenau, Wolfgang

    2012-01-01

    Secretion in blowfly (Calliphora vicina) salivary glands is stimulated by the biogenic amine serotonin (5-hydroxytryptamine, 5-HT), which activates both inositol 1,4,5-trisphosphate (InsP(3))/Ca(2+) and cyclic adenosine 3',5'-monophosphate (cAMP) signalling pathways in the secretory cells. In order to characterize the signal-inducing 5-HT receptors, we cloned two cDNAs (Cv5-ht2α, Cv5-ht7) that share high similarity with mammalian 5-HT(2) and 5-HT(7) receptor genes, respectively. RT-PCR demonstrated that both receptors are expressed in the salivary glands and brain. Stimulation of Cv5-ht2α-transfected mammalian cells with 5-HT elevates cytosolic [Ca(2+)] in a dose-dependent manner (EC(50) = 24 nM). In Cv5-ht7-transfected cells, 5-HT produces a dose-dependent increase in [cAMP](i) (EC(50) = 4 nM). We studied the pharmacological profile for both receptors. Substances that appear to act as specific ligands of either Cv5-HT(2α) or Cv5-HT(7) in the heterologous expression system were also tested in intact blowfly salivary gland preparations. We observed that 5-methoxytryptamine (100 nM) activates only the Cv5-HT(2α) receptor, 5-carboxamidotryptamine (300 nM) activates only the Cv5-HT(7) receptor, and clozapine (1 µM) antagonizes the effects of 5-HT via Cv5-HT(7) in blowfly salivary glands, providing means for the selective activation of each of the two 5-HT receptor subtypes. This study represents the first comprehensive molecular and pharmacological characterization of two 5-HT receptors in the blowfly and permits the analysis of the physiological role of these receptors, even when co-expressed in cells, and of the modes of interaction between the Ca(2+)- and cAMP-signalling cascades.

  18. Molecular and pharmacological characterization of serotonin 5-HT2α and 5-HT7 receptors in the salivary glands of the blowfly Calliphora vicina.

    Directory of Open Access Journals (Sweden)

    Claudia Röser

    Full Text Available Secretion in blowfly (Calliphora vicina salivary glands is stimulated by the biogenic amine serotonin (5-hydroxytryptamine, 5-HT, which activates both inositol 1,4,5-trisphosphate (InsP(3/Ca(2+ and cyclic adenosine 3',5'-monophosphate (cAMP signalling pathways in the secretory cells. In order to characterize the signal-inducing 5-HT receptors, we cloned two cDNAs (Cv5-ht2α, Cv5-ht7 that share high similarity with mammalian 5-HT(2 and 5-HT(7 receptor genes, respectively. RT-PCR demonstrated that both receptors are expressed in the salivary glands and brain. Stimulation of Cv5-ht2α-transfected mammalian cells with 5-HT elevates cytosolic [Ca(2+] in a dose-dependent manner (EC(50 = 24 nM. In Cv5-ht7-transfected cells, 5-HT produces a dose-dependent increase in [cAMP](i (EC(50 = 4 nM. We studied the pharmacological profile for both receptors. Substances that appear to act as specific ligands of either Cv5-HT(2α or Cv5-HT(7 in the heterologous expression system were also tested in intact blowfly salivary gland preparations. We observed that 5-methoxytryptamine (100 nM activates only the Cv5-HT(2α receptor, 5-carboxamidotryptamine (300 nM activates only the Cv5-HT(7 receptor, and clozapine (1 µM antagonizes the effects of 5-HT via Cv5-HT(7 in blowfly salivary glands, providing means for the selective activation of each of the two 5-HT receptor subtypes. This study represents the first comprehensive molecular and pharmacological characterization of two 5-HT receptors in the blowfly and permits the analysis of the physiological role of these receptors, even when co-expressed in cells, and of the modes of interaction between the Ca(2+- and cAMP-signalling cascades.

  19. The role of the serotonin receptor subtypes 5-HT1A and 5-HT7 and its interaction in emotional learning and memory

    Directory of Open Access Journals (Sweden)

    Oliver eStiedl

    2015-08-01

    Full Text Available Serotonin (5-hydroxytryptamine, 5-HT is a multifunctional neurotransmitter innervating cortical and limbic areas involved in cognition and emotional regulation. Dysregulation of serotonergic transmission is associated with emotional and cognitive deficits in psychiatric patients and animal models. Drugs targeting the 5-HT system are widely used to treat mood disorders and anxiety-like behaviors. Among the fourteen 5-HT receptor (5-HTR subtypes, the 5-HT1AR and 5-HT7R are associated with the development of anxiety, depression and cognitive function linked to mechanisms of emotional learning and memory. In rodents fear conditioning and passive avoidance (PA are associative learning paradigms to study emotional memory. This review assesses the role of 5-HT1AR and 5-HT7R as well as their interplay at the molecular, neurochemical and behavioral level. Activation of postsynaptic 5-HT1ARs impairs emotional memory through attenuation of neuronal activity, whereas presynaptic 5-HT1AR activation reduces 5-HT release and exerts pro-cognitive effects on PA retention. Antagonism of the 5-HT1AR facilitates memory retention possibly via 5-HT7R activation and evidence is provided that 5HT7R can facilitate emotional memory upon reduced 5-HT1AR transmission. These findings highlight the differential role of these 5-HTRs in cognitive/emotional domains of behavior. Moreover, the results indicate that tonic and phasic 5-HT release can exert different and potentially opposing effects on emotional memory, depending on the states of 5-HT1ARs and 5-HT7Rs and their interaction. Consequently, individual differences due to genetic and/or epigenetic mechanisms play an essential role for the responsiveness to drug treatment, e.g., by SSRIs which increase intrasynaptic 5-HT levels thereby activating multiple pre- and postsynaptic 5-HTR subtypes.

  20. 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

  1. 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

  2. Effects of acupuncture at HT7 on glucose metabolism in a rat model of Alzheimer's disease: an 18F-FDG-PET study

    OpenAIRE

    Lai, Xinsheng; Ren, Jie; Lu, Yangjia; Cui, Shaoyang; Chen, Junqi; Huang, Yong; Tang, Chunzhi; Shan, Baoci; Nie, Bingbing

    2015-01-01

    Objective To explore the effects of acupuncture at HT7 on different cerebral regions in a rat model of Alzheimer's disease (AD) with the application of 18F-2-fluoro-deoxy-D-glucose positron emission tomography (FDG-PET). Methods Sixty Wistar rats were included after undergoing a Y-maze electric sensitivity test. Ten rats were used as a healthy control group. The remaining 50 rats were injected stereotaxically with ibotenic acid into the right nucleus basalis magnocellularis and injected intra...

  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. First measurement of the magnetic turbulence induced Reynolds stress in a tokamak

    International Nuclear Information System (INIS)

    Xu Guosheng; Wan Baonian; Song Mei

    2003-01-01

    Reynolds stress component due to magnetic turbulence was first measured in the plasma edge region of the HT-7 superconducting tokamak using an insertable magnetic probe. A radial gradient of magnetic Reynolds stress was observed to be close to the velocity shear layer location; however, in this experiment its contribution to driving the poloidal flows is small compared to the electrostatic component. The electron heat transport driven by magnetic turbulence is quite small and cannot account for the total energy transport at the plasma edge

  5. Magnetic sensorless control of plasma position and shape in a tokamak

    International Nuclear Information System (INIS)

    Nakamura, K.; Luo, J.R.; Wang, H.Z.

    2005-01-01

    Magnetic sensorless sensing and control experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The sensing is made focusing on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem of integrator of magnetic sensors. Two kinds of control experiments were carried out, to keep the position constant and swing the position in a triangular waveform. And magnetic sensorless sensing of plasma shape is discussed. (author)

  6. 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

  7. Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

    Science.gov (United States)

    Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

    2013-10-01

    The present approach to magnetic fusion which relies on high recycling plasma-wall interaction has exhausted itself at the level of TFTR, JET, JT-60 devices with no realistic path to the burning plasma. Instead, magnetic fusion needs a return to its original idea of insulation of the plasma from the wall, which was the dominant approach in the 1970s and upon implementations has a clear path to the DEMO device with PDT ~= 100 MW and Qelectric > 1 . The SFLiLi systems of this talk is the technology tool for implementation of the guiding idea of magnetic fusion. It utilizes the unique properties of flowing LiLi to pump plasma particles and, thus, insulate plasma from the walls. The necessary flow rate, ~= 1 g3/s, is very small, thus, making the use of lithium practical and consistent with safety requirements. The talk describes how chemical activity of LiLi, which is the major technology challenge of using LiLi in tokamaks, is addressed by SFLiLi systems at the level of already performed (HT-7) experiment, and in ongoing implementations for a prototype of SFLiLi for tokamak divertors and the mid-plane limiter for EAST tokamak (to be tested in the next experimental campaign). This work is supported by US DoE contract No. DE-AC02-09-CH11466.

  8. 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)

  9. Synthesis, radiolabeling and in vivo evaluation of [11C](R)-1-[4-[2-(4-methoxyphenyl)phenyl]piperazin-1-yl]-3-(2-pyrazinyloxy)-2-propanol, a potential PET radioligand for the 5-HT7 receptor

    DEFF Research Database (Denmark)

    Hansen, Hanne Demant; Lacivita, Enza; Di Pilato, Pantaleo

    2014-01-01

    In the search for a novel serotonin 7 (5-HT7) receptor PET radioligand we synthesized and evaluated a new series of biphenylpiperazine derivatives in vitro. Among the studied compounds, (R)-1-[4-[2-(4-methoxyphenyl)phenyl]piperazin-1-yl]-3-(2-pyrazinyloxy)-2-propanol ((R)-16), showed the best com...

  10. 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

  11. 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

  12. Multiscale coherent structures in tokamak plasma turbulence

    International Nuclear Information System (INIS)

    Xu, G. S.; Wan, B. N.; Zhang, W.; Yang, Q. W.; Wang, L.; Wen, Y. Z.

    2006-01-01

    A 12-tip poloidal probe array is used on the HT-7 superconducting tokamak [Li, Wan, and Mao, Plasma Phys. Controlled Fusion 42, 135 (2000)] to measure plasma turbulence in the edge region. Some statistical analysis techniques are used to characterize the turbulence structures. It is found that the plasma turbulence is composed of multiscale coherent structures, i.e., turbulent eddies and there is self-similarity in a relative short scale range. The presence of the self-similarity is found due to the structural similarity of these eddies between different scales. These turbulent eddies constitute the basic convection cells, so the self-similar range is just the dominant scale range relevant to transport. The experimental results also indicate that the plasma turbulence is dominated by low-frequency and long-wavelength fluctuation components and its dispersion relation shows typical electron-drift-wave characteristics. Some large-scale coherent structures intermittently burst out and exhibit a very long poloidal extent, even longer than 6 cm. It is found that these large-scale coherent structures are mainly contributed by the low-frequency and long-wavelength fluctuating components and their presence is responsible for the observations of long-range correlations, i.e., the correlation in the scale range much longer than the turbulence decorrelation scale. These experimental observations suggest that the coexistence of multiscale coherent structures results in the self-similar turbulent state

  13. Assignment of the 5HT7 receptor gene (HTR7) to chromosome 10q and exclusion of genetic linkage with Tourette syndrome

    Energy Technology Data Exchange (ETDEWEB)

    Gelernter, J.; Rao, P.A.; Pauls, D.L. [Yale Univ. School of Medicine, West Haven, CT (United States)] [and others

    1995-03-20

    A novel serotonin receptor designated 5HT7 (genetic locus HTR7) was cloned in 1993. This receptor has interesting properties related to ligand affinity and CNS distribution that render HTR7 a very interesting candidate gene for neuropsychiatric disorders. We mapped this gene, first by physical methods and then by genetic linkage. First, we made a tentative assignment to chromosome 10, based on hybridization of an HTR7 probe to a Southern blot of DNA from somatic cell hybrids. We then identified a genetic polymorphism at the HTR7 locus. We identified one extended pedigree where the polymorphism segregated. Using the LEPED computer program for pairwise linkage analysis, we confirmed the assignment of the gene to chromosome 10, specifically 10q21-q24, based on a lod score of 5.37 at 0% recombination between HTR7 and D10S20 (a chromosome 10 reference marker). Finally, we excluded genetic linkage between this locus and Tourette syndrome under a reasonable set of assumptions. 15 refs., 1 fig., 1 tab.

  14. 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

  15. In search of zonal flows using cross-bispectrum analysis in the boundary plasma of the Hefei Tokamak-7

    International Nuclear Information System (INIS)

    Xu, G.S.; Wan, B.N.; Song, M.

    2002-01-01

    Langmuir probes have been used to measure the electrostatic Reynolds stress and the floating potential fluctuation in the boundary plasma of the Hefei Tokamak-7 (HT-7) [J. Li, B. N. Wan, and J. S. Mao, Plasma Phys. Controlled Fusion 42, 135 (2000)]. The cross bispectrum of r V(tilde sign) θ φ(tilde sign) f > indicates the existence of difference-frequency nonlinear phase coupling and the generation of fluctuations near the geodesic acoustic mode frequency. The inverse cascade process might be linked to the generation of zonal flows by small-scale electrostatic drift-wave turbulence

  16. 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

  17. 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

  18. 5-HT2 and 5-HT7 receptor agonists facilitate plantar stepping in chronic spinal rats through actions on different populations of spinal neurons

    Directory of Open Access Journals (Sweden)

    Urszula eSlawinska

    2014-08-01

    Full Text Available There is considerable evidence from research in neonatal and adult rat and mouse preparations to warrant the conclusion that activation of 5-HT2 and 5-HT1A/7 receptors leads to activation of the spinal cord circuitry for locomotion. These receptors are involved in control of locomotor movements, but it is not clear how they are implicated in the responses to 5-HT agonists observed after spinal cord injury. Here we used agonists that are efficient in promoting locomotor recovery in paraplegic rats, 8-OHDPAT (acting on 5-HT1A/7 receptors and quipazine (acting on 5-HT2 receptors, to examine this issue. Analysis of intra- and interlimb coordination confirmed that the locomotor performance was significantly improved by either drug, but the data revealed marked differences in their mode of action. Interlimb coordination was significantly better after 8-OHDPAT application, and the activity of the extensor soleus muscle was significantly longer during the stance phase of locomotor movements enhanced by quipazine. Our results show that activation of both receptors facilitates locomotion, but their effects are likely exerted on different populations of spinal neurons. Activation of 5-HT2 receptors facilitates the output stage of the locomotor system, in part by directly activating motoneurons, and also through activation of interneurons of the locomotor CPG. Activation of 5-HT7/1A receptors facilitates the activity of the locomotor CPG, without direct actions on the output components of the locomotor system, including motoneurons. Although our findings show that the combined use of these two drugs results in production of well-coordinated weight supported locomotion with a reduced need for exteroceptive stimulation, they also indicate that there might be some limitations to the utility of combined treatment. Sensory feedback and some intraspinal circuitry recruited by the drugs can conflict with the locomotor activation.

  19. 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.

  20. 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

  1. 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.

  2. 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)

  3. 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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. 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

  6. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

  7. 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)

  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. 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)

  11. 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.

  12. 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

  13. 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)

  14. 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

  15. Real time equilibrium reconstruction algorithm in EAST tokamak

    International Nuclear Information System (INIS)

    Wang Huazhong; Luo Jiarong; Huang Qinchao

    2004-01-01

    The EAST (HT-7U) superconducting tokamak is a national project of China on fusion research, with a capability of long-pulse (∼1000 s) operation. In order to realize a long-duration steady-state operation of EAST, some significant capability of real-time control is required. It would be very crucial to obtain the current profile parameters and the plasma shapes in real time by a flexible control system. As those discharge parameters cannot be directly measured, so a current profile consistent with the magnetohydrodynamic equilibrium should be evaluated from external magnetic measurements, based on a linearized iterative least square method, which can meet the requirements of the measurements. The arithmetic that the EFIT (equilibrium fitting code) is used for reference will be given in this paper and the computational efforts are reduced by parameterizing the current profile linearly in terms of a number of physical parameters. In order to introduce this reconstruction algorithm clearly, the main hardware design will be listed also. (authors)

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

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

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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.)

  19. 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.

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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.)

  5. 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

  6. Prepuberal stimulation of 5-HT7-R by LP-211 in a rat model of hyper-activity and attention-deficit: permanent effects on attention, brain amino acids and synaptic markers in the fronto-striatal interface.

    Directory of Open Access Journals (Sweden)

    Lucia A Ruocco

    Full Text Available The cross-talk at the prefronto-striatal interface involves excitatory amino acids, different receptors, transducers and modulators. We investigated long-term effects of a prepuberal, subchronic 5-HT7-R agonist (LP-211 on adult behaviour, amino acids and synaptic markers in a model for Attention-Deficit/Hyperactivity Disorder (ADHD. Naples High Excitability rats (NHE and their Random Bred controls (NRB were daily treated with LP-211 in the 5th and 6th postnatal week. One month after treatment, these rats were tested for indices of activity, non selective (NSA, selective spatial attention (SSA and emotionality. The quantity of L-Glutamate (L-Glu, L-Aspartate (L-Asp and L-Leucine (L-Leu, dopamine transporter (DAT, NMDAR1 subunit and CAMKIIα, were assessed in prefrontal cortex (PFC, dorsal (DS and ventral striatum (VS, for their role in synaptic transmission, neural plasticity and information processing. Prepuberal LP-211 (at lower dose reduced horizontal activity and (at higher dose increased SSA, only for NHE but not in NRB rats. Prepuberal LP-211 increased, in NHE rats, L-Glu in the PFC and L-Asp in the VS (at 0.250 mg/kg dose, whereas (at 0.125 mg/kg dose it decreased L-Glu and L-Asp in the DS. The L-Glu was decreased, at 0.125 mg/kg, only in the VS of NRB rats. The DAT levels were decreased with the 0.125 mg/kg dose (in the PFC, and increased with the 0.250 mg/kg dose (in the VS, significantly for NHE rats. The basal NMDAR1 level was higher in the PFC of NHE than NRB rats; LP-211 treatment (at 0.125 mg/kg dose decreased NMDAR1 in the VS of NRB rats. This study represents a starting point about the impact of developmental 5-HT7-R activation on neuro-physiology of attentive processes, executive functions and their neural substrates.

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  1. 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. Natural current profiles in tokamaks

    International Nuclear Information System (INIS)

    Biskamp, D.

    1986-01-01

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

  3. Alcator C-Mod Tokamak

    Data.gov (United States)

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

  4. JUST: Joint Upgraded Spherical Tokamak

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  5. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.; Saramad, S.

    2012-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

  6. New directions in tokamak reactors

    International Nuclear Information System (INIS)

    Baker, C.C.

    1985-01-01

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

  7. The Tokamak IST-TOK

    International Nuclear Information System (INIS)

    Varandas, C.A.F.; Cabral, J.A.C.; Manso, M.E.

    1991-01-01

    A small tokamak is under construction at the Portuguese Technical Superior Institute. The main objective is to create a home based laboratory in which an independent scientific program might be developed. (L.C.J.A.). 14 refs, 6 figs

  8. 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

  9. 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)

  10. 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

  11. 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)

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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)

  2. 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

  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. 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

  5. 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)

  6. 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)

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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)

  17. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  18. Large aspect ratio tokamak study

    International Nuclear Information System (INIS)

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

    1979-01-01

    The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured

  19. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

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

  20. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

    1981-02-01

    Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

  1. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1994-01-01

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  2. 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

  3. 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

  4. 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.)

  5. 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 ...

  6. 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

  7. 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

  8. 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.

  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. 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.

  13. 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

  14. 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

  15. 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

  16. Microwave Tokamak Experiment: Overview and status

    International Nuclear Information System (INIS)

    1990-05-01

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs

  17. Combined confinement system applied to tokamaks

    International Nuclear Information System (INIS)

    Ohkawa, Tihiro

    1986-01-01

    From particle orbit point of view, a tokamak is a combined confinement configuration where a closed toroidal volume is surrounded by an open confinement system like a magnetic mirror. By eliminating a cold halo plasma, the energy loss from the plasma becomes convective. The H-mode in diverted tokamaks is an example. Because of the favorable scaling of the energy confinement time with temperature, the performance of the tokamak may be significantly improved by taking advantage of this effect. (author)

  18. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  19. Improvement of the tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, L

    1994-12-31

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

  20. Advanced commercial Tokamak optimization studies

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  1. Flux driven turbulence in tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Ghendrih, P.; Ottaviani, M.; Sarazin, Y.; Beyer, P.; Benkadda, S.; Waltz, R.E.

    1999-01-01

    This work deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts, are usually observed over a broad range of time and spatial scales. The existence of these fronts provide a way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyroBohm type in spite of these large scale transport events. Some departure from the gyroBohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux than at fixed temperature gradient, with the same time averaged profile. (author)

  2. Options for an ignited tokamak

    International Nuclear Information System (INIS)

    Sheffield, J.

    1984-02-01

    It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon β/sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed

  3. Plasma diagnostics on large tokamaks

    International Nuclear Information System (INIS)

    Orlinskij, D.V.; Magyar, G.

    1988-01-01

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

  4. Starfire: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

    1979-01-01

    The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

  5. Comprehensive numerical modelling of tokamaks

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  6. Magnetic island formation in tokamaks

    International Nuclear Information System (INIS)

    Yoshikawa, S.

    1989-04-01

    The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs

  7. 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)

  8. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

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

  9. Runaway electrons during tokamak startup

    International Nuclear Information System (INIS)

    Sharma, A.S.; Jayakumar, R.

    1988-01-01

    Runaway electrons significantly affect the plasma and impurity evolution during tokamak startup. During its rise, a runaway pulse stores magnetic flux inductively; this is then released during the decay phase of the runaway pulse. This process affects plasma formation, current initiation and current buildup. Because of their relativistic velocities the runaway electrons have higher ionization and excitation rates than the plasma electrons. This leads to a significant modification of the impurity behaviour and consequently the plasma evolution. (author). 20 refs, 8 figs

  10. Minimum scaling laws in tokamaks

    International Nuclear Information System (INIS)

    Zhang, Y.Z.; Mahajan, S.M.

    1986-10-01

    Scaling laws governing anomalous electron transport in tokamaks with ohmic and/or auxiliary heating are derived using renormalized Vlasov-Ampere equations for low frequency electromagnetic microturbulence. It is also shown that for pure auxiliary heating (or when auxiliary heating power far exceeds the ohmic power), the energy confinement time scales as tau/sub E/ ∼ P/sub inj//sup -1/3/, where P/sub inj/ is the injected power

  11. 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

  12. Tokamak plasma boundary layer model

    International Nuclear Information System (INIS)

    Volkov, T.F.; Kirillov, V.D.

    1983-01-01

    A model has been developed for the limiter layer and for the boundary region of the plasma column in a tokamak to facilitate analytic calculations of the thickness of the limiter layers, the profiles and boundary values of the temperature and the density under various conditions, and the difference between the electron and ion temperatures. This model can also be used to analyze the recycling of neutrals, the energy and particle losses to the wall and the limiter, and other characteristics

  13. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Kieras, C.E.

    1982-12-01

    Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma

  14. Discharge cleaning for a tokamak

    International Nuclear Information System (INIS)

    Ishii, Shigeyuki

    1983-01-01

    Various methods of discharge cleaning for tokamaks are described. The material of the first walls of tokamaks is usually stainless steel, inconel, titanium and so on. Hydrogen is exclusively used as the discharge gas. Glow discharge cleaning (GDC), Taylor discharge cleaning (TDC), and electron cyclotron resonance discharge cleaning (ECR-DC) are discussed in this paper. The cleaning by GDC is made by moving a movable anode to the center of a tokamak vassel. Taylor found the good cleaning effect of induced discharge by high pressure and low power discharge. This is called TDC. When the frequency of high frequency discharge in a magnetic field is equal to that of the electron cyclotron resonance, the break down potential is lowered if the pressure is sufficiently low. The ECR-CD is made by using this effect. In TDC and ECR-DC, the electron temperature, which has a close relation to the production rate of H 0 , can be controlled by the pressure. In GDC, the operating pressure was improved by the radio frequency glow (RG) method. However, there is still the danger of arcing. In case of GDC and ECR-DC, the position of plasma can be controlled, but not in case of TDC. The TDC is accepted by most of takamak devices in the world. (Kato, T.)

  15. 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

  16. Magnetic confinement by Tokamak: physical aspects

    International Nuclear Information System (INIS)

    Tachon, J.

    1980-01-01

    After describing the Tokamak configuration concept, the author provides an analysis of the principal physical aspects of this type of installation and concludes by estimating that the Tokamak concept is a 'plausible candidate' as a means of producing controlled thermonuclear fusion [fr

  17. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    Reid, R.L.; Steiner, D.

    1977-01-01

    This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  18. Simulation of a major tokamak disruption

    International Nuclear Information System (INIS)

    White, R.B.; Monticello, D.A.; Rosenbluth, M.N.

    1977-08-01

    It is known that the internal tokamak disruption leads to a current profile which is flattened inside the surface where the safety factor equals unity. It is shown that such a profile can lead to m = 2 magnetic islands which grow to fill a substantial part of the tokamak cross section in a time consistent with the observations of the major disruption

  19. Diagnostics for the Rijnhuizen Tokamak Project

    NARCIS (Netherlands)

    Donne, A. J. H.

    1994-01-01

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

  20. MAST: a Mega Amp Spherical Tokamak

    International Nuclear Information System (INIS)

    Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.

    1995-01-01

    The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)

  1. Mercier criterion for high-β tokamaks

    International Nuclear Information System (INIS)

    Galvao, R.M.O.

    1984-01-01

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

  2. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  3. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    Im, K.H.; Cho, S.; Her, N.I.

    2001-01-01

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  4. 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

  5. Cluster storage for COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír

    2012-01-01

    Roč. 87, č. 12 (2012), s. 2238-2241 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006

  6. Surface tearing modes in tokamaks

    International Nuclear Information System (INIS)

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

    1985-10-01

    Surface tearing modes in tokamaks are studied numerically and analytically. The eigenvalue problem is solved to obtain the growth rate and the mode structure. We investigate in detail dependences of the growth rate of the m/n = 2/1 resistive MHD modes on the safety factor at the plasma surface, current profile, wall position, and resistivity. The surface tearing mode moves the plasma surface even when the wall is close to the surface. The stability diagram for these modes is presented. (author)

  7. Major disruption process in tokamak

    International Nuclear Information System (INIS)

    Kurita, Gen-ichi; Azumi, Masafumi; Tuda, Takashi; Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji; Itoh, Kimitaka; Takeda, Tatsuoki

    1981-11-01

    The major disruption in a cylindrical tokamak is investigated by using the multi-helicity code, and the destabilization of the 3/2 mode by the mode coupling with the 2/1 mode is confirmed. The evolution of the magnetic field topology caused by the major disruption is studied in detail. The effect of the internal disruption on the 2/1 magnetic island width is also studied. The 2/1 magnetic island is not enhanced by the flattening of the q-profile due to the internal disruption. (author)

  8. Nuclear fusion research at Tokamak Energy Ltd

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  9. 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

  10. CAT-D-T tokamaks

    International Nuclear Information System (INIS)

    Greenspan, E.; Blue, T.; Miley, G.H.

    1981-01-01

    The domains of plasma fuel cycles bounded by the D-T and Cat-D, and by the D-T and SCD modes of operation are examined. These domains, referred to as, respectively, the Cat-D-T and SCD-T modes of operation, are characterized by the number (γ) of tritons per fusion neutron available from external (to the plasma) sources. Two external tritium sources are considered - the blankets of the Cat-D-T (SCD-T) reactors and fission reactors supported by the Cat-D-T (SCD-T) driven hybrid reactors. It is found that by using 6 Li for the active material of the control elements of the fission reactors, it is possible to achieve γ values close to unity. Cat-D-T tokamaks could be designed to have smaller size, higher power density, lower magnetic field and even lower plasma temperature than Cat-D tokamaks; the difference becomes significant for γ greater than or equal to .75. The SCD-T mode of operation appears to be even more attractive. Promising applications identified for these Cat-D-T and SCD-T modes of operation include hybrid reactors, fusion synfuel factories and fusion reactors which have difficulty in providing all their tritium needs

  11. Plasma position control in TCABR Tokamak

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  12. Estimation of Zeff in Novillo Tokamak

    International Nuclear Information System (INIS)

    Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.

    1996-01-01

    We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)

  13. Global gyrokinetic simulation of tokamak transport

    International Nuclear Information System (INIS)

    Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T.

    1998-10-01

    A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or η i (η i ≡ ∂ ell nT i /∂ ell n n i ) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling

  14. Fast IR diodes thermometer for tokamak

    International Nuclear Information System (INIS)

    Chen Xiangbo

    2001-01-01

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

  15. 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

  16. Power and particle exhaust in tokamaks

    International Nuclear Information System (INIS)

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER's nominal design positions; important directions for further research are identified

  17. Definition of total bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    Ross, D.W.

    1995-01-01

    Alternative definitions of the total bootstrap current are compared. An analogous comparison is given for the ohmic and auxiliary currents. It is argued that different definitions than those usually employed lead to simpler analyses of tokamak operating scenarios

  18. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    1981-01-01

    Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

  19. Neutral beam in ALVAND IIC tokamak

    International Nuclear Information System (INIS)

    Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

    1992-01-01

    Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

  20. 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

  1. 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)

  2. Research into controlled fusion in tokamaks

    International Nuclear Information System (INIS)

    Zacek, F.

    1992-01-01

    During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

  3. 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.

  4. 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

  5. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  6. Theory of incremental turbulent transport in tokamaks

    International Nuclear Information System (INIS)

    Similon, P.L.

    1991-01-01

    The goal of this research is to understand how the various aspect of turbulent transport operate in tokamaks, in the presence of low frequency fluctuations such as drift waves or trapped electron modes

  7. Numerical simulation of edge plasma in tokamak

    International Nuclear Information System (INIS)

    Chen Yiping; Qiu Lijian

    1996-02-01

    The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)

  8. Plasma diagnostics using synchrotron radiation in tokamaks

    International Nuclear Information System (INIS)

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

    1995-09-01

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

  9. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br

    2002-07-01

    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 historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  10. 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.)

  11. Erosion of the first wall of Tokamaks

    International Nuclear Information System (INIS)

    Guseva, M.I.; Ionova, E.S.; Martynenko, Yu.V.

    1980-01-01

    An estimate of the rate of erosion of the wall due to sputtering and blistering requires knowledge of the fluxes and energies of the particles which go from the plasma to the wall, of the sputtering coefficients S, and of the erosion coefficients S* for blistering. The overall erosion coefficient is equal to the sum of the sputtering coefficient and the erosion coefficient for blistering. Here the T-20 Tokamak is examined as an example of a large-scale Tokamak. 18 refs

  12. Advanced tokamak physics in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-12-01

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

  13. Optimization design for SST-1 Tokamak insulators

    International Nuclear Information System (INIS)

    Zhang Yuanbin; Pan Wanjiang

    2012-01-01

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

  14. Spherical tokamak without external toroidal fields

    International Nuclear Information System (INIS)

    Kaw, P.K.; Avinash, K.; Srinivasan, R.

    2001-01-01

    A spherical tokamak design without external toroidal field coils is proposed. The tokamak is surrounded by a spheromak shell carrying requisite force free currents to produce the toroidal field in the core. Such equilibria are constructed and it is indicated that these equilibria are likely to have robust ideal and resistive stability. The advantage of this scheme in terms of a reduced ohmic dissipation is pointed out. (author)

  15. Preliminary results of the TBR small tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Fagundes, A.N.; Da Silva, R.P.; Galvao, R.M.O.; Del Bosco, E.; Vuolo, J.H.; Sanada, E.K.; Dellaqua, R.

    1982-01-01

    The paper gives a short description of the TBR - small Brazilian tokamak and the first results obtained for plasma formation and equilibrium. Measured breakdown curves for hydrogen are shown to be confined within analytically calculated limits and to depend strongly on stray vertical magnetic fields. Time profiles of plasma current in equilibrium are shown and compared with the predictions of a simple analytical model for tokamak discharges. Reasonable agreement is obtained taking Zsub(eff) as a free parameter. (author)

  16. Full power in the European tokamak

    International Nuclear Information System (INIS)

    Lallia, P.P.; Hugon, M.

    1987-01-01

    A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years

  17. Facility approach to tokamak operation

    International Nuclear Information System (INIS)

    Edmonds, P.H.; Gabbard, W.A.

    1981-01-01

    In anticipation of the appearance of more advanced tokamaks and other fusion relevant experiments, program has been established at ORNL to systemically identify the requirements of an effective machine operations group. This program is presently applied to the ISX-B experiment. With its continuing development, it is expected to provide major support in the identification of potential problem areas and to assist in the generation of the necessary procedures for forthcoming devices. The present and future generations of large plasma devices will function as facilities, operated by an operations group as service to the plasma physicists and diagnosticians. The purpose of the program discussed here is to develop and to encourage an orderly transition to the facility-like style of operation

  18. Developments in tokamak transport modeling

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Attenberger; Lao, L.L.

    1981-01-01

    A variety of numerical methods for solving the time-dependent fluid transport equations for tokamak plasmas is presented. Among the problems discussed are techniques for solving the sometimes very stiff parabolic equations for particle and energy flow, treating convection-dominated energy transport that leads to large cell Reynolds numbers, optimizing the flow of a code to reduce the time spent updating the particle and energy source terms, coupling the one-dimensional (1-D) flux-surface-averaged fluid transport equations to solutions of the 2-D Grad-Shafranov equation for the plasma geometry, handling extremely fast transient problems such as internal MHD disruptions and pellet injection, and processing the output to summarize the physics parameters over the potential operating regime for reactors. Emphasis is placed on computational efficiency in both computer time and storage requirements

  19. Simulation models for tokamak plasmas

    International Nuclear Information System (INIS)

    Dimits, A.M.; Cohen, B.I.

    1992-01-01

    Two developments in the nonlinear simulation of tokamak plasmas are described: (A) Simulation algorithms that use quasiballooning coordinates have been implemented in a 3D fluid code and a 3D partially linearized (Δf) particle code. In quasiballooning coordinates, one of the coordinate directions is closely aligned with that of the magnetic field, allowing both optimal use of the grid resolution for structures highly elongated along the magnetic field as well as implementation of the correct periodicity conditions with no discontinuities in the toroidal direction. (B) Progress on the implementation of a likeparticle collision operator suitable for use in partially linearized particle codes is reported. The binary collision approach is shown to be unusable for this purpose. The algorithm under development is a complete version of the test-particle plus source-field approach that was suggested and partially implemented by Xu and Rosenbluth

  20. 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.

  1. Tokamak plasma interaction with limiters

    International Nuclear Information System (INIS)

    Pitcher, C.S.

    1987-11-01

    The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D 2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict

  2. Economic considerations of commercial tokamak options

    International Nuclear Information System (INIS)

    Dabiri, A.E.

    1986-05-01

    Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m 2 , which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m 2 , will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e)

  3. D-D tokamak reactor assessment

    International Nuclear Information System (INIS)

    Baxter, D.C.; Dabiri, A.E.

    1983-01-01

    A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

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

    International Nuclear Information System (INIS)

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

    2014-03-01

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

  5. 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

  6. Analysis of tokamak plasma confinement modes using the fast

    Indian Academy of Sciences (India)

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

  7. Comparative studies of stellarator and tokamak transport

    Energy Technology Data Exchange (ETDEWEB)

    Stroth, U; Burhenn, R; Geiger, J; Giannone, L.; Hartfuss, H J; Kuehner, G; Ledl, L; Simmet, E E; Walter, H [Max-Planck-Inst. fuer Plasmaphysik, IPP-Euratom Association, Garching (Germany); ECRH Team; W7-AS Team

    1997-09-01

    Transport properties in the W7-AS stellarator and in tokamaks are compared. The parameter dependences and the absolute values of the energy confinement time are similar. Indications are found that the density dependence, which is usually observed in stellarator confinement, can vanish above a critical density. The density dependence in stellarators seems to be similar to that in the linear ohmic confinement regime, which, in small tokamaks, extends to high density values, too. Because of the similarity in the gross confinement properties, transport in stellarators and tokamaks should not be dominated by the parameters which are very different in the two concepts, i.e. magnetic shear, major rational values of the rotational transform and plasma current. A difference in confinement is that there exists evidence for pinches in the particle and, possibly, energy transport channels in tokamaks whereas in stellarators no pinches have been observed, so far. In order to study the effect of plasma current and toroidal electric fields, stellarator discharges were carried out with an increasing amount of plasma current. From these experiments, no clear evidence of a connection of pinches with these parameters is found. The transient response in W7-AS plasmas can be described in terms of a non-local model. As in tokamaks, also cold pulse experiments in W7-AS indicate the importance of non-local transport. (author). 8 refs, 5 figs.

  8. 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

  9. Edge plasma diagnostics on Tore Supra tokamak

    International Nuclear Information System (INIS)

    Fujita, Junji

    1991-01-01

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

  10. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J A

    1994-11-30

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

  11. Neoclassical MHD equations for tokamaks

    International Nuclear Information System (INIS)

    Callen, J.D.; Shaing, K.C.

    1986-03-01

    The moment equation approach to neoclassical-type processes is used to derive the flows, currents and resistive MHD-like equations for studying equilibria and instabilities in axisymmetric tokamak plasmas operating in the banana-plateau collisionality regime (ν* approx. 1). The resultant ''neoclassical MHD'' equations differ from the usual reduced equations of resistive MHD primarily by the addition of the important viscous relaxation effects within a magnetic flux surface. The primary effects of the parallel (poloidal) viscous relaxation are: (1) Rapid (approx. ν/sub i/) damping of the poloidal ion flow so the residual flow is only toroidal; (2) addition of the bootstrap current contribution to Ohm's laws; and (3) an enhanced (by B 2 /B/sub theta/ 2 ) polarization drift type term and consequent enhancement of the perpendicular dielectric constant due to parallel flow inertia, which causes the equations to depend only on the poloidal magnetic field B/sub theta/. Gyroviscosity (or diamagnetic vfiscosity) effects are included to properly treat the diamagnetic flow effects. The nonlinear form of the neoclassical MHD equations is derived and shown to satisfy an energy conservation equation with dissipation arising from Joule and poloidal viscous heating, and transport due to classical and neoclassical diffusion

  12. Alfven Eigenmodes in spherical tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, Mikhail P.; Sharapov, Sergei E.; Berk, Herbert L.; Pinches, Simon D.

    2005-01-01

    Electromagnetic instabilities are often excited by fast super-Alfvenic ions produced by neutral beam injection (NBI) in plasmas of the spherical tokamaks START and MAST (toroidal magnetic confinement devices in which the minor a and major R 0 radii of the torus are comparable, R 0 /a≅1.2/1.8). These instabilities are seen as discrete weakly-damped toroidal and elliptical Alfven Eigenmodes (TAEs and EAEs) with frequencies tracing in time the Alfven scaling with the equilibrium magnetic field and plasma density, or as energetic particle modes (EPMs) whose frequencies don't start from TAE-frequency and sweep down in time faster than the equilibrium parameters change. In some discharges the beam drives Aflvenic-type modes that start from the TAE frequency and sweep in both up- and down- directions. Such electromagnetic perturbations are interpreted as 'hole-clump' long-living nonlinear fluctuations of the fast ion distribution function predicted by Berk-Breizman-Petviashvili [Phys. Lett. A238 (1998) 408]. It is found on both START and MAST that the Alfven instabilities weaken in their mode amplitude and in the number of unstable modes as the pressure of the thermal plasma increases, in agreement with increased thermal ion Landau damping and the pressure effect on core-localised TAEs. (author)

  13. 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

  14. 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

  15. Progress of the ECH·ECCD experiments. Research progress of the ECH·ECCD experiments in tokamaks and spherical tokamaks

    International Nuclear Information System (INIS)

    Isayama, Akihiko; Tanaka, Hitoshi

    2009-01-01

    Recent progress in the ECH·ECCD study in tokamak and spherical tokamak devices is described. As for the tokamak study, results on the control of neoclassical tearing modes and sawtooth oscillations, the current profile, the internal transport barrier, the plasma start-up and the discharge cleaning are given. As for the spherical tokamak study, the plasma start-up by ECH·ECCD and the electron-Bernstein-wave heating and the current drive are described. (T.I.)

  16. 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

  17. 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.

  18. Heat load material studies: Simulated tokamak disruptions

    International Nuclear Information System (INIS)

    Gahl, J.M.; McDonald, J.M.; Zakharov, A.; Tserevitinov, S.; Barabash, V.; Guseva, M.

    1991-01-01

    It is clear that an improved understanding of the effects of tokamak disruptions on plasma facing component materials is needed for the ITER program. very large energy fluxes are predicted to be deposited in ITER and could be very damaging to the machine. During 1991, Sandia National Laboratories and the University of New Mexico conducted cooperative tokamak disruption simulation experiments at several Soviet facilities. These facilities were located at the Efremov Institute in Leningrad, the Kurchatov Atomic Energy Institute (Troisk and Moscow) and the Institute for Physical Chemistry of the Soviet Adademy of Sciences in Moscow. Erosion of graphite from plasma stream impact is seen to be much less than that observed with laser or electron beams with similar energy fluxes. This, along with other data obtained, seem to suggest that the ''vapor shielding'' effect is a very important phenomenon in the study of graphite erosion during tokamak disruption

  19. 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)

  20. Spherical tokamak power plant design issues

    International Nuclear Information System (INIS)

    Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.

    2000-01-01

    The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed

  1. Time - resolved thermography at Tokamak T-10

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  2. Magnet design considerations for Tokamak fusion reactors

    International Nuclear Information System (INIS)

    Purcell, J.R.; Chen, W.; Thomas, R.

    1976-01-01

    Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)

  3. 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

  4. Activation analysis of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Selcow, E.C.

    1986-01-01

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak

  5. Helicity content and tokamak applications of helicity

    International Nuclear Information System (INIS)

    Boozer, A.H.

    1986-05-01

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

  6. Proposed tokamak poloidal field system development program

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, J.D.; Vogel, H.F.; Warren, R.W.; Weldon, D.M.

    1977-05-01

    A program is proposed to develop poloidal field components for TNS and EPR size tokamak devices and to test these components in realistic circuits. Emphasis is placed upon the development of the most difficult component, the superconducting ohmic-heating coil. Switches must also be developed for testing the coils, and this switching technology is to be extended to meet the requirements for the large scale tokamaks. Test facilities are discussed; power supplies, including a homopolar to drive the coils, are considered; and poloidal field systems studies are proposed.

  7. 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.

  8. Ballooning stable high beta tokamak equilibria

    International Nuclear Information System (INIS)

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

    1981-04-01

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

  9. Energy storage for tokamak reactor cycles

    International Nuclear Information System (INIS)

    Buchanan, C.H.

    1979-01-01

    The inherent characteristic of a tokamak reactor requiring periodic plasma quench and reignition introduces the problem of energy storage to permit continuous electrical output to the power grid. The cycle under consideration in this paper is a 1000 second burn followed by a 100 second reignition phase. The physical size of a typical toroidal plasma reaction chamber for a tokamak reactor has been described earlier. The thermal energy storage requirements described in this reference will serve as a basis for much of the ensuing discussion

  10. Tokamak power systems studies, FY 1985

    International Nuclear Information System (INIS)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs

  11. A Fast Shutdown Technique for Large Tokamaks

    International Nuclear Information System (INIS)

    Fredrickson, E.; Schmidt, G.L.; Hill, K.; Jardin, S.C.

    1999-01-01

    A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR

  12. Two-ion ICRF heating in Tokamaks

    International Nuclear Information System (INIS)

    Tennfors, E.

    1985-03-01

    The practical consequences for tokamak plasma heating in the ion cyclotron frequency regime of the two-dimensional treatment of the two-ion mode conversion layer are analyzed. The problem of evaluation of the condition for fast wave resonance is analyzed, as well as the limitations imposed by warm plasma effects. Simple ways to find the mode conversion surfaces when they exist are presented. Also for large tokamaks, it is possible to obtain mode conversion conditions for realistic antenna spectra provided species concentration and frequency are chosen such that the surface Epsilon = 0 intersects the plasma midplane just outside of the magnetic axis. (Author)

  13. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    Downum, W.B.

    1981-01-01

    The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

  14. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  15. 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

  16. 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

  17. Periodic disruptions in the MT-1 tokamak

    International Nuclear Information System (INIS)

    Zoletnik, S.

    1988-11-01

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

  18. Gas blanket fueling of a tokamak reactor

    International Nuclear Information System (INIS)

    Gralnick, S.L.

    1978-01-01

    The purpose of this paper is a speculative investigation of the potential of fueling a Tokamak by introducing a sufficiently large quantity of gaseous deuterium and tritium at the vacuum wall boundary. It is motivated by two factors: current generation tokamaks are, in a manner of speaking, fueled from the edge quite successfully as is evidenced by pulse lengths that are long compared to particle recycling times, and by rapid plasma density increase produced by gas puffing, alternative, deep penetration fueling techniques that have been proposed possess severe technological problems and large costs

  19. Electronic system of TBR tokamak device

    International Nuclear Information System (INIS)

    Silva, R.P. da.

    1980-01-01

    The electronics developed as a part of the TBR project, which involves the construction of a small tokamak at the Physics Institute of the University of Sao Paulo, is described. On the basis of tokamak parameter values, the electronics for the toroidal field, ohmic/heating and vertical field systems is presented, including capacitors bank, switches, triggering circuits and power supplies. A controlled power oscilator used in discharge cleaning and pre-ionization is also described. The performance of the system as a function of the desired plasma parameters is discussed. (Author) [pt

  20. 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

  1. Alfven wave heating in a tokamak reactor

    International Nuclear Information System (INIS)

    Borg, G.G.; Appert, K.; Knight, A.J.; Lister, J.B.; Vaclavik, J.

    1990-01-01

    A number of features of Alfven wave heating make it potentially attractive for use in large tokamak reactors. Among them are the availability and relativity low cost of the power supplies, the potential ability to act selectively on the current profile, and the probable absence of operational limits in size, fields or density. The physics of Alfven wave heating in a large tokamak is assessed. Present theoretical understanding of mode coupling and antenna loading is extrapolated to a large machine. The problem of a recessed antenna is analysed. Calculations of loading and discussion of various heating scenarios for the particular case of NET are also presented. (author). 23 refs, 18 figs, 4 tabs

  2. 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

  3. Increase in beta limit in tokamak plasmas

    International Nuclear Information System (INIS)

    Kamada, Yutaka

    2003-01-01

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

  4. Development of Operation Scenario for Spherical Tokamak at SNU

    International Nuclear Information System (INIS)

    Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.

    2009-01-01

    Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed

  5. Tokamak startup: problems and scenarios related to the transient phases of ignited tokamak operations

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma

  6. A need for non-tokamak approaches to magnetic fusion energy

    International Nuclear Information System (INIS)

    Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

    1992-01-01

    Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

  7. Tokamak startup with electron cyclotron heating

    International Nuclear Information System (INIS)

    Holly, D.J.; Prager, S.C.; Shepard, D.A.; Sprott, J.C.

    1980-04-01

    Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed

  8. Transformer Recharging with Alpha Channeling in Tokamaks

    International Nuclear Information System (INIS)

    Fisch, N.J.

    2009-01-01

    Transformer recharging with lower hybrid waves in tokamaks can give low average auxiliary power if the resistivity is kept high enough during the radio frequency (rf) recharging stage. At the same time, operation in the hot ion mode via alpha channeling increases the effective fusion reactivity. This paper will address the extent to which these two large cost saving steps are compatible.

  9. Criteria for initiation of tokamak disruptions

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  10. Feedback stabilization of axisymmetric modes in tokamaks

    International Nuclear Information System (INIS)

    Jardin, S.C.; Larrabee, D.A.

    1982-01-01

    Noncircular tokamak plasmas can be unstable to ideal MHD axisymmetric instabilities. Passive conductors with finite resistivity will at best slow down these instabilities to the resistive (L/R) time of the conductors. An active feedback system far from the plasma which responds on this resistive time can stabilize the system provided its mutual inductance with the passive coils is small enough

  11. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

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

    1996-01-01

    We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

  12. Scattering measurements in Tokamak type devices

    International Nuclear Information System (INIS)

    Matoba, Tohru

    1975-03-01

    Theories, experiments and proposals for light scattering in Tokamak type devices are reviewed. Thomson scattering, measuring method of the current density distribution by scattering and resonance fluorescence are summarily described. These methods may be useful for diagnosis of the fusion plasmas. The report may help planning of the measuring apparatus for the fusion plasmas in future. (auth.)

  13. Maintenance considerations of the STARFIRE commercial tokamak

    International Nuclear Information System (INIS)

    Trachsel, C.A.; Zahn, H.S.; Field, R.E.; Stevens, H.C.

    1979-01-01

    This paper presents the maintenance approach, the commercial tokamak design features that enhance maintenance and preliminary repair time and required mean-time-between-failures for major subsystems. Reactor hall building and maintenance equipment requirements including hot cells, coil rewinding, and cranes are discussed

  14. 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

  15. Tokamak power plant burn cycle options

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1994-06-01

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

  16. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1980-11-01

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  17. Preliminary measurements on Tokamak KT-5

    International Nuclear Information System (INIS)

    Wen Yizhi; Wan Shude; Rong Furui; Haan Shengshen; Liu Wandong; Liu Lei

    1987-01-01

    A small tokamak, KT-5, has been put in to operation since 1984. The major and minor radius of the plasma are 30 and 4.5 cm, respectively. The parameters obtained in the first phase of KT-5 experiments are as follows B t = 0.45 T, I p ≥ 5 kA, q(α) σ = 50 eV

  18. 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)

  19. MHD stability of vertically asymmetric tokamak equilibria

    International Nuclear Information System (INIS)

    Dalhed, H.E.; Grimm, R.C.; Johnson, J.L.

    1981-03-01

    The ideal MHD stability properties of a special class of vertically asymmetric tokamak equilibria are examined. The calculations confirm that no major new physical effects are introduced and the modifications can be understood by conventional arguments. The results indicate that significant departures from up-down symmetry can be tolerated before the reduction in β becomes important for reactor operation

  20. [High beta tokamak research and plasma theory

    International Nuclear Information System (INIS)

    1990-01-01

    Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report

  1. Computation of tokamak equilibria with steady flow

    International Nuclear Information System (INIS)

    Kerner, W.; Tokuda, Shinji

    1987-08-01

    The equations for ideal MHD equilibria with stationary flow are reexamined and addressed as numerically applied to tokamak configurations with a free plasma boundary. Both the isothermal (purely toroidal flow) and the poloidal flow cases are treated. Experiment-relevant states with steady flow (so far only in the toroidal direction) are computed by the modified SELENE40 code. (author)

  2. Ion diagnostics for the tokamak boundary

    International Nuclear Information System (INIS)

    Matthews, G.F.

    1991-01-01

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

  3. Anomalous periodic disruptions in tokamak plasma

    International Nuclear Information System (INIS)

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

    1982-09-01

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

  4. Energy confinement of high-density tokamaks

    NARCIS (Netherlands)

    Schüller, F.C.; Schram, D.C.; Coppi, B.; Sadowski, W.

    1977-01-01

    Neoclassical ion heat conduction is the major energy loss mechanism in the center of an ohmically heated high-d. tokamak discharge (n>3 * 1020 m-3). This fixes the mutual dependence of plasma quantities on the axis and leads to scaling laws for the poloidal b and energy confinement time, given the

  5. Supravodivý tokamak dobyl Asii

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2006-01-01

    Roč. 54, č. 18 (2006), s. 58 ISSN 0040-1064 Institutional research plan: CEZ:AV0Z20430508 Keywords : superconducting tokamak * ITER * Tore Supra * Institute of Plasma Physics AV CR Subject RIV: BL - Plasma and Gas Discharge Physics

  6. Runaway electrons in the TRIAM-1 tokamak

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  7. Runaway electrons in the TRIAM-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-09-01

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

  8. Plasma position control in SST1 tokamak

    Indian Academy of Sciences (India)

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

  9. Central control system for the EAST tokamak

    International Nuclear Information System (INIS)

    Sun Xiaoyang; Ji Zhenshan; Wu Yicun; Luo Jiarong

    2008-01-01

    The architecture, the main function and the design scheme of the central control system and the collaboration system of EAST tokamak are described. The main functions of the central control system are to supply a union control interface for all the control, diagnoses, and data acquisition (DAQ) subsystem and it is also designed to synchronize all those subsystem. (authors)

  10. Microinstabilities in weak density gradient tokamak systems

    International Nuclear Information System (INIS)

    Tang, W.M.; Rewoldt, G.; Chen, L.

    1986-04-01

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient

  11. 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.)

  12. Modular pulse sequencing in a tokamak system

    International Nuclear Information System (INIS)

    Chew, A.C.; Lee, S.; Saw, S.H.

    1992-01-01

    Pulse technique applied in the timing and sequencing of the various part of the MUT tokamak system are discussed. The modular architecture of the pulse generating device highlights the versatile application of the simple physical concepts in precise and complicated research experiment. (author)

  13. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    HUMPHREYS, D.A.; FERRON, J.R.; JOHNSON, R.D; LEUER, J.A.; PENAFLOR, B.G.; WALKER, M.L.; WELANDER, A.S.; KHAYRUTDINOV, R.R; DOKOUKA, V.; EDGELL, D.H.; FRANSSON, C.M.

    2004-03-01

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance

  14. Tokamak startup with electron cyclotron heating

    Energy Technology Data Exchange (ETDEWEB)

    Holly, D J; Prager, S C; Shepard, D A; Sprott, J C

    1980-04-01

    Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed.

  15. The ARIES tokamak fusion reactor study

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  16. Plea for stellarator funding raps tokamaks

    International Nuclear Information System (INIS)

    Blake, M.

    1992-01-01

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

  17. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1983-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

  18. An overview on plasma disruption mitigation and avoidance in tokamak

    International Nuclear Information System (INIS)

    He Kaihui; Pan Chuanhong; Feng Kaiming

    2002-01-01

    Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  20. Disruption generated secondary runaway electrons in present day tokamaks

    International Nuclear Information System (INIS)

    Pankratov, I.M.; Jaspers, R.

    2000-01-01

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

  1. Tokamak residual zonal flow level in near-separatrix region

    International Nuclear Information System (INIS)

    Bing-Ren, Shi

    2010-01-01

    Residual zonal flow level is calculated for tokamak plasmas in the near-separatrix region of a diverted tokamak. A recently developed method is used to construct an analytic divertor tokamak configuration. It is shown that the residual zonal flow level becomes smaller but still keeps finite near the separatrix because the neoclassical polarisation mostly due to the trapped particles goes larger in this region. (fluids, plasmas and electric discharges)

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

    International Nuclear Information System (INIS)

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

    1992-03-01

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

  3. Summary of the 1982 small tokamak users meeting

    International Nuclear Information System (INIS)

    Sprott, J.C.

    1982-11-01

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

  4. Numerical studies of edge localized instabilities in tokamaks

    International Nuclear Information System (INIS)

    Wilson, H.R.; Snyder, P.B.; Huysmans, G.T.A.; Miller, R.L.

    2002-01-01

    A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code

  5. Start of the international tokamak physics activity

    International Nuclear Information System (INIS)

    Campbell, D.

    2001-01-01

    This newsletter comprises a summary on the start of the International Tokamak Physics activity (ITPA) by Dr. D. Campbell, Chair of the ITPA Co-ordinating Committee. As the ITER EDA drew to a close, it became clear that it was desirable to establish a new mechanism in order to promote the continued development of the physics basis for burning plasma experiments and to preserve the invaluable collaborations between the major international fusion communities which had been established through the ITER physics expert groups. As a result of the discussions of the representatives of the European Union, Japan, the Russian Federation and the United States the agreed principles for conducting the International Tokamak Physics Activity (ITPA) were elaborated and ITPA topical physics groups were organized

  6. Assembly study for JT-60SA tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Shibanuma, K., E-mail: shibanuma.kiyoshi@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Arai, T.; Hasegawa, K.; Hoshi, R.; Kamiya, K.; Kawashima, H.; Kubo, H.; Masaki, K.; Saeki, H.; Sakurai, S.; Sakata, S.; Sakasai, A.; Sawai, H.; Shibama, Y.K.; Tsuchiya, K.; Tsukao, N.; Yagyu, J.; Yoshida, K.; Kamada, Y. [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Mizumaki, S. [Toshiba Corporation, Minato-ku, Tokyo 105-8001 (Japan); and others

    2013-10-15

    The assembly scenarios and assembly tools of the major tokamak components for JT-60SA are studied in the following. (1) The assembly frame (with a dedicated 30-tonne crane), which is located around the JT-60SA tokamak, is adopted for effective assembly works in the torus hall and the temporary support of the components during assembly. (2) Metrology for precise positioning of the components is also studied by defining the metrology points on the components. (3) The sector segmentation for weld joints and positioning of the vacuum vessel (VV), the assembly scenario and tools for VV thermal shield (TS), the connection of the outer intercoil structure (OIS) and the installation of the final toroidal field coil (TFC) are studied, as typical examples of the assembly scenarios and tools for JT-60SA.

  7. Module description of TOKAMAK equilibrium code MEUDAS

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Masaei; Hayashi, Nobuhiko; Matsumoto, Taro; Ozeki, Takahisa [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-01-01

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  8. 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

  9. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1981-01-01

    The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and USSR. The Zero-Phase of the INTOR Workshop, which was conducted during 1979, assessed the technical data base that would support the construction of the next major device in the tokamak program to operate in the early 1990s and defined the objectives and characteristics of this device. The INTOR workshop was extended into phase-1, the Definition Phase, in early 1980. The objective of the Phase-1 Workshop was to develop a conceptual design of the INTOR experiment. The purpose of this paper is to give an overview of the work of the Phase-1 INTOR Workshop (January 1980-June 1981, with emphasis upon the conceptual design

  10. Magnetohydrodynamic stability of tokamak edge plasmas

    International Nuclear Information System (INIS)

    Connor, J.W.; Hastie, R.J.; Wilson, H.R.; Miller, R.L.

    1998-01-01

    A new formalism for analyzing the magnetohydrodynamic stability of a limiter tokamak edge plasma is developed. Two radially localized, high toroidal mode number n instabilities are studied in detail: a peeling mode and an edge ballooning mode. The peeling mode, driven by edge current density and stabilized by edge pressure gradient, has features which are consistent with several properties of tokamak behavior in the high confinement open-quotes Hclose quotes-mode of operation, and edge localized modes (or ELMs) in particular. The edge ballooning mode, driven by the pressure gradient, is identified; this penetrates ∼n 1/3 rational surfaces into the plasma (rather than ∼n 1/2 , expected from conventional ballooning mode theory). Furthermore, there exists a coupling between these two modes and this coupling provides a picture of the ELM cycle

  11. 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

  12. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.

    2011-01-01

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  13. The tokamak - an imperfect frame of refernce

    International Nuclear Information System (INIS)

    Schmitter, K.U.

    1981-03-01

    It is attempted to assess the suitability of tokamaks for fusion power plants on the basis of existing design studies by reference to the reality of energy production in fission power plants. A definition of suitability criteria and a discussion of their relation to the most important features of power plants are followed by a comparative treatment. For example, the mean volumetric net electric power density in the nuclear islands of tokamak power plant designs is only 2,5 to 4 E of the value common today in light water reactor nuclear islands. In addition, configuration problems, auxiliary power requirements and energy payback time are discussed and taken into account in the assessment. (orig.)

  14. Tokamak Fusion Core Experiment maintenance study

    International Nuclear Information System (INIS)

    Snyder, A.M.; Watts, K.D.

    1985-01-01

    The recently completed Tokamak Fusion Core Experiment (TFCX) design project was carried out to investigate potential next generation tokamak concepts. An important aspect of this project was the early development and incorporation of remote maintainability throughout the design process. This early coordination and incorporation of maintenance aspects to the design of the device and facilities would assure that the machine could ultimately be maintained and repaired in an efficient and cost effective manner. To meet this end, a rigorously formatted engineering trade study was performed to determine the preferred configuration for the TFCX reactor based primarily on maintenance requirements. The study indicated that the preferred design was one with an external vacuum vessel and torrodial field coils that could be removed via a simple radial motion. The trade study is presented and the preferred TFCX configuration is described

  15. 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

  16. KTM Tokamak operation scenarios software infrastructure

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  17. 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)

  18. 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.

  19. Runaway acceleration during magnetic reconnection in tokamaks

    International Nuclear Information System (INIS)

    Helander, P; Eriksson, L-G; Andersson, F

    2002-01-01

    In this paper, the basic theory of runaway electron production is reviewed and recent progress is discussed. The mechanisms of primary and secondary generation of runaway electrons are described and their dynamics during a tokamak disruption is analysed, both in a simple analytical model and through numerical Monte Carlo simulation. A simple criterion for when these mechanisms generate a significant runaway current is derived, and the first self-consistent simulations of the electron kinetics in a tokamak disruption are presented. Radial cross-field diffusion is shown to inhibit runaway avalanches, as indicated in recent experiments on JET and JT-60U. Finally, the physics of relativistic post-disruption runaway electrons is discussed, in particular their slowing down due to emission of synchrotron radiation, and their ability to produce electron-positron pairs in collisions with bulk plasma ions and electrons

  20. 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)

  1. Currents in the DIII-D Tokamak

    Science.gov (United States)

    Azari, A.; Eidietis, N. W.

    2012-10-01

    Loss of vertical control of an elongated tokamak plasma results in a vertical displacement event (VDE) which can induce large currents on open field lines and exert high JxB forces on in-vessel components. An array of first-wall tile current monitors on DIII-D provides direct measurement of the poloidal halo currents. These measurements are analyzed to create a database of halo current magnitude and asymmetry, which are found to lie within the ranges seen by numerous other tokamaks in the ITPA Disruption Database. In addition, an analysis of halo asymmetry rotation is presented, as rotation at the resonance frequencies of in-vessel components could lead to significant amplification of the halo forces. Halo current rotation is found to be far more prevalent in old (1997-2002) DIII-D halo current data than recent data (2009), perhaps due to a change in divertor geometry over that time.

  2. Lower hybrid current drive in shaped tokamaks

    International Nuclear Information System (INIS)

    Kesner, J.

    1993-01-01

    A time dependent lower hybrid current drive tokamak simulation code has been developed. This code combines the BALDUR tokamak simulation code and the Bonoli/Englade lower hybrid current drive code and permits the study of the interaction of lower hybrid current drive with neutral beam heating in shaped cross-section plasmas. The code is time dependent and includes the beam driven and bootstrap currents in addition to the current driven by the lower hybrid system. Examples of simulations are shown for the PBX-M experiment which include the effect of cross section shaping on current drive, ballooning mode stabilization by current profile control and sawtooth stabilization. A critical question in current drive calculations is the radial transport of the energetic electrons. The authors have developed a response function technique to calculate radial transport in the presence of an electric field. The consequences of the combined influences of radial diffusion and electric field acceleration are discussed

  3. 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.

  4. Boundary Plasma Turbulence Simulations for Tokamaks

    International Nuclear Information System (INIS)

    Xu, X.; Umansky, M.; Dudson, B.; Snyder, P.

    2008-05-01

    The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T e ; T i ) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics

  5. 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

  6. Module description of TOKAMAK equilibrium code MEUDAS

    International Nuclear Information System (INIS)

    Suzuki, Masaei; Hayashi, Nobuhiko; Matsumoto, Taro; Ozeki, Takahisa

    2002-01-01

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  7. Turbulence and abnormal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Garbet, X.

    1988-09-01

    Microinstabilities in linear and nonlinear tokamak plasmas were studied. A variational method based on the existence of a system of angular variables and action for the charged particles in the magnetic configuration of a tokamak is described. The corresponding functional, extremal in relation to the fluctuating electromagnetic field, is calculated analytically, taking into account the effects of the toroidal geometry. A numerical code, TORRID, was derived from these principles and the main instabilities, especially ion instabilities and microtearing, were studied linearly. Nonlinear methods were also applied to microtearing. Quasi-linear transport coefficients are derived from a principle of minimum entropy production. Thermal ionic conductivity and viscosity are calculated for an ionic turbulence [fr

  8. 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

  9. Constrained ripple optimization of Tokamak bundle divertors

    International Nuclear Information System (INIS)

    Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.

    1983-02-01

    Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ω B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple ( 0 ) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded

  10. A review of lifetime analyses for tokamaks

    International Nuclear Information System (INIS)

    Harkness, S.D.; Cramer, B.

    1979-01-01

    System studies have shown that economic fusion power can best be achieved from the use of long lived components. The stresses generated in a first wall module are a complex function of its geometry, the chosen structural material and the tokamak burn cycle characteristics. A means of applying ASME Code Case 1592 to preliminary design has been established. Methods of incorporating some of the material property changes expected from irradiation are discussed. Cyclic stresses imposed by tokamak operation are expected to cause fatigue related properties to govern the life of the structure. Stress assisted bubble growth is also discussed. This may be the critical mechanism in establishing the creep rupture life of a fusion first wall component. (orig.)

  11. 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

  12. 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

  13. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  14. Resonant helical fields in the TBR tokamak

    International Nuclear Information System (INIS)

    Bender, O.W.

    1986-01-01

    The influence of external resonant helical fields (RHF) in the tokamak TBR plasma discharges was investigated. These fields were created by helical windings wounded on the TBR vessel with the same helicity of rational magnetic surfaces, producing resonant efects on these surfaces. The characteristics of the MHZ activity (amplitude, frequency and poloidal and toroidal wave numbers, m=2,3,4 and n=1, respectively) during the plasma discharges were modified by eletrical winding currents of the order of 2% of the plasma current. These characterisitics were measured for diferent discharges safety factors at the limiter (q) between 3 and 4, with and without the RHF, with the atenuation of the oscillation amplitudes and the increasing of their frequencies. The existente of expontaneous and induced magnetic islands were investigated. The data were compared with results obtained in other tokamaks. (author) [pt

  15. User's manual of Tokamak Simulation Code

    International Nuclear Information System (INIS)

    Nakamura, Yukiharu; Nishino, Tooru; Tsunematsu, Toshihide; Sugihara, Masayoshi.

    1992-12-01

    User's manual for use of Tokamak Simulation Code (TSC), which simulates the time-evolutional process of deformable motion of axisymmetric toroidal plasma, is summarized. For the use at JAERI computer system, the TSC is linked with the data management system GAEA. This manual is forcused on the procedure for the input and output by using the GAEA system. Model equations to give axisymmetric motion, outline of code system, optimal method to get the well converged solution are also described. (author)

  16. 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

  17. 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

  18. Discrete Alfven waves in the TORTUS tokamak

    International Nuclear Information System (INIS)

    Amagishi, Y.; Ballico, M.J.; Cross, R.C.; Donnely, I.J.

    1989-01-01

    Discrete Alfven Waves (DAWs) have been observed as antenna resistance peaks and as enhanced edge fields in the TORTUS tokamak during Alfven wave heating experiments. A kinetic theory code has been used to calculate the antenna loading and the structure of the DAW fields for a range of plasma current and density profiles. There is fair agreement between the measured and predicted amplitude of the DAW fields in the plasma edge when both are normalized to the same antenna power

  19. Fusion technology applications of the spherical tokamak

    International Nuclear Information System (INIS)

    Robinson, D.C.; Akers, R.; Allfrey, S.J.

    1999-01-01

    Fusion technology applications of the spherical tokamak are presented, exploiting its high β capability, normal conducting TF coils, compact core, high natural elongation, disruption resilience and low capital cost. We concentrate here on two particular applications: a volume neutron source (VNS) for component testing and a power plant, addressing engineering and physics issues for steady state operation. The prospect of nearer term burning plasma ST devices are discussed in the conclusions. (author)

  20. Fusion technology applications of the spherical tokamak

    International Nuclear Information System (INIS)

    Robinson, D.C.; Akers, R.; Allfrey, S.J.

    2001-01-01

    Fusion technology applications of the spherical tokamak are presented, exploiting its high β capability, normal conducting TF coils, compact core, high natural elongation, disruption resilience and low capital cost. We concentrate here on two particular applications: a volume neutron source (VNS) for component testing and a power plant, addressing engineering and physics issues for steady state operation. The prospect of nearer term burning plasma ST devices are discussed in the conclusions. (author)

  1. 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

  2. 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

  3. Neoclassical tearing modes in a tokamak

    International Nuclear Information System (INIS)

    Hahm, T.S.

    1988-08-01

    Linear tearing instability is studied in the banana collisionality regime in tokamak geometry. Neoclassical effects produce significant modifications of Ohm's law and the vorticity equation so that the growth rate of tearing modes driven by Δ' is dramatically reduced compared to the usual resistive MHD value. Consequences of this result, regarding the presence of pressure-gradient-driven neoclassical resistive interchange instabilities and the evolution of magnetic islands in the Rutherford regime, are discussed. 10 refs

  4. Development of Atomic Beam Probe for tokamaks

    Czech Academy of Sciences Publication Activity Database

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

    2013-01-01

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

  5. 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

  6. 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

  7. 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

  8. Confinement scaling and ignition in tokamaks

    International Nuclear Information System (INIS)

    Perkins, F.W.; Sun, Y.C.

    1985-10-01

    A drift wave turbulence model is used to compute the scaling and magnitude of central electron temperature and confinement time of tokamak plasmas. The results are in accord with experiment. Application to ignition experiments shows that high density (1 to 2) . 10 15 cm -3 , high field, B/sub T/ > 10 T, but low temperature T approx. 6 keV constitute the optimum path to ignition

  9. 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

  10. 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

  11. Tokamak with liquid metal toroidal field coil

    International Nuclear Information System (INIS)

    Ohkawa, T.; Schaffer, M.J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof

  12. High beta plasmas in the PBX tokamak

    International Nuclear Information System (INIS)

    Bol, K.; Buchenauer, D.; Chance, M.

    1986-04-01

    Bean-shaped configurations favorable for high β discharges have been investigated in the Princeton Beta Experiment (PBX) tokamak. Strongly indented bean-shaped plasmas have been successfully formed, and beta values of over 5% have been obtained with 5 MW of injected neutral beam power. These high beta discharges still lie in the first stability regime for ballooning modes, and MHD stability analysis implicates the external kink as responsible for the present β limit

  13. Physics parameter space of tokamak ignition devices

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  14. 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

  15. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2005-01-01

    SST-1, a steady state superconducting tokamak, is undergoing commissioning tests at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in a tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. The auxiliary current drive is based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. Detailed commissioning tests on the cryogenic system and experiments on the hydraulic characters and cool down features of single TF coils have been completed prior to the cool down of the entire superconducting system. Results of the single TF magnet cool down, and testing of the magnet system are presented. First experiments related to the breakdown and the current ramp up will subsequently be carried out. (author)

  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. Design of the ITER tokamak assembly tools

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyunki [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)], E-mail: hkpark@nfri.re.kr; Lee, Jaehyuk; Kim, Taehyung [SFA Engineering Corp., 42-7 Palyong-dong, Changwon-si, Gyeongsangnam-do 641-847 (Korea, Republic of); Song, Yunju [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of); Im, Kihak [ITER Organization, CEA Cadarasche, 13108 Saint Paul-lez-Durance (France); Kim, Byungchul; Lee, Hyeongon; Jung, Ki-Jung [National Fusion Research Institute, 52 Eoeun-Dong, Yuseong-Gu, Daejon 305-333 (Korea, Republic of)

    2008-12-15

    ITER tokamak assembly is mainly composed of lower cryostat activities, sector sub-assembly, sector assembly, in-vessel activities and ex-vessel activities. The main tools for sector sub-assembly procedures consists of upending tool, sector lifting tool, vacuum vessel support and bracing tool and sector sub-assembly tool. Conceptual design of assembly tools for sector sub-assembly procedures is described herein. The basic structure for upending tool has been developed under the assumption that upending is performed with crane which will be installed in Tokamak building. Sector lifting tool is designed to adjust the position of a sector to minimize the difference between the center of the tokamak building crane and the center of gravity of the sector. Sector sub-assembly tool is composed of special frame for the fine adjustment of position control with 6 degrees of freedom. The design of VV support and bracing tool for four kinds of VV 40 deg. sectors has been developed. Also, structural analysis for upending tool, sector sub-assembly tool has been studied using ANSYS for the situation of an applied load with the same dead weight multiplied by 3/4. The results of structural analyses for these tools were below the allowable values.

  18. The spheric tokamak programme at Culham

    International Nuclear Information System (INIS)

    Sykes, A.

    1999-01-01

    The Spherical Tokamak (ST) is the low aspect ratio limit of the conventional tokamak, and appears to offer attractive physics properties in a simpler device. The START (Small Tight Aspect Ratio Tokamak) experiment provided the world's first demonstration of the properties of hot plasmas in an ST configuration, and was operational at Culham from January 1991 to March 1998, obtaining plasma current of up to 300 kA and pulse durations of ∼ 50 ms. Its successor, MAST is scheduled to obtain first plasma in Autumn 1998 and is a purpose built, high vacuum machine designed to have a tenfold increase in plasma volume with plasma currents up to 2 MA. Current drive and heating will be by a combination of induction-compression as on START, a high-performance central solenoid, 1.5 MW ECRH and 5 MW of Neutral Beam Injection. The promising results from START are reviewed, and the many challenges posed for the next generation of purpose-built STs (such as MAST) are described. (author)

  19. Dust limit management strategy in tokamaks

    International Nuclear Information System (INIS)

    Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Delaporte, P.; Douai, D.; Garnier, D.; Gauthier, E.; Gulden, W.; Hong, S.H.; Pitcher, S.; Rodriguez, L.; Taylor, N.; Tesini, A.; Vartanian, S.; Vatry, A.; Wykes, M.

    2009-01-01

    Dust is produced in tokamaks by the interaction between the plasma and the plasma facing components. Dust has not yet been of a major concern in existing tokamaks mainly because the quantity is small and these devices are not nuclear facilities. However, in ITER and in future reactors, it will represent operational and potential safety issues. From a safety point of view, in order to control the potential dust hazard, the current ITER strategy is based on a defense in depth approach designed to provide reliable confinement systems, to avoid failures, and to measure and minimise the dust inventory. In addition, R and D is put in place for optimisation of the proposed methods, such as improvement of measurement, dust cleaning and the reduction of dust production. The aim of this paper is to present the approach for the control of the dust inventory, relying on the monitoring of envelope values and the development of removal techniques already developed in the existing tokamaks or plasma dedicated devices or which will need further research and development in order to be integrated in ITER.

  20. Accessibility of high β tokamak states

    International Nuclear Information System (INIS)

    Hogan, J.T.

    1978-05-01

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

  1. Dust limit management strategy in tokamaks

    Science.gov (United States)

    Rosanvallon, S.; Grisolia, C.; Andrew, P.; Ciattaglia, S.; Delaporte, P.; Douai, D.; Garnier, D.; Gauthier, E.; Gulden, W.; Hong, S. H.; Pitcher, S.; Rodriguez, L.; Taylor, N.; Tesini, A.; Vartanian, S.; Vatry, A.; Wykes, M.

    2009-06-01

    Dust is produced in tokamaks by the interaction between the plasma and the plasma facing components. Dust has not yet been of a major concern in existing tokamaks mainly because the quantity is small and these devices are not nuclear facilities. However, in ITER and in future reactors, it will represent operational and potential safety issues. From a safety point of view, in order to control the potential dust hazard, the current ITER strategy is based on a defense in depth approach designed to provide reliable confinement systems, to avoid failures, and to measure and minimise the dust inventory. In addition, R&D is put in place for optimisation of the proposed methods, such as improvement of measurement, dust cleaning and the reduction of dust production. The aim of this paper is to present the approach for the control of the dust inventory, relying on the monitoring of envelope values and the development of removal techniques already developed in the existing tokamaks or plasma dedicated devices or which will need further research and development in order to be integrated in ITER.

  2. 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)

  3. Initial DEMO tokamak design configuration studies

    Energy Technology Data Exchange (ETDEWEB)

    Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

    2015-10-15

    Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

  4. 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

  5. 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.

  6. Super high field ohmically heated tokamak operation

    International Nuclear Information System (INIS)

    Cohn, D.R.; Bromberg, L.; Leclaire, R.J.; Potok, R.E.; Jassby, D.L.

    1986-01-01

    The authors discuss a super high field mode of tokamak operation that uses ohmic heating or near ohmic heating to ignition. The super high field mode of operation uses very high values of Β/sup 2/α, where Β is the magnetic field and a is the minor radius (Β/sup 2/α > 100 T/sup 2/m). We analyze copper magnet devices with major radii from 1.7 to 3.0 meters. Minimizing or eliminating the need for auxiliary heating has the potential advantages of reducing uncertainty in extrapolating the energy confinement time of current tokamak devices, and reducing engineering problems associated with large auxiliary heating requirements. It may be possible to heat relatively short pulse, inertially cooled tokamaks to ignition with ohmic power alone. However, there may be advantages in using a very small amount of auxiliary power (less than the ohmic heating power) to boost the ohmic heating and provide a faster start-up, expecially in relatively compact devices

  7. The ICRH tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Perkins, F.W.

    1976-01-01

    A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

  8. On the density limit of Tokamaks

    International Nuclear Information System (INIS)

    Lehnert, B.

    1982-12-01

    Under the conditions of so far performed quasi-steady tokamak experiments near the density limit, the plasma pressure gradient in the outer layers of the plasma body becomes mainly determined by the plasma-neutral gas balance. An earlier analysis of ballooning instabilities driven by this gradient in regions of bad curvature has been extended to deduce an explicit stability criterion which determines the density limit. This criterion is closely related to the empirical Murakami limit. At relevant tokamak data, the deduced limit becomes proportional to J(sub)zR(sup)1/2 where J(sub)z is the average current density and R the major plasma radius. It is further found to be independent of the toroidal magnetic field strength and anomalous transport, as well as to be a slow function of the outer layer temperature and the mass number. The deduced stability criterion is consistent with so far performed experiments. Provided that the present analysis can be extrapolated to a wider range of parameter data and be combined with Alcator scaling, conditions near ignition appear to become realizable in small tokamaks by ohmic heating alone. These conditions can be satisfied at relevant magnetic field strengths and plasma currents, by imposing a high plasma current density. (author)

  9. 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

  10. The design of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, G.S.; Kim, J.; Hwang, S.M.

    1999-01-01

    The Korea superconducting tokamak advanced research (KSTAR) project is the major effort of the Korean national fusion program (KNFP) to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. Major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 Tesla, and plasma current 2 MA with a strongly shaped plasma cross-section and double-null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beam, ion cyclotron waves, lower hybrid waves, and electron-cyclotron waves for flexible profile control. A comprehensive set of diagnostics is planned for plasma control and performance evaluation and physics understanding. The project has completed its conceptual design phase and moved to the engineering design phase. The target date of the first plasma is set for year 2002. (orig.)

  11. Current drive by spheromak injection into a tokamak

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  12. 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

  13. Recent progress on the Compact Ignition Tokamak (CIT)

    International Nuclear Information System (INIS)

    Ignat, D.W.

    1987-01-01

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

  14. Tokamak WEST připraven ke startu!

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    Květen (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * WEST * Tora Supra * divertor Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/2014-tokamak-west-pripraven-ke- start u

  15. Compact fusion energy based on the spherical tokamak

    Science.gov (United States)

    Sykes, A.; Costley, A. E.; Windsor, C. G.; Asunta, O.; Brittles, G.; Buxton, P.; Chuyanov, V.; Connor, J. W.; Gryaznevich, M. P.; Huang, B.; Hugill, J.; Kukushkin, A.; Kingham, D.; Langtry, A. V.; McNamara, S.; Morgan, J. G.; Noonan, P.; Ross, J. S. H.; Shevchenko, V.; Slade, R.; Smith, G.

    2018-01-01

    Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor magnets as a possible route to fusion power using relatively small devices. We present an overview of the development programme including details of the enabling technologies, the key modelling methods and results, and the remaining challenges on the path to compact fusion.

  16. Commercial feasibility of fusion power based on the tokamak concept

    International Nuclear Information System (INIS)

    Reid, R.L.; Steiner, D.

    1977-01-01

    The impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants is determined. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  17. Lower hybrid heating experiments in tokamaks: an overview

    International Nuclear Information System (INIS)

    Porkolab, M.

    1985-10-01

    Lower hybrid wave propagation theory relevant to heating fusion grade plasmas (tokamaks) is reviewed. A brief discussion of accessibility, absorption, and toroidal ray propagation is given. The main part of the paper reviews recent results in heating experiments on tokamaks. Both electron and ion heating regimes will be discussed. The prospects of heating to high temperatures in reactor grade plasmas will be evaluated

  18. 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)

  19. 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

  20. Physics design requirements for the Tokamak Physics Experiment (TPX)

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem)

    2013-01-01

    htmlabstractOne of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma

  2. Design of Tokamak plasma with high Tc superconducting coils

    International Nuclear Information System (INIS)

    Uchimoto, T.; Miya, K.; Yoshida, Y.; Yamada, T.

    1999-01-01

    This paper presents a design of tokamak plasma in light of how the small ignited tokamak is possible with use of the HTSC coils as plasma stabilizer. The same data base and formulas as ITER are here used and any innovative technology other than the HTSC stabilizing coils is not assumed. (author)

  3. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-08-01

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

  4. Tokamak Physics Experiment (TPX) power supply design and development

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  5. High performance operational limits of tokamak and helical systems

    International Nuclear Information System (INIS)

    Yamazaki, Kozo; Kikuchi, Mitsuru

    2003-01-01

    The plasma operational boundaries of tokamak and helical systems are surveyed and compared with each other. Global confinement scaling laws are similar and gyro-Bohm like, however, local transport process is different due to sawtooth oscillations in tokamaks and ripple transport loss in helical systems. As for stability limits, achievable tokamak beta is explained by ideal or resistive MHD theories. On the other hand, beta values obtained so far in helical system are beyond ideal Mercier mode limits. Density limits in tokamak are often related to the coupling between radiation collapse and disruptive MHD instabilities, but the slow radiation collapse is dominant in the helical system. The pulse length of both tokamak and helical systems is on the order of hours in small machines, and the longer-pulsed good-confinement plasma operations compatible with radiative divertors are anticipated in both systems in the future. (author)

  6. Physics design of an ultra-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  7. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

    Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.

    2002-10-01

    A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)

  8. The physics of an ignited tokamak

    International Nuclear Information System (INIS)

    Troyon, F.

    1990-10-01

    There appears to be a consensus that time has come to embark on the design and construction of the next generation of tokamaks which is at the origin of the ITER initiative. Different proposals have been made based on different appreciation as to the size of the step which can be taken, related to considerations of cost, risk and duration of construction. A class of devices which may be considered the last the very high-field, high density ALCATOR-Frascati line of tokamaks have been proposed for some years specifically for this purpose. Today there remain three such projects: Ignitor, Ignitex and CIT. The technology chosen limits the pulse length to a few seconds. These devices have evolved through the years becoming larger and much more expensive than originally anticipated, increasing the pressure to do more than just a simple demonstration of ignition. There is another class of more ambitious devices which aim at creating long burning plasmas in conditions as close as possible to those of a tokamak reactor in order to address all the plasma physics problems associated with long burn. Three such projects, NET, the european next step after JET, ITER and JIT are good examples of this approach. The ideal would be to design a device with sufficient margin to study burning plasmas over a wide range of parameters. The object of this didactic presentation is to describe the common physics basis of all these projects, compare their expected performance using present knowledge and list the physics problems associated with a burning plasma experiment. The comparison is not meant to be a judgement since the important parameter is the cost/benefit ratio which is a matter of appreciation at this stage. 6 refs., 3 figs., 1 tab

  9. 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

  10. Dynamics of fast ions in Tokamaks

    International Nuclear Information System (INIS)

    Helander, P.

    1994-01-01

    Fast ions play a prominent role in the heating of tokamak plasmas by, e.g. neutral-beam injection, ion-cyclotron-resonance heating, and alpha-particle heating. In this thesis, a number of physical and mathematical problems concerning the dynamics of fast ions in tokamaks are addressed. First, the motion under adiabatic perturbations is studied. The frequencies of instabilities excited in tokamaks sometimes vary slowly with time. The existence of an adiabatic invariant of particle motion in such circumstances is shown to lead to a rapid convection of particles in the radial direction. Generalized adiabatic invariants are constructed for systems where the slowly varying parameter is subjected to small, but rapidly varying, fluctuations. Second, the onset of stochastic motion under resonant perturbations is considered. It is shown that the finite width of fast-ion drift orbits significantly affects the threshold for stochastic motion caused by magnetic field ripple or ion-cyclotron-resonance heating. Finite-orbit-width effects are also shown to reduce the strength of resonant interaction between alpha particles and internal kink modes. Third, the diffusive motion in the stochastic regime is analysed mathematically. Monte Carlo operators for the motion on long time-scales are constructed, and the validity of the quasilinear diffusion coefficient is examined. Finally, the effects of close ion collisions are investigated. It is demonstrated that close encounters with fast ions produce a high-energy tail in the distribution functions of impurity ions, and that close collisions between fusion-generated alpha particles give rise to a population of such particles with energies extending up to twice the birth energy. 44 refs

  11. 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%.

  12. 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%

  13. Neutral beam injection on the PLT tokamak

    International Nuclear Information System (INIS)

    Schilling, G.; Ashcroft, D.L.; Eubank, H.P.; Grisham, L.R.; Knauer, R.C.; Stewart, L.D.; Stooksberry, R.W.; Ulrickson, M.; Williams, M.D.

    1981-01-01

    We describe the operation of the neutral beam injection system on the PLT tokamak. Improvements, retrofits, and conditioning have changed the injection system from an experiment in itself to a fairly reliable and useful plasma heating tool. We will present a brief overview of our physics achievements and then describe the system as it exists now. This will include injector performance, conditioning needs, maintenance needs, reliability, and daily operating sequences. We will also include hardware modifications and additions, electrical and mechanical, and point out remaining problem areas

  14. The calculation of Tritium burnup in Tokamaks

    International Nuclear Information System (INIS)

    Bittoni, E.; Haegi, M.

    1987-01-01

    In a deuterium plasma tokamak, the contained fusion-produced tritons are supposed to be decelerated down to thermalization according to classical Coulomb scattering. A fraction of these fast tritons undergoes the DT fusion reaction producing 14.1 MeV neutrons. It is thus possible to get information on the confinement of these fast tritons by comparing the measured and the calculated ratio of the 14.1 MeV to the 2.45 MeV neutron flux. This report describes the calculation of this flux ratio by means of a numerical Monte Carlo-like code

  15. The steady-state tokamak program

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  16. 'Crescent'-shaped tokamak for compact ignition

    International Nuclear Information System (INIS)

    Yamazaki, K.; Reiersen, W.T.

    1985-12-01

    A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross-section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design dose not rely on the second stability and is easily extendable from the present dee-shaped design. (author)

  17. 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

  18. 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

  19. 'Crescent'-shaped tokamak for compact ignition

    International Nuclear Information System (INIS)

    Yamazaki, K.; Reiersen, W.T.

    1986-01-01

    A compact high-beta tokamak configuration with ''crescent''-shaped (or ''boomerang''-shaped) cross section is proposed as a next-generation ignition machine. This configuration with a small indentation but a large triangularity is more compact than the normal dee-shaped design because of its high-beta characteristics in the first-second transition regime of stability. This may also be a more reliable next-generation compact device than the bean-shaped design with large indentation and small triangularity, because this design does not rely on the second stability and is easily extendable from the present dee-shaped design. (author)

  20. Filamentary probe on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Kovařík, Karel; Ďuran, Ivan; Stöckel, Jan; Seidl, Jakub; Adámek, Jiří; Spolaore, M.; Vianello, N.; Háček, Pavel; Hron, Martin; Pánek, Radomír

    2017-01-01

    Roč. 88, č. 3 (2017), č. článku 035106. ISSN 0034-6748 R&D Projects: GA MŠk(CZ) 8D15001; GA ČR(CZ) GA15-10723S; GA ČR(CZ) GA16-25074S Institutional support: RVO:61389021 Keywords : tokamak * filaments * scrape-off layer Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.515, year: 2016 http://aip.scitation.org/doi/10.1063/1.4977591