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Sample records for tokamak fusion plasmas

  1. Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Scharer, J.E.

    1992-01-01

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.

  2. Nuclear fusion as energetic source: plasma of deuterium and tritium in TFTR Tokamak. La fusion nuclear como alternativa energetica: plasmas de deuterio y tritio en el Tokamak TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Tagle, J.A.; Loarte, A.

    1994-01-01

    In the last two years some scientifical and technological developments in fusion energy have contributed to consider this energy as an alternative source of electric power energy. The Physics plasma laboratory of Princeton University worked with plasma of 50% deuterium and 50% tritium and produced 3 Mw of fusion power. The Tokamak Fusion Test Reactor (TFTR) opens a new way to find new energy sources.

  3. Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically in influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the part of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material inter actions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (author)

  4. Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next-step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several cm from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the engineering design activities of the international thermonuclear experimental reactor project (ITER) and significant progress has been made in better understanding these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/re-deposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D avenues for their resolution are presented. (orig.)

  5. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  6. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    International Nuclear Information System (INIS)

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.

    2001-01-01

    The major increase in discharge duration and plasma energy in a next-step DT (deuterium-tritium) fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D (Research and Development) avenues for their resolution are presented

  7. Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Bell, M.G.; Beer, M.

    1997-02-01

    Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l i ). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q a ∼ 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l i plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q 0 > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions

  8. 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

  9. Using plasma waves to create in tokamaks the necessary quasi-stationary conditions for controlled fusion

    International Nuclear Information System (INIS)

    Moreau, D.

    1993-04-01

    It is studied, on the one hand, how using hybrid waves with frequency near from lower hybrid frequency in fusion plasma. Works about coupling waves in plasma (chap.I), their propagation and response of the plasma to the absorption of the waves (chap.II). This method is the most effective until today. Because of limits, it has been investigated, on the other hand, fast magnetosonic wave to control current density in the centre of the discharge in a reactor or a very hot plasma. Theoretical study (chap.III) and experimental results (chap.IV) are presented. Experiments are in progress or planned in following tokamaks: D3-D (USA), JET (Europe), TORE SUPRA (France), JT-60 (Japan). figs. refs. tabs

  10. 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.

  11. 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

  12. Initial confinement studies of ohmically heated plasmas in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Efthimion, P.C.; Bell, M.; Blanchard, W.R.

    1984-06-01

    Initial operation of the Tokamak Fusion Test Reactor (TFTR) has concentrated upon confinement studies of ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (tau/sub E/) are 0.1 to 0.2 s for a line-average density range (anti n/sub e/) of 1 to 2.5 x 10 19 m -3 with electron temperatures of T/sub e/(o) approx. 1.2 to 2.2 keV, ion temperatures of T/sub i/(o) approx. 0.9 to 1.5 keV, and Z/sub eff/ approx. 3. A comparison of PLT, PDX, and TFTR plasma confinement supports a dimension-cubed scaling law

  13. Predictive modelling of edge transport phenomena in ELMy H-mode tokamak fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Loennroth, J.-S.

    2009-07-01

    This thesis discusses a range of work dealing with edge plasma transport in magnetically confined fusion plasmas by means of predictive transport modelling, a technique in which qualitative predictions and explanations are sought by running transport codes equipped with models for plasma transport and other relevant phenomena. The focus is on high confinement mode (H-mode) tokamak plasmas, which feature improved performance thanks to the formation of an edge transport barrier. H-mode plasmas are generally characterized by the occurrence of edge localized modes (ELMs), periodic eruptions of particles and energy, which limit confinement and may turn out to be seriously damaging in future tokamaks. The thesis introduces schemes and models for qualitative study of the ELM phenomenon in predictive transport modelling. It aims to shed new light on the dynamics of ELMs using these models. It tries to explain various experimental observations related to the performance and ELM-behaviour of H-mode plasmas. Finally, it also tries to establish more generally the potential effects of ripple-induced thermal ion losses on H-mode plasma performance and ELMs. It is demonstrated that the proposed ELM modelling schemes can qualitatively reproduce the experimental dynamics of a number of ELM regimes. Using a theory-motivated ELM model based on a linear instability model, the dynamics of combined ballooning-peeling mode ELMs is studied. It is shown that the ELMs are most often triggered by a ballooning mode instability, which renders the plasma peeling mode unstable, causing the ELM to continue in a peeling mode phase. Understanding the dynamics of ELMs will be a key issue when it comes to controlling and mitigating the ELMs in future large tokamaks. By means of integrated modelling, it is shown that an experimentally observed increase in the ELM frequency and deterioration of plasma confinement triggered by external neutral gas puffing might be due to a transition from the second to

  14. Predictive modelling of edge transport phenomena in ELMy H-mode tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Loennroth, J.-S.

    2009-01-01

    This thesis discusses a range of work dealing with edge plasma transport in magnetically confined fusion plasmas by means of predictive transport modelling, a technique in which qualitative predictions and explanations are sought by running transport codes equipped with models for plasma transport and other relevant phenomena. The focus is on high confinement mode (H-mode) tokamak plasmas, which feature improved performance thanks to the formation of an edge transport barrier. H-mode plasmas are generally characterized by the occurrence of edge localized modes (ELMs), periodic eruptions of particles and energy, which limit confinement and may turn out to be seriously damaging in future tokamaks. The thesis introduces schemes and models for qualitative study of the ELM phenomenon in predictive transport modelling. It aims to shed new light on the dynamics of ELMs using these models. It tries to explain various experimental observations related to the performance and ELM-behaviour of H-mode plasmas. Finally, it also tries to establish more generally the potential effects of ripple-induced thermal ion losses on H-mode plasma performance and ELMs. It is demonstrated that the proposed ELM modelling schemes can qualitatively reproduce the experimental dynamics of a number of ELM regimes. Using a theory-motivated ELM model based on a linear instability model, the dynamics of combined ballooning-peeling mode ELMs is studied. It is shown that the ELMs are most often triggered by a ballooning mode instability, which renders the plasma peeling mode unstable, causing the ELM to continue in a peeling mode phase. Understanding the dynamics of ELMs will be a key issue when it comes to controlling and mitigating the ELMs in future large tokamaks. By means of integrated modelling, it is shown that an experimentally observed increase in the ELM frequency and deterioration of plasma confinement triggered by external neutral gas puffing might be due to a transition from the second to

  15. Forthcoming Break-Even Conditions of Tokamak Plasma Performance for Fusion Energy Development

    Science.gov (United States)

    Hiwatari, Ryoji; Okano, Kunihiko; Asaoka, Yoshiyuki; Tokimatsu, Koji; Konishi, Satoshi; Ogawa, Yuichi

    The present study reveals forthcoming break-even conditions of tokamak plasma performance for the fusion energy development. The first condition is the electric break-even condition, which means that the gross electric power generation is equal to the circulating power in a power plant. This is required for fusion energy to be recognized as a suitable candidate for an alternative energy source. As for the plasma performance (normalized beta value ΒN), confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density fnGW), the electric break-even condition requires the simultaneous achievement of 1.2 market. By using a long-term world energy scenario, a break-even price for introduction of fusion energy in the year 2050 is estimated to lie between 65 mill/kWh and 135 mill/kWh under the constraint of 550 ppm CO2 concentration in the atmosphere. In the present study, this break-even price is applied to the economic break-even condition. However, because this break-even price is based on the present energy scenario including uncertainties, the economic break-even condition discussed here should not be considered the sufficient condition, but a necessary condition. Under the conditions of Btmax = 16 T, ηe = 40 %, plant availability 60 %, and a radial build with/without CS coil, the economic break-even condition requires ΒN ˜ 5.0 for 65 mill/kWh of lower break-even price case. Finally, the present study reveals that the demonstration of steady-state operation with ΒN ˜ 3.0 in the ITER project leads to the upper region of the break-even price in the present world energy scenario, which implies that it is necessary to improve the plasma performance beyond that of the ITER advanced plasma operation.

  16. Forthcoming break-even conditions of tokamak plasma performance for fusion energy development

    International Nuclear Information System (INIS)

    Hiwatari, Ryoji; Okano, Kunihiko; Asaoka, Yoshiyuki; Tokimatsu, Koji; Konishi, Satoshi; Ogawa, Yuichi

    2005-01-01

    The present study reveals forthcoming break-even conditions of tokamak plasma performance for the fusion energy development. The first condition is the electric break-even condition, which means that the gross electric power generation is equal to the circulating power in a power plant. This is required for fusion energy to be recognized as a suitable candidate for an alternative energy source. As for the plasma performance (normalized beta value β N , confinement improvement factor for H-mode HH, the ratio of plasma density to Greenwald density fn GW ), the electric break-even condition requires the simultaneous achievement of 1.2 N GW tmax =16 T, thermal efficiency η e =30%, and current drive power P NBI N ∼1.8, HH≠1.0, and fn GW ∼0.9, which correspond to the ITER reference operation parameters, have a strong potential to achieve the electric break-even condition. The second condition is the economic break-even condition, which is required for fusion energy to be selected as an alternative energy source in the energy market. By using a long-term world energy scenario, a break-even price for introduction of fusion energy in the year 2050 is estimated to lie between 65 mill/kWh and 135 mill/kWh under the constraint of 550 ppm CO 2 concentration in the atmosphere. In the present study, this break-even price is applied to the economic break-even condition. However, because this break-even price is based on the present energy scenario including uncertainties, the economic break-even condition discussed here should not be considered the sufficient condition, but a necessary condition. Under the conditions of B tmax =16 T, η e =40%, plant availability 60%, and a radial build with/without CS coil, the economic break-even condition requires β N ∼5.0 for 65 mill/kWh of lower break-even price case. Finally, the present study reveals that the demonstration of steady-state operation with β N ∼3.0 in the ITER project leads to the upper region of the break

  17. Liquid Metals as Plasma-facing Materials for Fusion Energy Systems: From Atoms to Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Stone, Howard A. [Princeton Univ., NJ (United States); Koel, Bruce E. [Princeton Univ., NJ (United States); Bernasek, Steven L. [Princeton Univ., NJ (United States); Carter, Emily A. [Princeton Univ., NJ (United States); Debenedetti, Pablo G. [Princeton Univ., NJ (United States); Panagiotopoulos, Athanassios Z. [Princeton Univ., NJ (United States)

    2017-06-23

    The objective of our studies was to advance our fundamental understanding of liquid metals as plasma-facing materials for fusion energy systems, with a broad scope: from atoms to tokamaks. The flow of liquid metals offers solutions to significant problems of the plasma-facing materials for fusion energy systems. Candidate metals include lithium, tin, gallium, and their eutectic combinations. However, such liquid metal solutions can only be designed efficiently if a range of scientific and engineering issues are resolved that require advances in fundamental fluid dynamics, materials science and surface science. In our research we investigated a range of significant and timely problems relevant to current and proposed engineering designs for fusion reactors, including high-heat flux configurations that are being considered by leading fusion energy groups world-wide. Using experimental and theoretical tools spanning atomistic to continuum descriptions of liquid metals, and bridging surface chemistry, wetting/dewetting and flow, our research has advanced the science and engineering of fusion energy materials and systems. Specifically, we developed a combined experimental and theoretical program to investigate flows of liquid metals in fusion-relevant geometries, including equilibrium and stability of thin-film flows, e.g. wetting and dewetting, effects of electromagnetic and thermocapillary fields on liquid metal thin-film flows, and how chemical interactions and the properties of the surface are influenced by impurities and in turn affect the surface wetting characteristics, the surface tension, and its gradients. Because high-heat flux configurations produce evaporation and sputtering, which forces rearrangement of the liquid, and any dewetting exposes the substrate to damage from the plasma, our studies addressed such evaporatively driven liquid flows and measured and simulated properties of the different bulk phases and material interfaces. The range of our studies

  18. Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks. Annual report, November 16, 1991--November 15, 1992

    Energy Technology Data Exchange (ETDEWEB)

    Scharer, J.E.

    1992-12-31

    The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.

  19. Reaction-rate coefficients, high-energy ions slowing-down, and power balance in a tokamak fusion reactor plasma

    International Nuclear Information System (INIS)

    Tone, Tatsuzo

    1978-07-01

    Described are the reactivity coefficient of D-T fusion reaction, slowing-down processes of deuterons injected with high energy and 3.52 MeV alpha particles generated in D-T reaction, and the power balance in a Tokamak reactor plasma. Most of the results were obtained in the first preliminary design of JAERI Experimental Fusion Reactor (JXFR) driven with stationary neutral beam injection. A manual of numerical computation program ''BALTOK'' developed for the calculations is given in the appendix. (auth.)

  20. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  1. 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.)

  2. 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.)

  3. Fokker--Planck/transport analyses of fusion plasmas in contemporary beam-driven tokamaks

    International Nuclear Information System (INIS)

    Mirin, A.A.; McCoy, M.G.; Killeen, J.; Rensink, M.E.; Shumaker, D.E.; Jassby, D.L.; Post, D.E.

    1978-04-01

    The properties of deuterium plasmas in experimental tokamaks heated and fueled by intense neutral-beam injection are evaluated with a Fokker-Planck/radial transport code coupled with a Monte Carlo neutrals treatment. Illustrative results are presented for the Poloidal Divertor Experiment at PPPL as a function of beam power and plasma recycling coefficient, R/sub c/. When P/sub beam/ = 8 MW at E/sub b/ = 60 keV, and R/sub c/ = 0.2, then approximately 0.5, [ 2 / 3 ] = 22 keV approximately 6 , and the D-D neutron intensity is 10 16 n/sec

  4. Study of fusion plasma microturbulence by reflectometry in Tore Supra tokamak

    International Nuclear Information System (INIS)

    Gerbaud, Th.

    2008-11-01

    Fast-sweeping reflectometry in extraordinary mode allows direct measurement of radial wave-number local spectra S(δ n /n](k r ,r), and radial profiles of density fluctuations, on Tore Supra tokamak. Wavelet-based approach - a mathematical tool for position-frequency analysis - made possible to consider the strong radial variation of the measured turbulence. Special consideration was given to the validation of spectra and turbulent profiles measurements, by comparing with experimental measurements (reflectometry, probes) and numerical non-linear gyrokinetic simulations. This density fluctuations measurement method has been used to analyse the local transport, by performing a dimensionless scaling on collisionality, ν * . The scaling experiments allow direct comparisons of plasmas from different tokamaks. A clear decrease of the normalized confinement time of the plasma energy with the normalized collisionality was observed: Bτ E ∼ ν *-0.5±0.15 . These new measurements of density fluctuations profiles have shown an intense rise of the edge turbulence (r/a > 0.8) when increasing - also observed by Doppler reflectometry diagnostic - providing a physical explanation of the loss of confinement with the normalized collisionality. More central regions did not present apparent variations (δ n /n, χ(eff)). Core plasma simulations (linear stability code KineZero and non-linear gyrokinetic GYRO) were performed, in order to analyse the experimental behaviour of the plasma. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Intrator, T.

    2000-06-01

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

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

    International Nuclear Information System (INIS)

    Intrator, T.

    2000-01-01

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

  7. ORNL TNS Program: plasma engineering considerations and innovations for a medium field tokamak fusion reactor

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Attenberger, S.E.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

    1977-12-01

    Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma anti β values up to 10%, and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors; some of the suggested innovations will be tested in upcoming experiments

  8. 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).

  9. A study on the fusion reactor - Study of ICRF coupling in the KAIST tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Duk In; Chang, Hong Young; Lee, Sun Chil; Jun, Sang Jin; Kwon, Gi Chung; Seo, Sung Hun; Heo, Sung Hoi; You, Kwang Il; Song, Soo Bin; Lee, Sung Chul; Kim, Min Chul; Lee, Chan Hui [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-09-01

    Research objectives are to design and fabricate the antenna, measure t property of absorption transmitted to the plasma, and research the physical phenomena about the ICRF coupling, Main heating method is ohmic heating at the KAIST tokamak. The power of the plasma produced by ohmic heating is about 100 kW. Because the toroidal field is 5 {approx} 8 kG, the RF system`s output power is about 10 kW and frequency range is 7 {approx} 30 MHz. In the first year, a 1 kW RF preamplifier was bought. In this year, a CW 2 kW RF main amp. and RF power monitoring system was bought. In the research on antenna, we study the method how to measure electric field emitted from antenna using piezo elements. The matching network composed of two VVC (35 kV), 100 {approx} 1000 pF match firmly up to 50 kW power. We studied the measurement method of antenna impedance theoretically, and measured power efficiency and antenna impedance in the helicon plasma. 32 refs., 5 tabs., 29 figs. (author)

  10. Passive MHD Spectroscopy for Enabling Magnetic Reconstructions on Spherical Tokamak Plasmas at General Fusion Inc

    Science.gov (United States)

    O'Shea, Peter; Laberge, Michel; Mossman, Alex; Reynolds, Meritt

    2017-10-01

    Magnetic reconstructions on lab based plasma injectors at General Fusion relies heavily on edge magnetic (``Bdot'') probes. On plasma experiments built for field compression (PCS) tests, the number and locations of Bdot probes is limited by mechanical constraints. Additional information about the q profiles near the core in our Spector plasmas is obtained using passive MHD spectroscopy. The coaxial helicity injection (CHI) formation process naturally generates hollow current profiles and reversed shear early in each discharge. Central Ohmic heating naturally peaks the current profiles as our plasmas evolve in time, simultaneously reducing the core safety factor, q(0), and reverse shear. As the central, non-monotonic q-profile crosses rational flux surfaces, we observe transient magnetic reconnection events (MRE's) due to the double tearing mode. Modal analysis allows us to infer the q surfaces involved in each burst. The parametric dependence of the timing of MRE's allows us to estimate the continuous time evolution of the core q profile. Combining core MHD spectroscopy with edge magnetic probe measurements greatly enhances our certainty of the overall q profile.

  11. Bibliography of fusion product physics in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Hively, L. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sigmar, D. J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    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.

  12. The spherical tokamak fusion power plant

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  13. 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

  14. 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)

  15. Controlled fusion and plasma physics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).

  16. Controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    1995-01-01

    This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC)

  17. Fusion Energy-Production from a Deuterium-Tritium Plasma in the Jet Tokamak

    NARCIS (Netherlands)

    Rebut, P. H.; Gibson, A.; Huguet, M.; Adams, J. M.; Alper, B.; Altmann, H.; Andersen, A.; Andrew, P.; Angelone, M.; Aliarshad, S.; Baigger, P.; Bailey, W.; Balet, B.; Barabaschi, P.; Barker, P.; Barnsley, R.; Baronian, M.; Bartlett, D. V.; Baylor, L.; Bell, A. C.; Benali, G.; Bertoldi, P.; Bertolini, E.; Bhatnagar, V.; Bickley, A. J.; Binder, D.; Bindslev, H.; Bonicelli, T.; Booth, S. J.; Bosia, G.; Botman, M.; Boucher, D.; Boucquey, P.; Breger, P.; Brelen, H.; Brinkschulte, H.; Brooks, D.; Brown, A.; Brown, T.; Brusati, M.; Bryan, S.; Brzozowski, J.; Buchse, R.; Budd, T.; Bures, M.; Businaro, T.; Butcher, P.; Buttgereit, H.; Caldwellnichols, C.; Campbell, D. J.; Card, P.; Celentano, G.; Challis, C. D.; Chankin, A. V.; Cherubini, A.; Chiron, D.; Christiansen, J.; Chuilon, P.; Claesen, R.; Clement, S.; Clipsham, E.; Coad, J. P.; Coffey, I. H.; Colton, A.; Comiskey, M.; Conroy, S.; Cooke, M.; Cooper, D.; Cooper, S.; Cordey, J. G.; Core, W.; Corrigan, G.; Corti, S.; Costley, A. E.; Cottrell, G.; Cox, M.; Cripwell, P.; Dacosta, O.; Davies, J.; Davies, N.; de Blank, H.; De Esch, H.; Dekock, L.; Deksnis, E.; Delvart, F.; Dennehinnov, G. B.; Deschamps, G.; Dickson, W. J.; Dietz, K. J.; Dmitrenko, S. L.; Dmitrieva, M.; Dobbing, J.; Doglio, A.; Dolgetta, N.; Dorling, S. E.; Doyle, P. G.; Duchs, D. F.; Duquenoy, H.; Edwards, A.; Ehrenberg, J.; Ekedahl, A.; Elevant, T.; Erents, S.K.; Eriksson, L. G.; Fajemirokun, H.; Falter, H.; Freiling, J.; Freville, F.; Froger, C.; Froissard, P.; Fullard, K.; Gadeberg, M.; Galetsas, A.; Gallagher, T.; Gambier, D.; Garribba, M.; Gaze, P.; Giannella, R.; Gill, R. D.; Girard, A.; Gondhalekar, A.; Goodall, D.; Gormezano, C.; Gottardi, N. A.; Gowers, C.; Green, B. J.; Grievson, B.; Haange, R.; Haigh, A.; Hancock, C. J.; Harbour, P. J.; Hartrampf, T.; Hawkes, N. C.; Haynes, P.; Hemmerich, J. L.; Hender, T.; Hoekzema, J.; Holland, D.; Hone, M.; Horton, L.; How, J.; Huart, M.; Hughes, I.; Hughes, T. P.; Hugon, M.; Huo, Y.; Ida, K.; Ingram, B.; Irving, M.; Jacquinot, J.; Jaeckel, H.; Jaeger, J. F.; Janeschitz, G.; Jankovicz, Z.; Jarvis, O. N.; Jensen, F.; Jones, E. M.; Jones, H. D.; Jones, Lpdf; Jones, S.; Jones, T. T. C.; Junger, J. F.; Junique, F.; Kaye, A.; Keen, B. E.; Keilhacker, M.; Kelly, G. J.; Kerner, W.; Khudoleev, A.; Konig, R.; Konstantellos, A.; Kovanen, M.; Kramer, G.; Kupschus, P.; Lasser, R.; Last, J. R.; Laundy, B.; Laurotaroni, L.; Laveyry, M.; Lawson, K.; Lennholm, M.; Lingertat, J.; Litunovski, R. N.; Loarte, A.; Lobel, R.; Lomas, P.; Loughlin, M.; Lowry, C.; Lupo, J.; Maas, A. C.; Machuzak, J.; Macklin, B.; Maddison, G.; Maggi, C. F.; Magyar, G.; Mandl, W.; Marchese, V.; Marcon, G.; Marcus, F.; Mart, J.; Martin, D.; Martin, E.; Martinsolis, R.; Massmann, P.; Matthews, G.; McBryan, H.; McCracken, G.; McKivitt, J.; Meriguet, P.; Miele, P.; Miller, A.; Mills, J.; Mills, S. F.; Millward, P.; Milverton, P.; Minardi, E.; Mohanti, R.; Mondino, P. L.; Montgomery, D.; Montvai, A.; Morgan, P.; Morsi, H.; Muir, D.; Murphy, G.; Myrnas, R.; Nave, F.; Newbert, G.; Newman, M.; Nielsen, P.; Noll, P.; Obert, W.; Obrien, D.; Orchard, J.; Orourke, J.; Ostrom, R.; Ottaviani, M.; Pain, M.; Paoletti, F.; Papastergiou, S.; Parsons, W.; Pasini, D.; Patel, D.; Peacock, A.; Peacock, N.; Pearce, R. J. M.; Pearson, D.; Peng, J. F.; Desilva, R. P.; Perinic, G.; Perry, C.; Petrov, M.; Pick, M. A.; Plancoulaine, J.; Poffe, J. P.; Pohlchen, R.; Porcelli, F.; Porte, L.; Prentice, R.; Puppin, S.; Putvinskii, S.; Radford, G.; Raimondi, T.; Deandrade, M. C. R.; Reichle, R.; Reid, J.; Richards, S.; Righi, E.; Rimini, F.; Robinson, D.; Rolfe, A.; Ross, R. T.; Rossi, L.; Russ, R.; Rutter, P.; Sack, H. C.; Sadler, G.; Saibene, G.; Salanave, J. L.; Sanazzaro, G.; Santagiustina, A.; Sartori, R.; Sborchia, C.; Schild, P.; Schmid, M.; Schmidt, G.; Schunke, B.; Scott, S. M.; Serio, L.; Sibley, A.; Simonini, R.; Sips, A.C.C.; Smeulders, P.; Smith, R.; Stagg, R.; Stamp, M.; Stangeby, P.; Stankiewicz, R.; Start, D. F.; Steed, C. A.; Stork, D.; Stott, P.E.; Stubberfield, P.; Summers, D.; Summers, H.; Svensson, L.; Tagle, J. A.; Talbot, M.; Tanga, A.; Taroni, A.; Terella, C.; Terrington, A.; Tesini, A.; Thomas, P. R.; Thompson, E.; Thomsen, K.; Tibone, F.; Tiscornia, A.; Trevalion, P.; Tubbing, B.; Vanbelle, P.; Vanderbeken, H.; Vlases, G.; von Hellermann, M.; Wade, T.; Walker, C.; Walton, R.; Ward, D.; Watkins, M. L.; Watkins, N.; Watson, M. J.; Weber, S.; Wesson, J.; Wijnands, T. J.; Wilks, J.; Wilson, D.; Winkel, T.; Wolf, R.; Wong, D.; Woodward, C.; Wu, Y.; Wykes, M.; Young, D.; Young, I. D.; Zannelli, L.; Zolfaghari, A.; Zwingmann, W.

    1992-01-01

    The paper describes a series of experiments in the Joint European Torus (JET), culminating in the first tokamak discharges in deuterium-tritium fuelled mixtures. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. The objectives were: (i) to

  18. 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)

  19. Fusion Plasma Theory: Task 1, Magnetic confinement Fusion Plasma Theory

    International Nuclear Information System (INIS)

    Callen, J.D.

    1993-01-01

    The research performed under this grant during the current year has concentrated on few tokamak plasma confinement issues: applications of our new Chapman-Enskog-like approach for developing hybrid fluid/kinetic descriptions of tokamak plasmas; multi-faceted studies as part of our development of a new interacting island paradigm for the tokamak equilibrium'' and transport; investigations of the resolution power of BES and ECE diagnostics for measuring core plasma fluctuations; and studies of net transport in the presence of fluctuating surfaces. Recent progress and publications in these areas, and in the management of the NERSC node and the fusion theory workstations are summarized briefly in this report

  20. 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

  1. Tokamak fusion reactor

    International Nuclear Information System (INIS)

    Nohara, Kiyohiko

    2009-01-01

    The structural material is one of key issues for the development of reliable superconducting magnets and peripheral equipments of fusion reactors. Standard stainless steels like SUS 304 and 316 steels available at present do not meet requirements. We are developing a new austenitic steel that has proposed target properties named 'JAERI BOX'. Additions of N and V at different amounts were tested to improve strength and fracture toughness of a base alloy SUS316LN at 4.2 K. Mechanical properties of the developed steel were examined. It is found that the charpy absorbed energy and the fracture toughness of the developed steel at 4.2 K are within JAERI BOX. (T.I.)

  2. Stability analysis of tokamak plasmas

    International Nuclear Information System (INIS)

    Bourdelle, C.

    2000-10-01

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  3. Particle and energy balances in tokamak plasmas

    International Nuclear Information System (INIS)

    Tazima, Teruhiko

    1978-06-01

    Computational and experimental studies on particle and energy balances in tokamak plasmas are described. Firstly, concerning the modeling of tokamak plasmas, the particle balance considering diffusion and recycling, and the energy balance considering transport and energy losses due to impurities are discussed. Production mechanisms of gaseous and metallic impurities, which play important role in tokamak plasmas, are also discussed from a viewpoint of plasma-wall interactions. Scaling laws of density, temperature and energy confinement time are shown on the basis of recent data. Secondarily, tokamak plasmas are simulated with the above model, and anomalous diffusion and electron thermal conduction are indicated. Characteristics of a future tokamak plasma are also simulated. Stationary impurity density distributions and related energy losses, such as bremsstrahlung, ionization and excitation, are calculated taking into account diffusion and ionization processes. Edge cooling by oxygen impurities is described quantitatively compared with experiments. Permissible impurity levels of carbon, oxygen and iron in future large tokamaks are estimated. Thirdly, experimental studies on surface cleaning methods of the first wall are described; discharge cleaning in JFT-2, baking effect on the outgassing rates of wall materials, surface treatment of high-temperature molybdenum by oxygen and hydrogen gases, and in-situ coating of molybdenum by a coaxial magnetron sputter method. Lastly, problems in future large tokamaks aiming at break-even or self-ignited plasma are discussed quantitatively, such as trapped particle instabilities, impurities and additional heating. It is predicted that new conceptions will be necessary to overcome the problems and attain the fusion goal. (auth.)

  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. Use of plasma waves to create in Tokamaks quasi-stationary conditions required for controlled fusion

    International Nuclear Information System (INIS)

    Moreau, D.

    1993-04-01

    In this thesis are studied the coupling of hybrid waves to the plasma, multijunction antennas, hybrid wave stochastic propagation, fast wave current drive and lower-hybrid current drive experiments in Tore Supra and Jet. The possibility of decoupling current density profile and temperature give one more degree of freedom for the control of plasma in a configuration which is not very flexible

  6. Self-consistent kinetic simulations of lower hybrid drift instability resulting in electron current driven by fusion products in tokamak plasmas

    International Nuclear Information System (INIS)

    Cook, J W S; Chapman, S C; Dendy, R O; Brady, C S

    2011-01-01

    We present particle-in-cell (PIC) simulations of minority energetic protons in deuterium plasmas, which demonstrate a collective instability responsible for emission near the lower hybrid frequency and its harmonics. The simulations capture the lower hybrid drift instability in a parameter regime motivated by tokamak fusion plasma conditions, and show further that the excited electromagnetic fields collectively and collisionlessly couple free energy from the protons to directed electron motion. This results in an asymmetric tail antiparallel to the magnetic field. We focus on obliquely propagating modes excited by energetic ions, whose ring-beam distribution is motivated by population inversions related to ion cyclotron emission, in a background plasma with a temperature similar to that of the core of a large tokamak plasma. A fully self-consistent electromagnetic relativistic PIC code representing all vector field quantities and particle velocities in three dimensions as functions of a single spatial dimension is used to model this situation, by evolving the initial antiparallel travelling ring-beam distribution of 3 MeV protons in a background 10 keV Maxwellian deuterium plasma with realistic ion-electron mass ratio. These simulations provide a proof-of-principle for a key plasma physics process that may be exploited in future alpha channelling scenarios for magnetically confined burning plasmas.

  7. Plasma physics for controlled fusion

    CERN Document Server

    Miyamoto, Kenro

    2016-01-01

    This new edition presents the essential theoretical and analytical methods needed to understand the recent fusion research of tokamak and alternate approaches. The author describes magnetohydrodynamic and kinetic theories of cold and hot plasmas in detail. The book covers new important topics for fusion studies such as plasma transport by drift turbulence, which depend on the magnetic configuration and zonal flows. These are universal phenomena of microturbulence. They can modify the onset criterion for turbulent transport, instabilities driven by energetic particles as well as alpha particle generation and typical plasma models for computer simulation. The fusion research of tokamaks with various new versions of H modes are explained. The design concept of ITER, the international tokamak experimental reactor, is described for inductively driven operations as well as steady-state operations using non-inductive drives. Alternative approaches of reversed-field pinch and its relaxation process, stellator includi...

  8. Controlled fusion and plasma physics

    CERN Document Server

    Miyamoto, Kenro

    2006-01-01

    Resulting from ongoing, international research into fusion processes, the International Tokamak Experimental Reactor (ITER) is a major step in the quest for a new energy source.The first graduate-level text to cover the details of ITER, Controlled Fusion and Plasma Physics introduces various aspects and issues of recent fusion research activities through the shortest access path. The distinguished author breaks down the topic by first dealing with fusion and then concentrating on the more complex subject of plasma physics. The book begins with the basics of controlled fusion research, foll

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-07-01

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

  10. Fusion-product transport in axisymmetric tokamaks: losses and thermalization

    Energy Technology Data Exchange (ETDEWEB)

    Hively, L.M.

    1980-01-01

    High-energy fusion-product losses from an axisymmetric tokamak plasma are studied. Prompt-escape loss fluxes (i.e. prior to slowing down) are calculated including the non-separable dependence of flux as a function of poloidal angle and local angle-of-incidence at the first wall. Fusion-product (fp) thermalization and heating are calculated assuming classical slowing down. The present analytical model describes fast ion orbits and their distribution function in realistic, high-..beta.., non-circular tokamak equilibria. First-orbit losses, trapping effects, and slowing-down drifts are also treated.

  11. The Tokamak Fusion Core Experiment studies

    International Nuclear Information System (INIS)

    Schmidt, J.A.; Sheffield, G.V.; Bushnell, C.

    1985-01-01

    The basic objective of the next major step in the US fusion programme has been defined as the achievement of ignition and long pulse equilibrium burn of a fusion plasma in the Tokamak Fusion Core Experiment (TFCX) device. Preconceptual design studies have seen completion of four candidate versions to provide the comparative information needed to narrow down the range of TFCX options before proceeding to the conceptual design phase. All four designs share the same objective and conform to common physics, engineering and costing criteria. The four base options considered differed mainly in the toroidal field coil design, two employing superconducting coils and the other two copper coils. In each case (copper and superconducting), one relatively conventional version was carried as well as a version employing more exotic toroidal field coil design assumptions. Sizes range from R=2.6 m for the smaller of the two copper versions to R=4.08 m for the larger superconducting option. In all cases, the plasma current was about 10 MA and the toroidal field about 4 T. (author)

  12. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    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

    2003-10-01

    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.

  13. 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

  14. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

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

  15. Massachusetts Institute of Technology Plasma Fusion Center 1992--1993 report to the President

    International Nuclear Information System (INIS)

    1993-07-01

    This report discusses research being conducted at MIT's plasma fusion center. Some of the areas covered are: plasma diagnostics; rf plasma heating; gyrotron research; treatment of solid waste by arc plasma; divertor experiments; tokamak studies; and plasma and fusion theory

  16. 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

  17. Plasma detachment in divertor tokamaks

    Science.gov (United States)

    Leonard, A. W.

    2018-04-01

    Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma E× B drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.

  18. 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...wavelength polariza- tion field produced by the curvature and field gradient drifts [15]. The growth rate is y = Vs[k/R 2 = [T(eV)/X(cm)J 2 3.3 x 105 sec

  19. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    F. Ed Cecil

    2011-06-30

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

  1. Fusion Plasma Theory project summaries

    International Nuclear Information System (INIS)

    1993-10-01

    This Project Summary book is a published compilation consisting of short descriptions of each project supported by the Fusion Plasma Theory and Computing Group of the Advanced Physics and Technology Division of the Department of Energy, Office of Fusion Energy. The summaries contained in this volume were written by the individual contractors with minimal editing by the Office of Fusion Energy. Previous summaries were published in February of 1982 and December of 1987. The Plasma Theory program is responsible for the development of concepts and models that describe and predict the behavior of a magnetically confined plasma. Emphasis is given to the modelling and understanding of the processes controlling transport of energy and particles in a toroidal plasma and supporting the design of the International Thermonuclear Experimental Reactor (ITER). A tokamak transport initiative was begun in 1989 to improve understanding of how energy and particles are lost from the plasma by mechanisms that transport them across field lines. The Plasma Theory program has actively-participated in this initiative. Recently, increased attention has been given to issues of importance to the proposed Tokamak Physics Experiment (TPX). Particular attention has been paid to containment and thermalization of fast alpha particles produced in a burning fusion plasma as well as control of sawteeth, current drive, impurity control, and design of improved auxiliary heating. In addition, general models of plasma behavior are developed from physics features common to different confinement geometries. This work uses both analytical and numerical techniques. The Fusion Theory program supports research projects at US government laboratories, universities and industrial contractors. Its support of theoretical work at universities contributes to the office of Fusion Energy mission of training scientific manpower for the US Fusion Energy Program

  2. Fusion Plasma Theory project summaries

    Energy Technology Data Exchange (ETDEWEB)

    1993-10-01

    This Project Summary book is a published compilation consisting of short descriptions of each project supported by the Fusion Plasma Theory and Computing Group of the Advanced Physics and Technology Division of the Department of Energy, Office of Fusion Energy. The summaries contained in this volume were written by the individual contractors with minimal editing by the Office of Fusion Energy. Previous summaries were published in February of 1982 and December of 1987. The Plasma Theory program is responsible for the development of concepts and models that describe and predict the behavior of a magnetically confined plasma. Emphasis is given to the modelling and understanding of the processes controlling transport of energy and particles in a toroidal plasma and supporting the design of the International Thermonuclear Experimental Reactor (ITER). A tokamak transport initiative was begun in 1989 to improve understanding of how energy and particles are lost from the plasma by mechanisms that transport them across field lines. The Plasma Theory program has actively-participated in this initiative. Recently, increased attention has been given to issues of importance to the proposed Tokamak Physics Experiment (TPX). Particular attention has been paid to containment and thermalization of fast alpha particles produced in a burning fusion plasma as well as control of sawteeth, current drive, impurity control, and design of improved auxiliary heating. In addition, general models of plasma behavior are developed from physics features common to different confinement geometries. This work uses both analytical and numerical techniques. The Fusion Theory program supports research projects at US government laboratories, universities and industrial contractors. Its support of theoretical work at universities contributes to the office of Fusion Energy mission of training scientific manpower for the US Fusion Energy Program.

  3. Fusion plasma theory project summaries

    Science.gov (United States)

    1993-10-01

    This Project Summary book is a published compilation consisting of short descriptions of each project supported by the Fusion Plasma Theory and Computing Group of the Advanced Physics and Technology Division of the Department of Energy, Office of Fusion Energy. The summaries contained in this volume were written by the individual contractors with minimal editing by the Office of Fusion Energy. Previous summaries were published in February of 1982 and December of 1987. The Plasma Theory program is responsible for the development of concepts and models that describe and predict the behavior of a magnetically confined plasma. Emphasis is given to the modelling and understanding of the processes controlling transport of energy and particles in a toroidal plasma and supporting the design of the International Thermonuclear Experimental Reactor (ITER). A tokamak transport initiative was begun in 1989 to improve understanding of how energy and particles are lost from the plasma by mechanisms that transport them across field lines. The Plasma Theory program has actively participated in this initiative. Recently, increased attention has been given to issues of importance to the proposed Tokamak Physics Experiment (TPX). Particular attention has been paid to containment and thermalization of fast alpha particles produced in a burning fusion plasma as well as control of sawteeth, current drive, impurity control, and design of improved auxiliary heating. In addition, general models of plasma behavior are developed from physics features common to different confinement geometries. This work uses both analytical and numerical techniques. The Fusion Theory program supports research projects at U.S. government laboratories, universities and industrial contractors. Its support of theoretical work at universities contributes to the office of Fusion Energy mission of training scientific manpower for the U.S. Fusion Energy Program.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. Fusion neutron diagnostics on ITER tokamak

    Science.gov (United States)

    Bertalot, L.; Barnsley, R.; Direz, M. F.; Drevon, J. M.; Encheva, A.; Jakhar, S.; Kashchuk, Y.; Patel, K. M.; Arumugam, A. P.; Udintsev, V.; Walker, C.; Walsh, M.

    2012-04-01

    ITER is an experimental nuclear reactor, aiming to demonstrate the feasibility of nuclear fusion realization in order to use it as a new source of energy. ITER is a plasma device (tokamak type) which will be equipped with a set of plasma diagnostic tools to satisfy three key requirements: machine protection, plasma control and physics studies by measuring about 100 different parameters. ITER diagnostic equipment is integrated in several ports at upper, equatorial and divertor levels as well internally in many vacuum vessel locations. The Diagnostic Systems will be procured from ITER Members (Japan, Russia, India, United States, Japan, Korea and European Union) mainly with the supporting structures in the ports. The various diagnostics will be challenged by high nuclear radiation and electromagnetic fields as well by severe environmental conditions (ultra high vacuum, high thermal loads). Several neutron systems with different sensitivities are foreseen to measure ITER expected neutron emission from 1014 up to almost 1021 n/s. The measurement of total neutron emissivity is performed by means of Neutron Flux Monitors (NFM) installed in diagnostic ports and by Divertor Neutron Flux Monitors (DNFM) plus MicroFission Chambers (MFC) located inside the vacuum vessel. The neutron emission profile is measured with radial and vertical neutron cameras. Spectroscopy is accomplished with spectrometers looking particularly at 2.5 and 14 MeV neutron energy. Neutron Activation System (NAS), with irradiation ends inside the vacuum vessel, provide neutron yield data. A calibration strategy of the neutron diagnostics has been developed foreseeing in situ and cross calibration campaigns. An overview of ITER neutron diagnostic systems and of the associated challenging engineering and integration issues will be reported.

  6. Tritium experience in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A. [Princeton Univ., NJ (United States). Princeton Plasma Physics Lab.; Brooks, J.N. [Argonne National Lab., IL (United States); Hogan, J. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.

  7. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  8. Ion cyclotron emission in tokamak plasmas

    International Nuclear Information System (INIS)

    Fraboulet, D.

    1996-01-01

    Detection of α(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, α particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. α particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central α density in a reactor. (author)

  9. Tokamak-FED plasma-engineering assessments

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Lyon, J.F.; Rutherford, P.H.

    1981-01-01

    A wide range of plasma assumptions and scenarios has been examined for the current US tokamak FED concept, which aims to provide a controlled, long pulse (approx. 100 s) burning plasma with an energy amplification of greater than or equal to 5, a fusion power of 180 MW, and a neutron wall load of greater than or equal to 0.4 MW/m 2 . The results of the assessment suggest that the current FED baseline parameters of R = 5.0 m, B/sub t/ = 3.6 T, a = 1.3 m, b = 2.1 m (D-shape), and I/sub p/ = 5.4 MA are appropriate in reaching the above plasma performance, despite uncertainties in several plasma physics areas, such as confinement scaling, achievable beta, impurity control, etc. To enhance the probability of achieving fusion ignition and to provide some margin against a short fall in our physics projections in FED, a limited operating capability at B/sub t/ = 4.6 T and I/sub p/ = 6.5 MA is incorporated. Various other options and remedies have also been assessed aiming to alleviate the impact of the uncertainties on the FED design concept. These approaches appear promising because they can be studied within the current fusion physics program and may lead to drastically more cost-effective FED concepts

  10. Plasma Sprayed Tungsten-based Coatings and their Usage in Edge Plasma Region of Tokamaks

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Dufková, Edita; Piffl, Vojtěch; Peřina, Vratislav

    2006-01-01

    Roč. 51, č. 2 (2006), s. 179-191 ISSN 0001-7043 Grant - others:Evropská unie EFDA Task TW-5-TVM-PSW (EU – Euratom) Institutional research plan: CEZ:AV0Z20430508; CEZ:AV0Z10480505 Keywords : plasma sprayed coatings * fusion * plasma facing components * tungsten * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics

  11. Measurements of ion energies in the tokamak plasma boundary

    Czech Academy of Sciences Publication Activity Database

    Kocan, M.; Gunn, J. P.; Carpentier-Chouchana, S.; Herrmann, A.; Kirk, A.; Komm, Michael; Müller, H.W.; Pascal, J.Y.; Pitts, R.A.; Rohde, V.; Tamain, P.

    2011-01-01

    Roč. 415, č. 1 (2011), S1133-S1138 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI)/19./. San Diego, CA, 24.05.2010-28.05.2010] Institutional support: RVO:61389021 Keywords : plasma * tokamak * scrape-off layer Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.052, year: 2011 http://www.sciencedirect.com/science/article/pii/S0022311510004277#

  12. PLASMA EQUILIBRIUM IN TOKAMAKS

    NARCIS (Netherlands)

    de Blank, H. J.

    2010-01-01

    This lecture treats the magnetohydrodynamic (MHD) equilibrium of axisymmetric plasmas, as given by the Grad-Shafranov equation. In a brief introduction, equilibrium parameters such as the q-profile, the internal inductance, and the poloidal beta are introduced. The properties of these quantities

  13. Plasma equilibrium in tokamaks

    NARCIS (Netherlands)

    de Blank, H. J.

    2008-01-01

    This lecture treats the magnetohydrodynamic (MHD) equilibrium of axisymmetric plasmas, as given by the Grad-Shafranov equation. In a brief introduction, equilibrium parameters such as the q-profile, the internal inductance, and the poloidal beta are introduced. The properties of these quantities

  14. Plasma equilibrium in tokamaks

    NARCIS (Netherlands)

    de Blank, H. J.

    2012-01-01

    This lecture treats the magnetohydrodynamic (MHD) equilibrium of axisymmetric plasmas, as given by the Grad-Shafranov equation. In a brief introduction, equilibrium parameters such as the q-profile, the internal inductance, and the poloidal beta are introduced. The properties of these quantities

  15. 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

  16. Remote operation of the GOLEM tokamak for Fusion Education

    International Nuclear Information System (INIS)

    Grover, O.; Kocman, J.; Odstrcil, M.; Odstrcil, T.; Matusu, M.; Stöckel, J.; Svoboda, V.; Vondrasek, G.; Zara, J.

    2016-01-01

    Highlights: • The remote operation of the tokamak GOLEM for educational purposes. - Abstract: Practically oriented education in the field of thermonuclear fusion is highly requested. However, the high complexity of appropriate experiments makes it difficult to develop and maintain laboratories where students can take part in hands-on experiments in this field of study. One possible solution is to establish centres with specific high temperature plasma experiments where students can visit such a laboratory and perform their experiments in-situ. With the advancements of IT technologies it naturally follows to make a step forward and connect these with necessary plasma physics technologies and thus allow to access even sophisticated experiments remotely. Tokamak GOLEM is a small, modest device with its infrastructure linked to web technologies allowing students to set-up necessary discharge parameters, submit them into a queue and within minutes obtain the results in the form of a discharge homepage.

  17. Remote operation of the GOLEM tokamak for Fusion Education

    Energy Technology Data Exchange (ETDEWEB)

    Grover, O.; Kocman, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Odstrcil, M. [University of Southampton, Southampton SO17 1BJ (United Kingdom); Odstrcil, T. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Matusu, M. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, Prague CZ-182 21 (Czech Republic); Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Vondrasek, G. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Zara, J. [Faculty of Electrical Engineering CTU Prague, CZ-166 27 (Czech Republic)

    2016-11-15

    Highlights: • The remote operation of the tokamak GOLEM for educational purposes. - Abstract: Practically oriented education in the field of thermonuclear fusion is highly requested. However, the high complexity of appropriate experiments makes it difficult to develop and maintain laboratories where students can take part in hands-on experiments in this field of study. One possible solution is to establish centres with specific high temperature plasma experiments where students can visit such a laboratory and perform their experiments in-situ. With the advancements of IT technologies it naturally follows to make a step forward and connect these with necessary plasma physics technologies and thus allow to access even sophisticated experiments remotely. Tokamak GOLEM is a small, modest device with its infrastructure linked to web technologies allowing students to set-up necessary discharge parameters, submit them into a queue and within minutes obtain the results in the form of a discharge homepage.

  18. Tokamak plasma position dynamics and feedback control

    International Nuclear Information System (INIS)

    Burenko, L.; Bailey, J.M.

    1983-01-01

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form

  19. 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

  20. Neoclassical transport of impurtities in tokamak plasmas

    International Nuclear Information System (INIS)

    Hirshman, S.P.; Sigmar, D.J.

    1981-05-01

    Tokamak plasmas are inherently comprised of multiple ion species. This is due to wall-bred impurities and, in future reactors, will result from fusion-born alpha particles. Relatively small concentrations of highly charged non-hydrogenic impurities can strongly influence plasma transport properties whenever n/sub I/e/sub I/ 2 /n/sub H/e 2 greater than or equal to (m/sub e//m/sub H/)/sup 1/2/. The determination of the complete neoclassical Onsager matrix for a toroidally confined multispecies plasma, which provides the linear relation between the surface averaged radial fluxes and the thermodynamic forces (i.e., gradients of density and temperature, and the parallel electric field), is reviewed. A closed set of one-dimensional moment equations is presented for the time evolution of thermodynamic and magnetic field quantities which results from collisional transport of the plasma and two dimensional motion of the magnetic flux surface geometry. The effects of neutral beam injection on the equilibrium and transport properties of a toroidal plasma are consistently included

  1. 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)

  2. 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.

  3. 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

  4. Fusion plasma physics

    CERN Document Server

    Stacey, Weston M

    2012-01-01

    This revised and enlarged second edition of the popular textbook and reference contains comprehensive treatments of both the established foundations of magnetic fusion plasma physics and of the newly developing areas of active research. It concludes with a look ahead to fusion power reactors of the future. The well-established topics of fusion plasma physics -- basic plasma phenomena, Coulomb scattering, drifts of charged particles in magnetic and electric fields, plasma confinement by magnetic fields, kinetic and fluid collective plasma theories, plasma equilibria and flux surface geometry, plasma waves and instabilities, classical and neoclassical transport, plasma-materials interactions, radiation, etc. -- are fully developed from first principles through to the computational models employed in modern plasma physics. The new and emerging topics of fusion plasma physics research -- fluctuation-driven plasma transport and gyrokinetic/gyrofluid computational methodology, the physics of the divertor, neutral ...

  5. Plasma surface interactions in controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Ghendrih, Ph.; Becoulet, M.; Costanzo, L. [and others

    2000-07-01

    This report brings together all the contributions of EURATOM/CEA association to the 14. international conference on plasma surface interactions in controlled fusion devices. 24 papers are presented and they deal mainly with the ergodic divertor and the first wall of Tore-supra tokamak.

  6. 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

  7. Neutral-beam-injected tokamak fusion reactors: a review

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1976-08-01

    The theories of energetic-ion velocity distributions, stability, injection, and orbits were summarized. The many-faceted role of the energetic ions in plasma heating, fueling, and current maintenance, as well as in the direct enhancement of fusion power multiplication and power density, is discussed in detail for three reactor types. The relevant implications of recent experimental results on several beam-injected tokamaks are examined. The behavior of energetic ions is found to be in accordance with classical theory, large total ion energy densities are readily achieved, and plasma equilibrium and stability are maintained. The status of neutral-beam injectors and of conceptual design studies of beam-driven reactors are briefly reviewed. The principal plasma-engineering problems are those associated directly with achieving quasi-stationary operation

  8. Diamond Wire Cutting of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Keith Rule; Erik Perry; Robert Parsells

    2003-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of DandD (Decontamination and Decommissioning) activity

  9. COAST-4, Design and Cost of Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sink, D.A.; Iwinski, E.M.

    1985-01-01

    1 - Description of program or function: COAST produces a generalized description of a D-T burning tokamak reactor and facility. In each complete calculation, the geometry, dimensions, and ratings of approximately 50 subsystems are determined. In addition, performance data associated with the tokamak and facility operation and a detailed cost estimate, subsystem-by-subsystem, are provided. The cost estimates include the evaluation of direct and indirect capital costs, time-related, and operating costs. Devices which can be sized and costed include: TFTR (Tokamak Fusion Test Reactor), The Next Step (TNS) type of devices, fusion-fission hybrids, engineering demonstration reactors, power producing reactors, and commercial (power and fissile fuel breeding) reactors. Both the ignition mode and the neutral-beam-driven mode of plasma operation are modeled. The plasma engineering calculations involve 0-D models which account for energy balance, particle balance, alpha-particle effects, slowing-down theory, plasma-plasma and beam-plasma fusion reaction rates, impurity effects, profile effects, and n(e)-tau(E) scaling. The tokamak engineering calculations account for poloidal and toroidal magnetic field coil assemblies (both superconducting and copper conductors are possible), neutral beam injectors, blanket/shield assemblies, plasma fueling, divertors, heat dissipation systems, and related power supplies. The modeling for the shaping field coils includes the positioning of each coil relative to the other device components, as well as scaling the coil currents from a reference set of current magnitudes and positions (which are supplied as input and should be consistent with a coil set providing equilibrium and stability conditions for the plasma). The reactor cell containing the tokamak device, the turbine-generator plant and facility, and balance-of-plant systems are taken into account. 2 - Method of solution: The COAST code 'builds' the tokamak around the plasma and

  10. Transport and turbulence in a magnetized plasma (application to tokamak plasmas); Transport et turbulence dans un plasma magnetise (application aux plasmas de tokamaks)

    Energy Technology Data Exchange (ETDEWEB)

    Sarazin, Y

    2004-03-01

    This document gathers the lectures made in the framework of a Ph.D level physics class dedicated to plasma physics. This course is made up of 3 parts : 1) collisions and transport, 2) transport and turbulence, and 3) study of a few exchange instabilities. More precisely the first part deals with the following issues: thermonuclear fusion, Coulomb collisions, particles trajectories in a tokamak, neo-classical transport in tokamaks, the bootstrap current, and ware pinch. The second part involves: particle transport in tokamaks, quasi-linear transport, resonance islands, resonance in tokamaks, from quasi to non-linear transport, and non-linear saturation of turbulence. The third part deals with: shift velocities in fluid theory, a model for inter-change instabilities, Rayleigh-Benard instability, Hasegawa-Wakatani model, and Hasegawa-Mima model. This document ends with a series of appendices dealing with: particle-wave interaction, determination of the curvature parameter G, Rossby waves.

  11. 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

  12. 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

  13. Three novel tokamak plasma regimes in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. The importance of the toroidal magnetic field for the feasibility of a tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    Mazzucato, E.

    2000-01-01

    The next step in the demonstration of the scientific feasibility of a tokamak fusion reactor is a DT burning plasma experiment for the study and control of self-heated plasmas. In this paper, the authors examine the role of the toroidal magnetic field on the confinement of a tokamak plasma in the ELMy H-mode regime--the operational regime foreseen for ITER

  15. Controlled Fusion and Plasma Physics

    Energy Technology Data Exchange (ETDEWEB)

    Engelmann, F [Max-Planck-Institut fuer Plasmaphysik (Germany)

    2007-07-15

    This new book by Kenro Miyamoto provides an up-to-date overview of the status of fusion research and the important parts of the underlying plasma physics at a moment where, due to the start of ITER construction, an important step in fusion research has been made and many new research workers will enter the field. For them, and also for interested graduate students and physicists in other fields, the book provides a good introduction into fusion physics as, on the whole, the presentation of the material is quite appropriate for getting acquainted with the field on the basis of just general knowledge in physics. There is overlap with Miyamoto's earlier book Plasma Physics for Nuclear Fusion (MIT Press, Cambridge, USA, 1989) but only in a few sections on subjects which have not evolved since. The presentation is subdivided into two parts of about equal length. The first part, following a concise survey of the physics basis of thermonuclear fusion and of plasmas in general, covers the various magnetic configurations studied for plasma confinement (tokamak; reversed field pinch; stellarator; mirror-type geometries) and introduces the specific properties of plasmas in these devices. Plasma confinement in tokamaks is treated in particular detail, in compliance with the importance of this field in fusion research. This includes a review of the ITER concept and of the rationale for the choice of ITER's parameters. In the second part, selected topics in fusion plasma physics (macroscopic instabilities; propagation of waves; kinetic effects such as energy transfer between waves and particles including microscopic instabilities as well as plasma heating and current drive; transport phenomena induced by turbulence) are presented systematically. While the emphasis is on displaying the essential physics, deeper theoretical analysis is also provided here. Every chapter is complemented by a few related problems, but only partial hints for their solution are given. A

  16. Stability analysis of tokamak plasmas; Analyse de stabilite de plasmas de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bourdelle, C

    2000-10-01

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  17. Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

    International Nuclear Information System (INIS)

    Mansfield, D.K.; Hill, K.W.; Strachan, J.D.; Bell, M.G.; Scott, S.D.; Budny, R.; Marmar, E.S.; Snipes, J.A.; Terry, J.L.; Batha, S.; Bell, R.E.; Bitter, M.; Bush, C.E.; Chang, Z.; Darrow, D.S.; Ernst, D.; Fredrickson, E.; Grek, B.; Herrmann, H.W.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Levinton, F.M.; Mikkelsen, D.R.; Mueller, D.; Owens, D.K.; Park, H.; Ramsey, A.T.; Roquemore, A.L.; Skinner, C.H.; Stevenson, T.; Stratton, B.C.; Synakowski, E.; Taylor, G.; von Halle, A.; von Goeler, S.; Wong, K.L.; Zweben, S.J.

    1996-01-01

    Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium endash tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10 21 m -3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun. copyright 1996 American Institute of Physics

  18. 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.)

  19. 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

  20. OTR-experimental fusion-fission tokamak-reactor concept

    International Nuclear Information System (INIS)

    Kadomtsev, B.B.; Kolbasov, B.N.; Orlov, V.V.

    1987-01-01

    An original concept of OTR-experimental hybrid fusion-fission tokamak-reactor is given. Tendencies to probable changes in the reactor parameters and some engineering alternatives are discussed. The feasibility of ''non-chemical'' fuel cycle with direct neutron enrichment (in the fusion-fission reactor blanket) of the fuel elements made of depleted uranium for their use in fission reactors is considered. Advantages and shortcomings of the blanket cooling with water, liquid metal and with a gas are briefly discussed. The parameters of experimental and commercial fusion-fission reactors and a composition of the nuclear fuel unloaded from such reactors are given. An option for the detail design which should be done by us needs some corrections in plasma geometry and in other parameters with account taken of new experimental and theoretical results, as well as that of the estimates and comparison of different engineering solutions referred to the plasma engineering, electrophysical and nuclear engineering aspects of the design. An initial OTR-concept, the tendencies to the correction of parameters and some engineering alternatives studied in the present stage of design are discussed in the paper. 12 refs, 2 tabs

  1. Quantify Plasma Response to Non-Axisymmetric (3D) Magnetic Fields in Tokamaks, Final Report for FES (Fusion Energy Sciences) FY2014 Joint Research Target

    Energy Technology Data Exchange (ETDEWEB)

    Strait, E. J. [General Atomics, San Diego, CA (United States); Park, J. -K. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Marmar, E. S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Ahn, J. -W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Berkery, J. W. [Columbia Univ., New York, NY (United States); Burrell, K. H. [General Atomics, San Diego, CA (United States); Canik, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Delgado-Aparicio, L. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ferraro, N. M. [General Atomics, San Diego, CA (United States); Garofalo, A. M. [General Atomics, San Diego, CA (United States); Gates, D. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Greenwald, M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kim, K. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); King, J. D. [General Atomics, San Diego, CA (United States); Lanctot, M. J. [General Atomics, San Diego, CA (United States); Lazerson, S. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Liu, Y. Q. [Culham Science Centre, Abingdon (United Kingdom). Euratom/CCFE Association; Logan, N. C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lore, J. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Menard, J. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Nazikian, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Shafer, M. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Paz-Soldan, C. [General Atomics, San Diego, CA (United States); Reiman, A. H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Rice, J. E. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Sabbagh, S. A. [Columbia Univ., New York, NY (United States); Sugiyama, L. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Turnbull, A. D. [General Atomics, San Diego, CA (United States); Volpe, F. [Columbia Univ., New York, NY (United States); Wang, Z. R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Wolfe, S. M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2014-09-30

    The goal of the 2014 Joint Research Target (JRT) has been to conduct experiments and analysis to investigate and quantify the response of tokamak plasmas to non-axisymmetric (3D) magnetic fields. Although tokamaks are conceptually axisymmetric devices, small asymmetries often result from inaccuracies in the manufacture and assembly of the magnet coils, or from nearby magnetized objects. In addition, non-axisymmetric fields may be deliberately applied for various purposes. Even at small amplitudes of order 10-4 of the main axisymmetric field, such “3D” fields can have profound impacts on the plasma performance. The effects are often detrimental (reduction of stabilizing plasma rotation, degradation of energy confinement, localized heat flux to the divertor, or excitation of instabilities) but may in some case be beneficial (maintenance of rotation, or suppression of instabilities). In general, the magnetic response of the plasma alters the 3D field, so that the magnetic field configuration within the plasma is not simply the sum of the external 3D field and the original axisymmetric field. Typically the plasma response consists of a mixture of local screening of the external field by currents induced at resonant surfaces in the plasma, and amplification of the external field by stable kink modes. Thus, validated magnetohydrodynamic (MHD) models of the plasma response to 3D fields are crucial to the interpretation of existing experiments and the prediction of plasma performance in future devices. The non-axisymmetric coil sets available at each facility allow well-controlled studies of the response to external 3D fields. The work performed in support of the 2014 Joint Research Target has included joint modeling and analysis of existing experimental data, and collaboration on new experiments designed to address the goals of the JRT. A major focus of the work was validation of numerical models through quantitative comparison to experimental data, in

  2. 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)

  3. Extremely shaped plasmas to improve the Tokamak concept

    International Nuclear Information System (INIS)

    Piras, F.

    2011-04-01

    Energy is essential for human existence and our future depends on plentiful and accessible sources of energy. The world population is fast growing and the average energy used per capita increases. One of the greatest challenges for human beings is that of meeting the growing demand for energy in a responsible, equitable and sustainable way. The possibility to obtain energy by ‘fusing’ light atoms addresses these needs. Nuclear fusion reactions are clean, safe and the amount of fuel present on Earth (hydrogen isotopes) is practically inexhaustible and well distributed. Nuclear fusion is a natural process that occurs in all active stars like our Sun. Since the first demonstration of a deuterium fusion reaction (Rutherford 1933), researchers worldwide have tried to replicate this process on Earth by building a thermonuclear fusion reactor. Nevertheless, the challenge posed by the construction of a nuclear fusion reactor is greater than the one presented earlier by the development of a fission reactor. During the IAEA Conference in Geneva in the early 1958, L.A. Artsimovich declared: ‘Plasma physics is very difficult. Worldwide collaboration is needed for progress’ and E.Teller, at the same conference: ‘Fusion technology is very complex. It is almost impossible to build a fusion reactor in this century’. They were right. The extremely high temperature and density necessary to fuse hydrogen isotopes makes it difficult indeed to create a successful fusion reactor. Even though the physics of the fusion reaction appears clear, we are still facing problems on the road towards building the ‘box’ that can efficiently confine the hot gas in the state of plasma. The best results so far have been obtained confining a plasma with strong magnetic fields in a toroidal configuration (‘tokamak’). The Centre de Recherches en Physique des Plasmas in Switzerland actively studies this promising configuration towards the development of a nuclear fusion reactor. The

  4. Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Abiteboul, J.

    2012-10-01

    The goal of magnetic confinement devices such as tokamaks is to produce energy from nuclear fusion reactions in plasmas at low densities and high temperatures. Experimentally, toroidal flows have been found to significantly improve the energy confinement, and therefore the performance of the machine. As extrinsic momentum sources will be limited in future fusion devices such as ITER, an understanding of the physics of toroidal momentum transport and the generation of intrinsic toroidal rotation in tokamaks would be an important step in order to predict the rotation profile in experiments. Among the mechanisms expected to contribute to the generation of toroidal rotation is the transport of momentum by electrostatic turbulence, which governs heat transport in tokamaks. Due to the low collisionality of the plasma, kinetic modeling is mandatory for the study of tokamak turbulence. In principle, this implies the modeling of a six-dimensional distribution function representing the density of particles in position and velocity phase-space, which can be reduced to five dimensions when considering only frequencies below the particle cyclotron frequency. This approximation, relevant for the study of turbulence in tokamaks, leads to the so-called gyrokinetic model and brings the computational cost of the model within the presently available numerical resources. In this work, we study the transport of toroidal momentum in tokamaks in the framework of the gyrokinetic model. First, we show that this reduced model is indeed capable of accurately modeling momentum transport by deriving a local conservation equation of toroidal momentum, and verifying it numerically with the gyrokinetic code GYSELA. Secondly, we show how electrostatic turbulence can break the axisymmetry and generate toroidal rotation, while a strong link between turbulent heat and momentum transport is identified, as both exhibit the same large-scale avalanche-like events. The dynamics of turbulent transport are

  5. Bridge between fusion plasma and plasma processing

    International Nuclear Information System (INIS)

    Ohno, Noriyasu; Takamura, Shuichi

    2008-01-01

    In the present review, relationship between fusion plasma and processing plasma is discussed. From boundary-plasma studies in fusion devices new applications such as high-density plasma sources, erosion of graphite in a hydrogen plasma, formation of helium bubbles in high-melting-point metals and the use of toroidal plasmas for plasma processing are emerging. The authors would like to discuss a possibility of knowledge transfer from fusion plasmas to processing plasmas. (T. Ikehata)

  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. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    International Nuclear Information System (INIS)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

  8. Deuterium-tritium experiments on the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D. [Princeton Plasma Physics Lab., NJ (United States); Anderson, J.L.; Barnes, C.W. [Los Alamos National Lab., NM (United States)] [and others

    1994-11-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.

  9. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D. [and others

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.

  10. TSC simulation of the ohmic discharges for the tokamak plasmas

    International Nuclear Information System (INIS)

    Li Jiaxian; Pan Yudong; Zhang Jihua

    2010-01-01

    TSC (Tokamak simulation Code) is a famous numerical simulation code in fusion research, which is developed by Princeton Plasma Physics Laboratory. The code has been used to provide the physics studies to support the design of the poloidal field system in the new tokamak in Southwestern of Institute of Physics (SWIP). These studies have been performed in some depth. We briefly summarize the analysis methods used, but concentrate on giving the key results. This work firstly confirmed the plasma configuration designed by the SWEQU and EFIT code. We presented the main results of standard ohmic discharge evolution of the new tokamak in SWIP. The real-time feedback control system has been combined which considering the plasma displacement control and the current control. The output physical parameters have been detailed analyzed in this thesis, which including the plasma magnetic axis evolution, current profile evoultion, loop voltage evolution, electron temperature profile evolution, ion temperature profile evolution, density profile, pressure profile and so on. These parameters may supply the benchmarch to future experimental diagnostic information. (authors)

  11. Compact Commercial Tokamak Reactor (CCTR): a concept for a 500-MWe commercial-tokamak fusion system

    International Nuclear Information System (INIS)

    Gillen, T.J.

    1980-11-01

    A detailed set of self-consistent parameters and costs for the conceptual design of a Compact Commercial Tokamak Reactor (CCTR) is given. Several of the basic design features are the following: an ignited plasma with a major radius of 4.9 m and minor radius of 1.4 m; a net electrical output of 500 MW; a borated-water-cooled, stainless steel shield; and a toroidal field of 12 T at the coil. The design, which utilizes the Westinghouse computer code for the COsting And Sizing of D-T burning Tokamaks (COAST), mainly provides the sizes and geometries associated with the definition of the main component features for which a detailed engineering design can be effectively undertaken. Design study alternatives, including a neutral beam driven design option, a design option with a toroidal field of 13 T at the coil, and a tungsten-shielded option are considered for the CCTR. Also included is the conceptual design of a Compact Fusion Engineering Device

  12. Tokamak Plasmas: Electron temperature $(T_ {e}) $ measurements ...

    Indian Academy of Sciences (India)

    Home; Journals; Pramana – Journal of Physics; Volume 55; Issue 5-6. Tokamak Plasmas : Electron temperature ( T e ) measurements by Thomson scattering system. R Rajesh B Ramesh Kumar S K Varshney Manoj Kumar Chhaya Chavda Aruna Thakkar N C Patel Ajai Kumar Aditya Team. Contributed Papers Volume 55 ...

  13. Ion cyclotron transmission spectroscopy in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Greene, G.J.

    1993-09-01

    The propagation of waves in the ion cyclotron range of frequencies has been investigated experimentally in the Tokamak Fusion Test Reactor. A small, broadband, radiofrequency (rf) magnetic probe located outside the plasma limiter, at a major radius near that of the plasma center, was excited with a low power, frequency swept source (1--200 MHz). Waves propagating to a distant location were detected with a second, identical probe. The rf transmission spectrum revealed a region of attenuation over a band of frequencies for which the minority fundamental resonance was located between the outer plasma edge and the major radius of the probe location. Distinct, non-overlapping attenuation bands were observed from hydrogen and helium-3 minority species; a distinct tritium band should be observed in future DT experiments. Rapid spectrum acquisition during a helium-3 gas puff experiment showed that the wave attenuation involved the plasma core and was not a surface effect. A model in which the received power varied exponentially with the minority density, averaged over the resonance region, fit the time evolution of the probe signal relatively well. Estimation of a 1-d tunneling parameter from the experimental observations is discussed. Minority concentrations of less than 0.5 % can be resolved with this measurement.

  14. Fusion programs in Applied Plasma Physics

    International Nuclear Information System (INIS)

    1992-07-01

    The Applied Plasma Physics (APP) program at General Atomics (GA) described here includes four major elements: (a) Applied Plasma Physics Theory Program, (b) Alpha Particle Diagnostic, (c) Edge and Current Density Diagnostic, and (d) Fusion User Service Center (USC). The objective of the APP theoretical plasma physics research at GA is to support the DIII-D and other tokamak experiments and to significantly advance our ability to design a commercially-attractive fusion reactor. We categorize our efforts in three areas: magnetohydrodynamic (MHD) equilibria and stability; plasma transport with emphasis on H-mode, divertor, and boundary physics; and radio frequency (rf). The objective of the APP alpha particle diagnostic is to develop diagnostics of fast confined alpha particles using the interactions with the ablation cloud surrounding injected pellets and to develop diagnostic systems for reacting and ignited plasmas. The objective of the APP edge and current density diagnostic is to first develop a lithium beam diagnostic system for edge fluctuation studies on the Texas Experimental Tokamak (TEXT). The objective of the Fusion USC is to continue to provide maintenance and programming support to computer users in the GA fusion community. The detailed progress of each separate program covered in this report period is described in the following sections

  15. Liquid gallium jet-plasma interaction studies in ISTTOK tokamak

    International Nuclear Information System (INIS)

    Gomes, R.B.; Fernandes, H.; Silva, C.; Sarakovskis, A.; Pereira, T.; Figueiredo, J.; Carvalho, B.; Soares, A.; Duarte, P.; Varandas, C.; Lielausis, O.; Klyukin, A.; Platacis, E.; Tale, I.; Alekseyv, A.

    2009-01-01

    Liquid metals have been pointed out as a suitable solution to solve problems related to the use of solid walls submitted to high power loads allowing, simultaneously, an efficient heat exhaustion process from fusion devices. The most promising candidate materials are lithium and gallium. However, lithium has a short liquid state temperature range when compared with gallium. To explore further this property, ISTTOK tokamak is being used to test the interaction of a free flying liquid gallium jet with the plasma. ISTTOK has been successfully operated with this jet without noticeable discharge degradation and no severe effect on the main plasma parameters or a significant plasma contamination by liquid metal. Additionally the response of an infrared sensor, intended to measure the jet surface temperature increase during its interaction with the plasma, has been studied. The jet power extraction capability is extrapolated from the heat flux profiles measured in ISTTOK plasmas.

  16. 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 ...

  17. 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

  18. Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

    Science.gov (United States)

    Greenfield, C. M.; DIII-D Team

    2005-01-01

    Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available.

  19. Advanced tokamak research at the DIII-D National Fusion Facility in support of ITER

    International Nuclear Information System (INIS)

    Greenfield, C M

    2005-01-01

    Fusion energy research aims to develop an economically and environmentally sustainable energy system. The tokamak, a doughnut shaped plasma confined by magnetic fields generated by currents flowing in external coils and the plasma, is a leading concept. Advanced Tokamak (AT) research in the DIII-D tokamak seeks to provide a scientific basis for steady-state high performance operation. This necessitates replacing the inherently pulsed inductive method of driving plasma current. Our approach emphasizes high pressure to maximize fusion gain while maximizing the self-driven bootstrap current, along with external current profile control. This requires integrated, simultaneous control of many characteristics of the plasma with a diverse set of techniques. This has already resulted in noninductive conditions being maintained at high pressure on current relaxation timescales. A high degree of physical understanding is facilitated by a closely coupled integrated modelling effort. Simulations are used both to plan and interpret experiments, making possible continued development of the models themselves. An ultimate objective is the capability to predict behaviour in future AT experiments. Analysis of experimental results relies on use of the TRANSP code via the FusionGrid, and our use of the FusionGrid will increase as additional analysis and simulation tools are made available

  20. A general comparison between tokamak and stellarator plasmas

    Directory of Open Access Journals (Sweden)

    Yuhong Xu

    2016-07-01

    Full Text Available This paper generally compares the essential features between tokamaks and stellarators, based on previous review work individually made by authors on several specific topics, such as theories, bulk plasma transport and edge divertor physics, along with some recent results. It aims at summarizing the main results and conclusions with regard to the advantages and disadvantages in these two types of magnetic fusion devices. The comparison includes basic magnetic configurations, magnetohydrodynamic (MHD instabilities, operational limits and disruptions, neoclassical and turbulent transport, confinement scaling and isotopic effects, plasma rotation, and edge and divertor physics. Finally, a concept of quasi-symmetric stellarators is briefly referred along with a comparison of future application for fusion reactors.

  1. Power supplies for plasma column control in COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Havlíček, Josef; Hauptmann, R.; Peroutka, Oldřich; Tadros, Momtaz; Hron, Martin; Janky, Filip; Vondráček, Petr; Cahyna, Pavel; Mikulín, Ondřej; Šesták, David; Junek, Pavel; Pánek, Radomír

    2013-01-01

    Roč. 88, 9-10 (2013), s. 1640-1645 ISSN 0920-3796. [Symposium on Fusion Technology ( SOFT -27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * Power supplies * Feedback control * Vertical displacement * Vertical kicks Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001543#

  2. Technology issues for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

    The approach for decommissioning the Tokamak Fusion Test Reactor has evolved from a conservative plan based on cutting up and burying all of the systems, to one that considers the impact tritium contamination will have on waste disposal, how large size components may be used as their own shipping containers, and even the possibility of recycling the materials of components such as the toroidal field coils and the tokamak structure. In addition, the project is more carefully assessing the requirements for using remotely operated equipment. Finally, valuable cost database is being developed for future use by the fusion community

  3. 10th International Conference and School on Plasma Physics and Controlled Fusion. Book of Abstracts

    International Nuclear Information System (INIS)

    Anon

    2004-01-01

    About 240 abstracts by Ukrainian and foreign authors submitted to 10-th International Conference and School on Plasma Physics and Controlled fusion have been considered by Conference Program Committee members. All the abstracts have been divided into 8 groups: magnetic confinement systems: stellarators, tokamaks, alternative conceptions; ITER and Fusion reactor aspects; basic plasma physics; space plasma; plasma dynamics and plasma-wall interaction; plasma electronics; low temperature plasma and plasma technologies; plasma diagnostics

  4. Abstracts of the 23rd European physical society conference on controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    Goutych, I.F.; Gresillon, D.; Sitenko, A.G.

    1996-01-01

    This document contains the abstracts of the invited and contributed papers presented at 23 EPS conference on controlled fusion and plasma physics. The main contents are: tokamaks, stellarators; alternative magnetic confinement; plasma edge physics; plasma heating and current drive; plasma diagnostics; basic collisionless plasma physics; high intensity laser produced plasmas and inertial confinement; low-temperature plasmas

  5. Remote operation of the vertical plasma stabilization @ the GOLEM tokamak for the plasma physics education

    Czech Academy of Sciences Publication Activity Database

    Svoboda, V.; Kocman, J.; Grover, O.; Krbec, Jaroslav; Stöckel, Jan

    96-97, October (2015), s. 974-979 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] Institutional support: RVO:61389021 Keywords : tokamak technology * remote participation * plasma stabilization Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://dx.doi.org/10.1016/j.fusengdes.2015.06.044

  6. A quasi-linear gyrokinetic transport model for tokamak plasmas

    International Nuclear Information System (INIS)

    Casati, A.

    2009-10-01

    After a presentation of some basics around nuclear fusion, this research thesis introduces the framework of the tokamak strategy to deal with confinement, hence the main plasma instabilities which are responsible for turbulent transport of energy and matter in such a system. The author also briefly introduces the two principal plasma representations, the fluid and the kinetic ones. He explains why the gyro-kinetic approach has been preferred. A tokamak relevant case is presented in order to highlight the relevance of a correct accounting of the kinetic wave-particle resonance. He discusses the issue of the quasi-linear response. Firstly, the derivation of the model, called QuaLiKiz, and its underlying hypotheses to get the energy and the particle turbulent flux are presented. Secondly, the validity of the quasi-linear response is verified against the nonlinear gyro-kinetic simulations. The saturation model that is assumed in QuaLiKiz, is presented and discussed. Then, the author qualifies the global outcomes of QuaLiKiz. Both the quasi-linear energy and the particle flux are compared to the expectations from the nonlinear simulations, across a wide scan of tokamak relevant parameters. Therefore, the coupling of QuaLiKiz within the integrated transport solver CRONOS is presented: this procedure allows the time-dependent transport problem to be solved, hence the direct application of the model to the experiment. The first preliminary results regarding the experimental analysis are finally discussed

  7. Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak

    International Nuclear Information System (INIS)

    Barbosa, Luis Filipe F.P.W.; Bosco, Edson del.

    1994-01-01

    This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs

  8. Development of a tokamak plasma optimized for stability and confinement

    International Nuclear Information System (INIS)

    Politzer, P.A.

    1995-02-01

    Design of an economically attractive tokamak fusion reactor depends on producing steady-state plasma operation with simultaneous high energy density (β) and high energy confinement (τ E ); either of these, by itself, is insufficient. In operation of the DIII-D tokamak, both high confinement enhancement (H≡ τ E /τ ITER-89P = 4) and high normalized β (β N ≡ β/(I/aB) = 6%-m-T/MA) have been obtained. For the present, these conditions have been produced separately and in transient discharges. The DIII-D advanced tokamak development program is directed toward developing an understanding of the characteristics which lead to high stability and confinement, and to use that understanding to demonstrate stationary, high performance operation through active control of the plasma shape and profiles. The authors have identified some of the features of the operating modes in DIII-D that contribute to better performance. These are control of the plasma shape, control of both bulk plasma rotation and shear in the rotation and Er profiles, and particularly control of the toroidal current profiles. In order to guide their future experiments, they are developing optimized scenarios based on their anticipated plasma control capabilities, particularly using fast wave current drive (on-axis) and electron cyclotron current drive (off-axis). The most highly developed model is the second-stable core VH-mode, which has a reversed magnetic shear safety factor profile [q(O) = 3.9, q min = 2.6, and q 95 = 6]. This model plasma uses profiles which the authors expect to be realizable. At β N ≥ 6, it is stable to n=l kink modes and ideal ballooning modes, and is expected to reach H ≥ 3 with VH-mode-like confinement

  9. Modelling and control of a tokamak plasma

    International Nuclear Information System (INIS)

    Bremond, S.

    1995-01-01

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes

  10. 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

  11. High beta plasmas in the PBX tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bol, K.; Buchenauer, D.; Chance, M.; Couture, P.; Fishman, H.; Fonck, R.; Gammel, G.; Grek, B.; Ida, K.; Itami, K.

    1986-04-01

    Bean-shaped configurations favorable for high ..beta.. 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 ..beta.. limit.

  12. Remote operation of the GOLEM tokamak for Fusion Education

    Czech Academy of Sciences Publication Activity Database

    Grover, O.; Kocman, J.; Odstrčil, M.; Odstrčil, T.; Matušů, M.; Stöckel, Jan; Svoboda, V.; Vondrášek, G.; Žára, J.

    2016-01-01

    Roč. 112, November (2016), s. 1038-1044 ISSN 0920-3796. [Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research IAEA /10./. Ahmedabad, 20.04.2015-24.04.2015] Institutional support: RVO:61389021 Keywords : Tokamak technology * Remote participation * Education * Nuclear fusion Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379616303441

  13. Electron cyclotron emission from tokamak plasmas

    International Nuclear Information System (INIS)

    Sillen, R.M.J.

    1986-01-01

    Emitted electron radiation can be used as a diagnostic signal to measure the electron temperature of a thermonuclear plasma. This type of diagnostics is well established in tokamak physics. In ch. 2 of this thesis the development, calibration and special design features are treated of a six-channel prototype of a twelve-channel grating spectrometer which is built for JET at Culham for electron cyclotron emission (ECE) measurements. In order to test this prototype measurements have been performed with the T-10 tokamak at the Kurchatov Institute in Moscow. With this prototype nearly half of the temperature profile of the T-10 could be measured. Detailed observations of sawteeth instabilities have been performed. Plasma heating by electron cyclotron resonance heating experiments was studied. A detailed description of these measurements and results is given in ch. 3. Often ECE spectra from tokamaks showed non-thermal features. In order to interprete them a computer code Notec has been developed. This code that calculates the ECE radiation emerging from the plasma for a 3-D configuration, is described in ch. 4. Some preliminary results and applications are presented. (Auth.)

  14. Study of intelligent system for control of the tokamak-ETE plasma positioning

    International Nuclear Information System (INIS)

    Barbosa, Luis Filipe de Faria Pereira Wiltgen

    2003-01-01

    The development of an intelligent neural control system of the neural type, capable to perform real time control of the plasma displacement in the experiment tokamak spheric - ETE (spherical tokamak experiment ) is presented. The ETE machine is in operation since Nov 2000, in the LAP - Plasma Associated Laboratory of the Brazilian Institute on Spatial Research (INPE) in Sao Jose dos Campos, S P, Brazil. The experiment is dedicated to study the magnetic confinement of a fusion plasma in a configuration favorable for the construction of future reactors. Nuclear fusion constitutes a renewable energy source with low environmental impact, which uses atomic energy in pacific applications for the sustainable development of humanity. One of the important questions for the attainment of fusion relates to the stability of the plasma and control of its position during the reactor operation. Therefore, the development of systems to control the plasma in tokamaks constitutes a necessary technological advance for the feasibility of nuclear fusion. In particular, the research carried out in this thesis concerns the proposal of a system to control the vertical displacement of the plasma in the ETE tokamak, aiming to obtain steady pulses in this machine. A Magnetic Levitation system (Mag Lev) was developed as part of this work, allowing to study the nonlinear behavior of a device that, from the aspect of position control, is similar (analogous) to the plasma in the ETE tokamak, This magnetic levitation system was designed, mathematically modeled and built in order to test both classical and intelligent type controllers. The results of this comparison are very promising for the use of intelligent controllers in the ETE tokamak as well as other control applications. (author)

  15. Tokamak Fusion Test Reactor neutral beam injection system vacuum chamber

    International Nuclear Information System (INIS)

    Pedrotti, L.R.

    1977-01-01

    Most of the components of the Neutral Beam Lines of the Tokamak Fusion Test Reactor (TFTR) will be enclosed in a 50 cubic meter box-shaped vacuum chamber. The chamber will have a number of unorthodox features to accomodate both neutral beam and TFTR requirements. The design constraints, and the resulting chamber design, are presented

  16. Edge plasma physical investigations of tokamak plasmas in CRIP

    International Nuclear Information System (INIS)

    Bakos, J.; Ignacz, P.; Koltai, L.; Paszti, F.; Petravich, G.; Szigeti, J.; Zoletnik, S.

    1988-01-01

    The results of the measurements performed in the field of thermonuclear high temperature plasma physics in CRIP (Hungary) are summarized. In the field of the edge plasma physics solid probes were used to test the external zone of plasma edges, and atom beams and balls were used to investigate both the external and internal zones. The plasma density distribution was measured by laser blow-off technics, using Na atoms, which are evaporated by laser pulses. The excitation of Na atom ball by tokamak plasma gives information on the status of the plasma edge. The toroidal asymmetry of particle transport in tokamak plasma was measured by erosion probes. The evaporated and transported impurities were collected on an other part of the plasma edge and were analyzed by SIMS and Rutherford backscattering. The interactions in plasma near the limiter were investigated by a special limiter with implemented probes. Recycling and charge exchange processes were measured. Disruption phenomena of tokamak plasma were analyzed and a special kind of disruptions, 'soft disruptions' and the related preliminary perturbations were discovered. (D.Gy.) 10 figs

  17. Edge plasma fluctuations measurements in fusion experiments

    International Nuclear Information System (INIS)

    Schrittwieser, R.; Ionitha, C.; Balan, P.C.; Varandas, C.A.F.; Figueiredo, H.F.C.; Silva, C.; Stoeckel, J.; Adamek, J.; Hron, M.; Tichy, M.; Hidalgo, C.; Pedrosa, M.A.; Calderon, E.; Martines, E.; Van Oost, G.; Rasmussen, J.J.; Naulin, V.

    2005-01-01

    We report on investigations on electrostatic fluctuations in the edge plasma region which have been carried out during the last few years at several European fusion experiments. Various methods and probe arrangements have been used to determine fluctuations of the plasma potential, the electric field and the electron temperature. Investigations were under-taken in ISTTOK (Instituto Superior Tecnico TOKamak), Lisbon, Portugal, in CASTOR (Czech Academy of Science TORus), Prague, Czech Republic, and the TJ-II Flexible Heliac at CIEMAT in Madrid, Spain. (author)

  18. Trade studies of plasma elongation for next-step tokamaks

    International Nuclear Information System (INIS)

    Galambos, J.D.; Strickler, D.J.; Peng, Y.K.M.; Reid, R.L.

    1988-09-01

    The effect of elongation on minimum-cost devices is investigated for elongations ranging from 2 to 3. The analysis, carried out with the TETRA tokamak systems code, includes the effects of elongation on both physics (plasma beta limit) and engineering (poloidal field coil currents) issues. When ignition is required, the minimum cost occurs for elongations from 2.3 to 2.9, depending on the plasma energy confinement scaling used. Scalings that include favorable plasma current dependence and/or degradation with fusion power tend to have minimum cost at higher elongation (2.5-2.9); scalings that depend primarily on size result in lower elongation (/approximately/2.3) for minimum cost. For design concepts that include steady-state current-driven operation, minimum cost occurs at an elongation of 2.3. 12 refs., 13 figs

  19. Safety aspects of activation products in a compact Tokamak Fusion Power Plant

    International Nuclear Information System (INIS)

    Willenberg, H.J.; Bickford, W.E.

    1978-10-01

    Neutron activation of materials in a compact tokamak fusion reactor has been investigated. Results of activation product inventory, dose rate, and decay heat calculations in the blanket and injectors are presented for a reactor design with stainless steel structures. Routine transport of activated materials into the plasma and vacuum systems is discussed. Accidental release of radioactive materials as a result of liquid lithium spills is also considered

  20. Recent developments in engineering and technology concepts for prospective tokamak fusion reactors

    International Nuclear Information System (INIS)

    Ford, G.W.K.

    1987-01-01

    The tokamak has become the most developed magnetic fusion system and it appears likely that break-even and possibly ignition will first be demonstrated in existing machines of this type. Yet larger tokamaks could also demonstrate the essential technologies for the production of useful power. World-wide, well over a hundred tritium-breeder/heat-removal blanket concepts have been devised and preliminary engineering design studies undertaken, but the effort deployed on breeding and power recovery systems has been very small compared with that assigned to plasma research and development. The European Communities' NET (Next European Torus) project may offer an opportunity to redress this imbalance. The NET pre-design stage now in progress for some three years has selected many of the best features of plasma and nuclear design from the world's total efforts in these fields, and the NET concept is described in this paper as exemplifying where magnetic fusion power reactor technology stands today. It is concluded that although there are numerous more advanced types of magnetic confinement fusion reactor at early stages of their physics development, the tokamak offers the best opportunity for the early demonstration of fusion power

  1. Study of electron density and its fluctuations in tokamaks plasmas by fast infrared interferometry

    International Nuclear Information System (INIS)

    Ryter, F.

    1982-10-01

    The electron density knowledge in tokamak plasma is fundamental for controlled fusion research. Its study can be made by interferometric measurement of plasma refraction index. Density and density fluctuation measurements are given for present and future tokamak, the wavelength used must be in the far infrared. The interferometer used type employs two identical lasers. Waveguide type submillimetric lasers, optically pumped by a CO 2 laser, have been developed and optimized. Detectors used are Schottky diodes. The interferometer allows a radial study of the plasma and presents a great stability during the measurement [fr

  2. Fusion

    International Nuclear Information System (INIS)

    Naraghi, M.

    1976-01-01

    It is proposed that Iran as a world's potential supplier of fossile fuel should participate in fusion research and gain experience in this new field. Fusion, as an ultimate source of energy in future, and the problems concerned with the fusion reactors are reviewed. Furthermore; plasma heating, magnetic and inertial confinement in a fusion reactor are discussed. A brief description of tokamak, theta pinch and magnetic mirror reactors is also included

  3. Materials experience with tokamak plasmas

    International Nuclear Information System (INIS)

    Colchin, R.J.

    1979-01-01

    The primary effects of using various wall and limiter materials have been in the amount and kind of impurities they introduce into the plasma. Only a limited number of materials have been employed to date. These include gold, stainless steel, inconel, glass, alumina, and titanium for first walls and carbon, stainless steel, inconel, alumina, silicon carbide, boron carbide, molybdenum, tantalum, titanium, and tungsten for limiters and divertors. Limiter surfaces bear the brunt of the plasma bombardment and so typically introduce impurities far out of proportion to their relative size. The ratio of the bombarded limiter area to that of the first wall is of order 1 to 10 3

  4. Materials experience with tokamak plasmas

    International Nuclear Information System (INIS)

    Colchin, R.J.

    1979-01-01

    The primary effects of using various wall and limiter materials have been in the amount and kind of impurities they introduce into the plasma. The materials employed to date include gold, stainless steel, inconel, glass, alumina, and titanium for first walls and carbon, stainless steel, inconel, alumina, silicon carbide, boron carbide, molybdenum, tantalum, titanium, and tungsten for limiters and divertors. Limiter surfaces bear the brunt of the plasma bombardment and so typically introduce impurities far out of proportion to their relative size. The ratio of the bombarded limiter area to that of the first wall is of order 1 to 10 3

  5. Current drive for spherical tokamak plasmas

    International Nuclear Information System (INIS)

    Storer, R.

    1999-01-01

    Very low aspect ratio spherical tokamaks have proved to have some very useful and remarkable properties including very high values of the plasma pressure to magnetic field pressure. Following the construction of the Start tokamak, a number of such configurations have been constructed. One of the difficulties encountered is in providing sufficient inductive current drive due to the competing requirements of the need to keep the aspect ratio low and providing the space for the central current-carrying rod with an internal inductive coil. An alternative current drive technique would be very useful. In a parallel development it has been shown that a rotating magnetic field can drive a significant non-linear Hall current in a spherical plasma. Successful experiments of this concept have been made with a device called the Rotamak. In its original configuration this device was a field reversed configuration without a toroidal magnetic field but with a vertical field to establish the magnetic hydrodynamical equilibrium. However, recent modifications have shown that increased current can be driven if a central current-carrying rod is used to provide an applied toroidal field. The new Rotamak has then a spherical tokamak magnetic field structure. This work will present new calculations which model the above structure and include the effect of the applied toroidal field in addition to the steady vertical field and the rotating (current-drive) magnetic field. The problem is fully three dimensional and non-linear and involves the application of interesting computational techniques. The potential of using the rotating field current drive technique for spherical tokamaks will be evaluated

  6. Sawtooth driven particle transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Nicolas, T.

    2013-01-01

    The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author) [fr

  7. Turbulent ion heating in TCV Tokamak plasmas

    International Nuclear Information System (INIS)

    Schlatter, Ch.

    2009-08-01

    The Tokamak à configuration variable (TCV) features the highest electron cyclotron wave power density available to resonantly heat (ECRH) the electrons and to drive noninductive currents in a fusion grade plasma (ECCD). In more than 15 years of exploitation, much effort has been expended on real and velocity space engineering of the plasma electron energy distribution function and thus making electron physics a major research contribution of TCV. When a plasma was first subjected to ECCD, a surprising energisation of the ions, perpendicular to the confining magnetic field, was observed on the charge exchange spectrum measured with the vertical neutral particle analyser (VNPA). It was soon concluded that the ion acceleration was not due to power equipartition between electrons and ions, which, due to the absence of direct ion heating on TCV, has thus far been considered as the only mechanism heating the ions. However, although observed for more than ten years, little attention was paid to this phenomenon, whose cause has remained unexplained to date. The key subject of this thesis is the experimental study of this anomalous ion acceleration, the characterisation in terms of relevant parameters and the presentation of a model simulation of the potential process responsible for the appearance of fast ions. The installation of a new compact neutral particle analyser (CNPA) with an extended high energy range (≥ 50 keV) greatly improved the fast ion properties diagnosis. The CNPA was commissioned and the information derived from its measurement (ion temperature and density, isotopic plasma composition) was validated against other ion diagnostics, namely the active carbon charge exchange recombination spectroscopy system (CXRS) and a neutron counter. In ohmic plasmas, where the ion heating agrees with classical theory, the radial ion temperature profile was successfully reconstructed by vertically displacing the plasma across the horizontal CNPA line of sight. Active

  8. 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%

  9. Control of a fusion burning in a tokamak with the help of fueling

    International Nuclear Information System (INIS)

    Bespoludennov, S.G.; Galkin, S.A.; AN SSSR, Moscow

    1990-01-01

    Analytical model to control fusion burning in tokamak using fueling is developed and studied. Model is based on two traditional differential equations of particle and energy balance, control of thermonuclear power is conducted using proportional feedback between plasma density and thermal energy characterized by β parameters. Range of control parameters providing for stationary fusion burning along low-temperature brand of ignition curve is determined. Comparison of predictions of analytical model with calculations of burning stabilization is conducted according to QUET sesquidimensional transport code. Good conformity between analytical estimations and calculation results is shown

  10. The effective cost of tritium for tokamak fusion power reactors with reduced tritium production systems

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Evans, K.

    1983-01-01

    If sufficient tritium cannot be produced and processed in tokamak blankets then at least two alternatives are possible. Tritium can be purchased; or reactors with reduced tritium (RT) content in the plasma can be designed. The latter choice may require development of magnet technology etc., but the authors show that the impact on the cost-of-electricity may be mild. Cost tradeoffs are compared to the market value of tritium. Adequate tritium production in fusion blankets is preferred, but the authors show there is some flexibility in the deployment of fusion if this is not possible

  11. Plasma startup patterns in tokamak reactors

    International Nuclear Information System (INIS)

    Maki, Koichi; Tone, Tatsuzo.

    1983-01-01

    Plasma startup patterns are studied from the viewpoint of net power loss represented by the total power loss less the α-particle heating power. The existence is shown of a critical temperature of plasma at which the net power loss becomes independent of plasma density. Observations are made which indicate that the net power loss decreases with lowering plasma density in the range below the critical temperature and vice versa, whether governed by empirical or trapped-ion scaling laws. A startup pattern is presented which minimizes the net power loss during startup, and which prescribes that: (1) The plasma density should be kept as low as possible until the plasma is heated up to the critical temperature; (2) thereafter, the plasma density should be increased to its steady state value while retaining the critical temperature; and (3) finally, with the density kept constant, the temperature should be further raised to its steady state value. The net power loss at critical temperature represents the lower limit of heating power required to bring the plasma to steady state in tokamak reactors. (author)

  12. Viscosity in the edge of tokamak plasmas

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1993-05-01

    A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the ''short-radial-gradient-scale-length'' (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates

  13. Plasma physics and controlled nuclear fusion research 1990. V. 2

    International Nuclear Information System (INIS)

    1991-01-01

    Volume 2 of the Proceedings of the Thirteenth International Conference on Plasma Physics and Controlled Fusion Research contains papers in two sessions: the first part, on Magnetic Confinement Theory and Modelling, was presented in session D at the conference; the second part, on Non-Tokamak Confinement Systems, was given in session C of the conference. Abstracts accompany each paper. Refs, figs and tabs

  14. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

    1992-01-01

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  15. Fundamentals of plasma physics and controlled fusion

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Kenro

    2000-10-01

    The present lecture note was written to fill a gap between text books for undergraduates and specific review articles written by specialists for their young colleagues. The note may be divided in three parts. The first part is on basic characteristics of a plasma in a magnetic field. The second part describes plasma confinement and heating with an emphasis on magnetohydrodynamic instabilities. In addition, propagation of plasma waves, plasma heating by electromagnetic waves are given. The third part is devoted to various specific concepts of nuclear fusion. Emphases are placed on toroidal devices, especially on tokamak devices and stellarators. One might feel heavy mathematics glimpsing the present note, especially in the part treating magnetohydrodynamic instabilities. (author)

  16. First fusion proton measurements in TEXTOR plasmas using activation technique

    Energy Technology Data Exchange (ETDEWEB)

    Bonheure, G.; Wassenhove, G. Van [ERM-KMS, Trilateral Euregio Cluster, B-1000 Brussels (Belgium); Mlynar, J. [Association Euratom-IPP.CR, Institute of Plasma Physics AS CR, v.v.i., Za Slovankou 3, CZ-182 00 Praha 8 (Czech Republic); Hult, M.; Gonzalez de Orduna, R.; Lutter, G. [Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Vermaercke, P. [SCK-CEN, Boeretang, B-2400 Mol (Belgium); Huber, A.; Schweer, B.; Esser, G.; Biel, W. [Forschungszentrum Juelich GmbH, Institut fuer Plasmaphysik, EURATOM-Assoziation, Trilateral Euregio Cluster, D-52425 Juelich (Germany)

    2012-10-15

    MeV particle loss measurements from fusion plasmas, in particular alpha particles, remain difficult in large fusion devices and further R and D is needed for ITER. This paper describes the first attempt to measure 3 MeV escaping fusion protons emitted from TEXTOR tokamak plasmas using activation technique. This technique was successfully demonstrated, initially, in 2006 on the JET tokamak. An ion camera equipped with a collimator and several types of activation detectors was installed inside the TEXTOR vacuum vessel to perform these measurements. After irradiation, the detectors were analyzed using ultra low level gamma-ray spectrometry at the HADES underground laboratory. 3 MeV escaping fusion protons were detected in larger number -{approx}6 times more - compared to earlier measurements using this technique on JET. Another major progress was the reduction of the cooling time by a factor of 50, which made possible to detect radionuclides with half-life of less than 90 min.

  17. Interaction of a spheromak-like compact toroid with a high beta spherical tokamak plasma

    International Nuclear Information System (INIS)

    Hwang, D.Q.; McLean, H.S.; Baker, K.L.; Evans, R.W.; Horton, R.D.; Terry, S.D.; Howard, S.; Schmidt, G.L.

    2000-01-01

    Recent experiments using accelerated spheromak-like compact toroids (SCTs) to fuel tokamak plasmas have quantified the penetration mechanism in the low beta regime; i.e. external magnetic field pressure dominates plasma thermal pressure. However, fusion reactor designs require high beta plasma and, more importantly, the proper plasma pressure profile. Here, the effect of the plasma pressure profile on SCT penetration, specifically, the effect of diamagnetism, is addressed. It is estimated that magnetic field pressure dominates penetration even up to 50% local beta. The combination of the diamagnetic effect on the toroidal magnetic field and the strong poloidal field at the outer major radius of a spherical tokamak will result in a diamagnetic well in the total magnetic field. Therefore, the spherical tokamak is a good candidate to test the potential trapping of an SCT in a high beta diamagnetic well. The diamagnetic effects of a high beta spherical tokamak discharge (low aspect ratio) are computed. To test the penetration of an SCT into such a diamagnetic well, experiments have been conducted of SCT injection into a vacuum field structure which simulates the diamagnetic field effect of a high beta tokamak. The diamagnetic field gradient length is substantially shorter than that of the toroidal field of the tokamak, and the results show that it can still improve the penetration of the SCT. Finally, analytic results have been used to estimate the effect of plasma pressure on penetration, and the effect of plasma pressure was found to be small in comparison with the magnetic field pressure. The penetration condition for a vacuum field only is reported. To study the diamagnetic effect in a high beta plasma, additional experiments need to be carried out on a high beta spherical tokamak. (author)

  18. In-vessel maintenance concepts for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Yount, J.A.

    1983-01-01

    Concepts for rail-mounted and guided in-vessel handling machines (IVM) for remote maintenance inside tokamak fusion reactors are described. The IVM designs are based on concepts for tethered remotely operated vehicles and feature the use of multiple manipulator arms for remote handling and remote-controlled TV cameras for remote viewing. The concepts include IVMs for both single or dual rail systems located in the top or bottom of the reactor vessel

  19. Stability of tearing modes in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Hegna, C.C.; Callen, J.D.

    1994-02-01

    The stability properties of m {ge} 2 tearing instabilities in tokamak plasmas are analyzed. A boundary layer theory is used to find asymptotic solutions to the ideal external kink equation which are used to obtain a simple analytic expression for the tearing instability parameter {Delta}{prime}. This calculation generalizes previous work on this topic by considering more general toroidal equilibria (however, toroidal coupling effects are ignored). Constructions of {Delta}{prime} are obtained for plasmas with finite beta and for islands that have nonzero width. A simple heuristic estimate is given for the value of the saturated island width when the instability criterion is violated. A connection is made between the calculation of the asymptotic matching parameter in the finite beta and island width case to the nonlinear analog of the Glasser effect.

  20. Technology and plasma-materials interaction processes of tokamak disruptions

    International Nuclear Information System (INIS)

    McGrath, R.T.; Kellman, A.G.

    1992-01-01

    A workshop on the technology and plasma-materials interaction processes of tokamak disruptions was held April 3, 1992 in Monterey, California, as a satellite meeting of the 10th International Conference on Plasma-Surface Interactions. The objective was to bring together researchers working on disruption measurements in operating tokamaks, those performing disruption simulation experiments using pulsed plasma gun, electron beam and laser systems, and computational physicists attempting to model the evolution and plasma-materials interaction processes of tokamak disruptions. This is a brief report on the workshop. 4 refs

  1. Net energy balance of tokamak fusion power plants

    International Nuclear Information System (INIS)

    Buende, R.

    1981-10-01

    The net energy balance for a tokamak fusion power plant was determined by using a PWR power plant as reference system, replacing the fission-specific components by fusion-specific components and adjusting the non-reactor-specific components to altered conditions. For determining the energy input to the fusion plant a method was developed that combines the advantages of the energetic input-output method with those of process chain analysis. A comparison with PWR, HTR, FBR, and coal-fired power plants is made. As a result the net energy balance of the fusion power plant turns out to be more advantageous than that of an LWR, HTR or coal-fired power plant and nearly in the same range as FBR power plants. (orig.)

  2. Net energy balance of tokamak fusion power plants

    International Nuclear Information System (INIS)

    Buende, R.

    1983-01-01

    The net energy balance for a tokamak fusion power plant of present day design is determined by using a PWR power plant as reference system, replacing the fission-specific components by fusion-specific components and adjusting the non-reactor-specific components to altered conditions. For determining the energy input to the fusion plant a method was developed that combines the advantages of the energetic input-output method with those of process chain analysis. A comparison with PWR, HTR, FBR, and coal-fired power plants is made. As a result the energy expenditures of the fusion power plant turn out to be lower than that of an LWR, HTR, or coal-fired power plant of equal net electric power output and nearly in the same range as FBR power plants. (orig.)

  3. Plasma diagnostics for the compact ignition tokamak

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  4. Impact of melt-layer ejection from metallic first wall on tokamak plasmas

    Science.gov (United States)

    Smirnov, R. D.; Krasheninnikov, S. I.; Pigarov, A. Yu.; Rognlien, T. D.

    2012-10-01

    At present, all-metallic tokamak first wall is preferred over carbon composite materials for next generation fusion devices, such as ITER, due to favorable thermo-physical and chemical properties of metals in fusion plasma environment. However, recent experiments demonstrate that surface of metallic components, including tungsten ones, under high transient heat load pertinent to next step tokamaks can melt and eject molten material into fusion plasma in form of droplets or fine spray [1]. The ejected material can be a source of impurity contamination of fusion plasmas and even in some cases cause discharge termination, as was observed recently on LHD. In this work, we investigate impact of ejection of beryllium droplets of various sizes on ITER-like plasmas using coupled dust-plasma edge transport code DUSTT/UEDGE [2]. Different ejection scenarios are modeled, including intermittent and prolonged ejection of molten material at the top, midplane and divertor poloidal locations in ITER. Using the modeling we assess modifications of the plasma profiles, radiation power losses, and impurity particle fluxes to the plasma core produced by various quantities of the ejectile. Critical amounts of the different materials ejected, which can lead to discharge termination, are evaluated.[4pt] [1] J.W. Coenen, et al., Nucl. Fusion 51 (2011) 113020;[0pt] [2] R.D. Smirnov, et al., J. Nucl. Mater. 415 (2011) S1067.

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

    International Nuclear Information System (INIS)

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

    1985-08-01

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

  6. NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0. 8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-08-01

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

  7. Measurement of Sheared Flows in the Edge Plasma of the CASTOR Tokamak

    Czech Academy of Sciences Publication Activity Database

    Brotánková, Jana; Stöckel, Jan; Horáček, Jan; Seidl, Jakub; Ďuran, Ivan; Hron, Martin; Van Oost, G.

    2009-01-01

    Roč. 35, č. 11 (2009), s. 980-986 ISSN 1063-780X. [IAEA Technical Meeting on Research Using Small Fusion Devices/18th./. Alushta (Krym), 25.09. 2008 -27.09. 2008 ] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * probe diagnostics * sheared flows * edge plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.584, year: 2009 http://www.springerlink.com/content/u571504gmq118314/

  8. A review of the methods to measure the ion temperature in a tokamak plasma

    International Nuclear Information System (INIS)

    Zurro Hernandez, B.; Perez-Navarro Gomez, A.

    1976-01-01

    The most important methods to measure the ion temperatu--re in a Tokamak plasma are reviewed, e.g. energy analysis of the fast neutrals which leave out the plasma, Doppler broadening of the emision spectral lines and fusion neutron analysis. It is discussed their bounds so as the advantages and drawbacks of each one. Other methods of some interest in the future are outlined. (author) [es

  9. BOOK REVIEW: Controlled Fusion and Plasma Physics

    Science.gov (United States)

    Engelmann, F.

    2007-07-01

    This new book by Kenro Miyamoto provides an up-to-date overview of the status of fusion research and the important parts of the underlying plasma physics at a moment where, due to the start of ITER construction, an important step in fusion research has been made and many new research workers will enter the field. For them, and also for interested graduate students and physicists in other fields, the book provides a good introduction into fusion physics as, on the whole, the presentation of the material is quite appropriate for getting acquainted with the field on the basis of just general knowledge in physics. There is overlap with Miyamoto's earlier book Plasma Physics for Nuclear Fusion (MIT Press, Cambridge, USA, 1989) but only in a few sections on subjects which have not evolved since. The presentation is subdivided into two parts of about equal length. The first part, following a concise survey of the physics basis of thermonuclear fusion and of plasmas in general, covers the various magnetic configurations studied for plasma confinement (tokamak; reversed field pinch; stellarator; mirror-type geometries) and introduces the specific properties of plasmas in these devices. Plasma confinement in tokamaks is treated in particular detail, in compliance with the importance of this field in fusion research. This includes a review of the ITER concept and of the rationale for the choice of ITER's parameters. In the second part, selected topics in fusion plasma physics (macroscopic instabilities; propagation of waves; kinetic effects such as energy transfer between waves and particles including microscopic instabilities as well as plasma heating and current drive; transport phenomena induced by turbulence) are presented systematically. While the emphasis is on displaying the essential physics, deeper theoretical analysis is also provided here. Every chapter is complemented by a few related problems, but only partial hints for their solution are given. A selection of

  10. The ARIES-ST study: Assessment of the spherical tokamak concept as fusion power plants

    International Nuclear Information System (INIS)

    Najmabadi, F.; Tillack, M.; Miller, R.; Mau, T.K.; Jardin, S.; Stambaugh, R.; Steiner, D.; Waganer, L.

    2001-01-01

    Recent experimental achievements and theoretical studies have generated substantial interest in the spherical tokamak concept. The ARIES-ST study was undertaken as a national U.S. effort to investigate the potential of the spherical tokamak concept as a fusion power plant and as a vehicle for fusion development. The 1000-MWe ARIES-ST power plant has an aspect ratio of 1.6, a major radius of 3.2 m, a plasma elongation (at 95% flux surface) of 3.4 and triangularity of 0.64. This configuration attains a β of 54% (which is 90% of the maximum theoretical β). While the plasma current is 31 MA, the almost perfect alignment of bootstrap and equilibrium current density profiles results in a current-drive power of only 31 MW. The on-axis toroidal field is 2.1 T and the peak field at the TF coil is 7.6 T, which leads to 288 MW of Joule losses in the normal-conducting TF system. The ARIES-ST study has highlighted many areas where tradeoffs among physics and engineering systems are critical in determining the optimum regime of operation for spherical tokamaks. Many critical issues also have been identified which must be resolved in R and D programs. (author)

  11. Charge-exchange and fusion reaction measurements during compression experiments with neutral beam heating in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.; Chan, A.A.; England, A.C.; Hendel, H.W.; Medley, S.S.; Nieschmidt, E.; Roquemore, A.L.; Scott, S.D.

    1986-04-01

    Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge-exchange neutral particle analyzer. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and /sup 3/He ions, respectively. When the plasma was compressed, the d(d,n)/sup 3/He fusion reaction rate increased a factor of five, and the /sup 3/He(d,p)/sup 4/He rate by a factor of twenty. These data were simulated with a bounce-averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling.

  12. Charge-exchange and fusion reaction measurements during compression experiments with neutral beam heating in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Kaita, R.; Heidbrink, W.W.; Hammett, G.W.

    1986-04-01

    Adiabatic toroidal compression experiments were performed in conjunction with high power neutral beam injection in the Tokamak Fusion Test Reactor (TFTR). Acceleration of beam ions to energies nearly twice the injection energy was measured with a charge-exchange neutral particle analyzer. Measurements were also made of 2.5 MeV neutrons and 15 MeV protons produced in fusion reactions between the deuterium beam ions and the thermal deuterium and 3 He ions, respectively. When the plasma was compressed, the d(d,n) 3 He fusion reaction rate increased a factor of five, and the 3 He(d,p) 4 He rate by a factor of twenty. These data were simulated with a bounce-averaged Fokker-Planck program, which assumed conservation of angular momentum and magnetic moment during compression. The results indicate that the beam ion acceleration was consistent with adiabatic scaling

  13. Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S.

    1995-10-18

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.

  14. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Rule, K.; Viola, M.; Williams, M.; Strykowsky, R.

    1999-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling

  15. Wall conditioning of fusion devices by reactive plasmas

    International Nuclear Information System (INIS)

    Winter, J.

    1988-06-01

    Conditioning of the plasma facing components of fusion devices like inner wall surfaces and limiters is a necessary prerequisite for the obtention of pure and stable fusion plasmas. Radiofrequency assisted dc-glow discharges in various gases (hydrogen, oxygen, methane) are a flexible tool to produce reactive species, which deplete the wall surfaces from reducible oxygen - and carbon compounds, or allow to plasmachemically deposite well defined thin layers onto them. The carbonization technique, i.e. the in situ deposition of amorphous hydrogenated carbon layers (a-C:H) and the resulting effects on tokamak plasmas are discussed in detail. Its application leads to significantly improved plasma performance. After its development and characterization at the Juelich tokamak TEXTOR it is now being used worldwide. (orig.)

  16. Tokamak Plasmas: Measurement of temperature fluctuations and ...

    Indian Academy of Sciences (India)

    ... Refresher Courses · Symposia · Live Streaming. Home; Journals; Pramana – Journal of Physics; Volume 55; Issue 5-6. Tokamak Plasmas : Measurement of temperature fluctuations and anomalous transport in the SINP tokamak. R Kumar S K Saha. Contributed Papers Volume 55 Issue 5-6 November-December 2000 pp ...

  17. First fusion proton measurements in TEXTOR plasmas using

    Czech Academy of Sciences Publication Activity Database

    Bonheure, G.; Mlynář, Jan; Van Wassenhove, G.; Hult, M.; González de Orduña, R.; Lutter, G.; Vermaercke, P.; Huber, A.; Schweer, B.; Esser, G.; Biel, W.

    2012-01-01

    Roč. 83, č. 10 (2012), 10D318 ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * fusion * activation * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://rsi.aip.org/resource/1/rsinak/v83/i10/p10D318_s1

  18. Plasma physics and controlled nuclear fusion research 1990. V. 1

    International Nuclear Information System (INIS)

    1991-01-01

    Volume 1 of the Proceedings of the Thirteenth International Conference on Plasma Physics and Controlled Nuclear Fusion Research contains papers given in two of the sessions: A and E. Session A contains the Artsimovich Memorial Lecture and papers on tokamaks; session E papers on plasma heating and current drive. The titles and authors of each paper are listed in the Contents. Abstracts accompany each paper. Refs, figs and tabs

  19. Transport Bifurcation in a Rotating Tokamak Plasma

    International Nuclear Information System (INIS)

    Highcock, E. G.; Barnes, M.; Schekochihin, A. A.; Parra, F. I.; Roach, C. M.; Cowley, S. C.

    2010-01-01

    The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.

  20. Plasma physics and controlled nuclear fusion research 1990. V. 3

    International Nuclear Information System (INIS)

    1991-01-01

    The thirteenth International Atomic Energy Agency Conference on Plasma Physics and Controlled Nuclear Fusion Research, held in Washington D.C., 1-6 October 1990, and organized in cooperation with the United States Department of Energy, was devoted to the exchange and dissemination of reports on the steady progress in the research on both inertial and magnetic confinement fusion, aiming ultimately for the production of commercial energy from controlled thermonuclear reactors. More than two hundred technical papers presented work on tokamak experiments, inertial confinement, non-tokamak confinement systems, magnetic confinement theory and modelling, plasma heating and current drive, the ITER project, technology and reactor concepts, and the economic, safety and environmental aspects of thermonuclear fusion. Refs, figs and tabs

  1. 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 ...

  2. 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 ...

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

    International Nuclear Information System (INIS)

    Martin, R.; Manero, F.

    1984-01-01

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

  4. ARIES-AT: An advanced tokamak, advanced technology fusion power plant

    International Nuclear Information System (INIS)

    Najmabadi, F.; Jardin, S.C.; Tillack, M.; Waganer, L.M.

    2001-01-01

    The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant. Several avenues were pursued in order to arrive at plasmas with a higher β and better bootstrap alignment compared to ARIES-RS that led to plasmas with higher β N and β. Advanced technologies that are examined in detail include: (1) Possible improvements to the overall system by using high-temperature superconductors, (2) Innovative SiC blankets that lead to a high thermal cycle efficiency of ∼60%; and (3) Advanced manufacturing techniques which aim at producing near-finished products directly from raw material, resulting in low-cost, and reliable components. The 1000-MWe ARIES-AT design has a major radius of 5.4 m, minor radius of 1.3 M, a toroidal β of 9.2% (β N =6.0) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current drive power is 24 MW. The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (5c/kWh), which is competitive with those projected for other sources of energy. (author)

  5. Deuterium-Tritium Simulations of the Enhanced Reversed Shear Mode in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D.R.; Manickam, J.; Scott, S.D.; Zarnstorff

    1997-04-01

    The potential performance, in deuterium-tritium plasmas, of a new enhanced con nement regime with reversed magnetic shear (ERS mode) is assessed. The equilibrium conditions for an ERS mode plasma are estimated by solving the plasma transport equations using the thermal and particle dif- fusivities measured in a short duration ERS mode discharge in the Tokamak Fusion Test Reactor [F. M. Levinton, et al., Phys. Rev. Letters, 75, 4417, (1995)]. The plasma performance depends strongly on Zeff and neutral beam penetration to the core. The steady state projections typically have a central electron density of {approx}2:5x10 20 m{sup -3} and nearly equal central electron and ion temperatures of {approx}10 keV. In time dependent simulations the peak fusion power, {approx} 25 MW, is twice the steady state level. Peak performance occurs during the density rise when the central ion temperature is close to the optimal value of {approx} 15 keV. The simulated pressure profiles can be stable to ideal MHD instabilities with toroidal mode number n = 1, 2, 3, 4 and {infinity} for {beta}{sub norm} up to 2.5; the simulations have {beta}{sub norm} {le} 2.1. The enhanced reversed shear mode may thus provide an opportunity to conduct alpha physics experiments in conditions imilar to those proposed for advanced tokamak reactors.

  6. Current drive in high density tokamak plasma

    International Nuclear Information System (INIS)

    Sakaguchi, Seiichiro; Jotaki, Eriko; Kawasaki, Shoji; Moriyama, Shin-ichi; Nagao, Akihiro; Nakamura, Kazuo; Nakamura, Yukio; Hiraki, Naoki; Itoh, Satoshi

    1989-01-01

    Current drive in high density tokamak plasma is investigated, with special attention given to mode conversion and proximity conditions that characterize the propagation of electromagnetic waves in the case of current drive by lower hybrid waves. A simple model is used to evaluate the current drive efficiency, and its dependence on various parameters associated with equipment is investigated to provide information required in designing experimental equipment. A strong troidal magnetic field is necessary to produce high density plasma, and incident electromagnetic waves should have a high frequency to prevent the mode conversion, suggesting that a high frequency and a strong troidal field are essential to permit desirable propagation of incident electromagnetic waves. The evaluation of the current drive efficiency shows that the proximity conditions and the power spectrum of the lower hybrid waves entering the plasma are of importance. The average refraction factor in the direction of the troidal field should be larger than but close to that determined from the proximity conditions in order to increase the drive efficiency. As the intensity of the troidal field increases, the refraction factor determined from the proximity conditions decreases, leading to an increase in the drive efficiency. (N.K.)

  7. Fundamental studies of fusion plasmas

    International Nuclear Information System (INIS)

    Aamodt, R.E.; Catto, P.J.; D'Ippolito, D.A.; Myra, J.R.; Russell, D.A.

    1993-01-01

    Work on ICRF interaction with the edge plasma is reported. ICRF generated convective cells have been established as an important mechanism for influencing edge transport and interaction with the H-mode, and for controlling profiles in the tokamak scrape-off-layer. Power dissipation by rf sheaths has been shown to be significant for some misaligned ICRF and IIBW antenna systems. Near-field antenna sheath work has been extended to the far-field case, important for experiments with low single pass absorption. Impurity modeling and Faraday screen design support has been provided for the ICRF community. In the area of core-ICRF physics, the kinetic theory of heating by applied ICRF waves has been extended to retain important geometrical effects relevant to modeling minority heated tokamak plasmas, thereby improving on the physics base that is standard in presently employed codes. Both the quasilinear theory of ion heating, and the plasma response function important in wave codes have been addressed. In separate studies, it has been shown that highly anisotropic minority heated plasmas can give rise to unstable field fluctuations in some situations. A completely separate series of studies have contributed to the understanding of tokamak confinement physics. Additionally, a diffraction formalism has been produced which will be used to access the focusability of lower hybrid, ECH, and gyrotron scattering antennas in dynamic plasma configurations

  8. TFTR/JET INTOR workshop on plasma transport tokamaks

    International Nuclear Information System (INIS)

    Singer, C.E.

    1985-01-01

    This report summarizes the proceedings of a Workshop on transport models for prediction and analysis of tokamak plasma confinement. Summaries of papers on theory, predictive modeling, and data analysis are included

  9. Study of heat and synchrotron radiation transport in fusion tokamak plasmas. Application to the modelling of steady state and fast burn termination scenarios for the international experimental fusion reactor ITER

    Energy Technology Data Exchange (ETDEWEB)

    Villar Colome, J. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Universitat Polytechnica de Catalunya (Spain)

    1997-12-01

    of fast burn termination scenarios in ITER is made by combining thermalhydraulic and fusion plasma modelling codes that implement the results obtained in the preceding sections. (author) 52 refs.

  10. Tokamak nonmaxwellian plasma dynamics in thermonuclear regime

    International Nuclear Information System (INIS)

    Cotsaftis, M.

    1987-01-01

    To reach ignition in a Tokamak plasma, large additional power P aux has to be injected in the device on top of the Joule heating P OH =VI r , V the plasma loop voltage, I r the resistive port of plasma current. Typi-cally JH ∼ 1 KeV, whereas ignition would requi- re IG ∼ 7-10 KeV. To gain this factor 7, one at least should inject additional power P aux ∼ 7P OH , supposing that nothing, especially the heat transport, is modified. This is by far not the case, with the so-called energy lifetime degradation, largely observed in oil experiments (but less dramatic with divertors), where energy lifetime tau E behaves like P tot -b with b∼1/2. In large machines where ignition temperature is the target to be imperiously reached, this implies to inject a very large power, typically P aux ∼ 50 to 100 MW, depending on size and parameters and on actual transport. So it is of importance with such figures, or even larger ones owing to uncertain ties, to optimize at best injected power by increasing its efficiency, both with respect to possible transport laws, and to physical phenomena governing heat flow in the system from the sources. This leads to the concept of scenarios, as time sequences of power input, where physical properties of the plasma system are used to build up ion temperature so that ignition is reached with minimum P tot = P OH + P aux and with fixed Q = Q o > 1. Elements for this study are given. The method is outlined. The resulting system of equations describing the evolution of a thermonuclear plasma is given

  11. STAR Power, an Interactive Educational Fusion CD with a Dynamic, Shaped Tokamak Power Plant Simulator

    Science.gov (United States)

    Leuer, J. A.; Lee, R. L.; Kellman, A. G.; Chapman Nutt, G. C., Jr.; Holley, G.; Larsen, T. A.

    2000-10-01

    We describe an interactive, educational fusion adventure game developed within our fusion education program. The theme of the adventure is start-up of a state-of-the-art fusion power plant. To gain access to the power plant control room, the student must complete several education modules, including topics on building an atom, fusion reactions, charged particle motion in electric and magnetic fields, and building a power plant. Review questions, a fusion video, library material and glossary provide additional resources. In the control room the student must start-up a complex, dynamic fusion power plant. The simulation model contains primary elements of a tokamak based device, including a magnetic shaper capable of producing limited and diverted elongated plasmas. A zero dimensional plasma model based on ITER scaling and containing rate based conservation equations provides dynamic feedback through major control parameters such as toroidal field, fueling rate and heating. The game is available for use on PC and Mac. computers. Copies will be available at the conference.

  12. A control approach for plasma density in tokamak machines

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  13. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  14. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

    2016-01-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO 2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO 2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO 2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO 2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO 2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  15. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  16. Minerals resource implications of a tokamak fusion reactor economy

    International Nuclear Information System (INIS)

    Cameron, E.; Conn, R.W.; Kulcinski, G.L.; Sviatoslavsky, I.

    1979-09-01

    The mineral resource implications of an economy of tokamak-type fusion reactors are assessed based upon the recent conceptual reactor design study, NUWMAK, developed at the University of Wisconsin. For comparative purposes, various structural alloys of vanadium and steel are assumed to be usable in the NUWMAK design in place of the titanium alloy originally selected. In addition, the inner blanket core and magnet system of the conceptual reactor, HFCTR, developed at the Massachusetts Institute of Technology, are assumed to be interchangeable with the comparable components in NUWMAK. These variations permit a range of likely requirements to be assessed

  17. TRAIL: a Tokamak Rail Gun Limiter for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Powell, J R; Yu, W S; Fillo, J A; Usher, J L

    1980-01-01

    An attractive new limiter concept is investigated. The Tokamak Rail Gun Limiter (TRAIL) system impacts a stream of moderate velocity pellets (100 to 200 m/sec through the plasma edge region to absorb energy and define the plasma boundary. The pellets are recycled, after cooling, to the injector in an E-M mass accelerator. Heat fluxes of approx. 30,000 W/cm/sup 2/ can be readily accommodated by the pellets, with very low recirculating power requirements (approx. 0.1%) for the accelerator. The mass accelerator velocity requirements are well within the present state-of-the-art (several km/sec). Accelerators injecting pellets at approx. 1 km/sec can be used to control local plasma temperature and current profiles and to act as energy absorbers to shut down the plasma without damage to the first wall if a plasma disruption occurs.

  18. The ARIES-AT advanced tokamak, Advanced technology fusion power plant

    International Nuclear Information System (INIS)

    Najmabadi, Farrokh; Abdou, A.; Bromberg, L.

    2006-01-01

    The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R and D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ x = 2.2) which is the result of a 'thinner' blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β N . ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb-17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 deg. C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb-17Li to about 1000 deg. C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES

  19. Reactor aspects of counterstreaming-ion tokamak plasmas

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1975-06-01

    Toroidal DT plasmas in which the D and T ions make up two distinct, quasi-thermal velocity distributions, oppositely displaced in velocity along the magnetic axis, are discussed. Such counterstreaming distributions can be set up by introducing all ions by tangential injection of neutral beams, and by removing ions from the plasma shortly after they have decelerated to an energy approximate to or less than 2T/sub e/ by Coulomb drag on the plasma electrons. A simple physical model for counterstreaming-ion operation is postulated, which allows one to deduce the ion velocity distributions and required energy and particle confinement times that are in good agreement with the results of previous Fokker-Planck calculations. The variations of fusion reactivity, power gain, and power density with injection energy and electron temperature are presented. The practical problems of implementing counter-streaming operation in a tokamak, such as charge-exchange losses, the prompt removal of cold ions, and the effect of impurities are discussed. (U.S.)

  20. Plasma physics and controlled nuclear fusion research 1992

    International Nuclear Information System (INIS)

    1993-01-01

    The 14th International Atomic Energy Agency Conference on Plasma Physics and Controlled Fusion Research was held in Wuerzburg, Germany, 30 September - 7 October 1992. Its proceedings were published in four volumes. Volume 1 contains major papers on tokamak experiments, the text of the Artsimovich Memorial Lecture held by Prof. P.K. Kaw, and papers on plasma heating and current drive. Volume 2 contains papers on magnetic confinement theory and modelling and on non-tokamak confinement systems. Volume 3 contains papers on inertial confinement, the ITER project, and technology and reactor concepts. Volume 4 contains summaries on (i) tokamak experiments, heating and current drive, (ii) alternative magnetic systems, (iii) magnetic confinement theory, (iv) inertial confinement, and (v) technology and reactor concepts. Refs, figs and tabs

  1. Plasma physics and controlled nuclear fusion research 1988. V.2

    International Nuclear Information System (INIS)

    1989-01-01

    Volume 2 of the Proceedings of the Twelfth International Conference on Plasma Physics and Controlled Nuclear Fusion contains papers given in two of the sessions: D and C. Session D contains papers on magnetic confinement theory and modelling tokamaks, and session C the papers on non-tokamak confinement system. Each of these papers and their authors is listed in the Contents; in turn, each paper contains an abstract for more information as to the contents of a specific paper. Refs, figs and tabs

  2. Measurements of plasma position in TJ-I Tokamak

    International Nuclear Information System (INIS)

    Quin, J.; Ascasibar, E.; Navarro, A. P.; Ochando, M. A.; Pastor, I.; Pedrosa, M. A.; Rodriguez, L.; Sanchez, J.

    1994-01-01

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

  3. JOINT EXPERIMENTS ON SMALL TOKAMAKS: EDGE PLASMA STUDIES ON CASTOR

    Czech Academy of Sciences Publication Activity Database

    Van Oost, G.; Berta, M.; Brotánková, Jana; Dejarnac, Renaud; Del Bosco, E.; Dufková, Edita; Ďuran, Ivan; Gryaznevich, M.P.; Horáček, Jan; Hron, Martin; Malaquias, A.; Mank, G.; Peleman, P.; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zoletnik, S.; Tál, B.; Ferrera, J.; Fonseca, A.; Hegazy, H.; Kuznetsov, Y.; Ossyannikov, A.; Singh, A.; Sokholov, M.; Talebitaher, A.

    2007-01-01

    Roč. 47, č. 5 (2007), s. 378-386 ISSN 0029-5515 R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * edge plasma * turbulence * Langmuir probe * plasma radiation * Hall probe Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.278, year: 2007

  4. Plasma recycling and marfe phenomenon in a tokamak

    International Nuclear Information System (INIS)

    Nedospasov, A.V.; Tokar', M.Z.; Vasin, N.L.

    1988-01-01

    The mechanism of Marfe formation as a result of the plasma transition to the state with strong recycling on the liner is investigated and calculated for T-10 tokamak conditions. The self-sustained oscillations of the peripheral plasma between the states with strong and weak recycling, explaining fluctuations of the plasma parameters in the discharges with Marfe, are studied

  5. Plan for 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) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D ampersand D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D ampersand D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D ampersand D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D ampersand D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates

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

    International Nuclear Information System (INIS)

    Stott, P. E.; Sanchez, J.

    1994-01-01

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

  7. Plasma physics and fusion plasma electrodynamics

    CERN Document Server

    Bers, Abraham

    2016-01-01

    Plasma is a ubiquitous state of matter at high temperatures. The electrodynamics of plasmas encompasses a large number of applications, from understanding plasmas in space and the stars, to their use in processing semiconductors, and their role in controlled energy generation by nuclear fusion. This book covers collective and single particle dynamics of plasmas for fully ionized as well as partially ionized plasmas. Many aspects of plasma physics in current fusion energy generation research are addressed both in magnetic and inertial confinement plasmas. Linear and nonlinear dynamics in hydrodynamic and kinetic descriptions are offered, making both simple and complex aspects of the subject available in nearly every chapter. The approach of dividing the basic aspects of plasma physics as "linear, hydrodynamic descriptions" to be covered first because they are "easier", and postponing the "nonlinear and kinetic descriptions" for later because they are "difficult" is abandoned in this book. For teaching purpose...

  8. A fusion power plant without plasma-material interactions

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, S.A.

    1997-04-01

    A steady-state fusion power plant is described which avoids the deleterious plasma-material interactions found in D-T fueled tokamaks. It is based on driven p-{sup 11}B fusion in a high-beta closed-field device, the field-reversed configuration (FRC), anchored in a gas-dynamic trap (GDT). The plasma outflow on the open magnetic-field lines is cooled by radiation in the GDT, then channeled through a magnetic nozzle, promoting 3-body recombination in the expansion region. The resulting supersonic neutral exhaust stream flows through a turbine, generating electricity.

  9. 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

  10. Feasibility study of a fission-suppressed tokamak fusion breeder

    International Nuclear Information System (INIS)

    Moir, R.W.; Lee, J.D.; Neef, W.S.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m 2 and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of 233 U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U 3 O 8 depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. Control of plasma position in the CASTOR tokamak

    International Nuclear Information System (INIS)

    Valovic, M.

    1988-11-01

    A simple servo-system designed for plasma position control in the CASTOR tokamak is described. Both radial and vertical plasma displacements were minimized using two servo-loops consisting of detection coils, a conventional electric controller and an amplifier operated as an unipolar voltage-controlled current source. To ensure the optimum conditions in the start-up phase of the discharge, currents in the servo-systems were externally preprogrammed. The prescribed plasma position was maintained with the accuracy of 3 mm. The feedback control improves plasma parameters, e.g. it removes the positional disruption at the end of the tokamak discharge. (J.U.). 4 figs., 3 refs

  13. [Fusion research/tokamak]. Final report, 1 May 1988 - 30 April 1994

    International Nuclear Information System (INIS)

    1994-01-01

    The objectives of the Fusion Research Center Program are: (1) to advance /the transport studies of tokamaks, including the development and maintenance of the Magnetic Fusion Energy Database, and (2) to provide theoretical interpretation, modeling and equilibrium and stability studies for the text-upgrade tokamak. Work is described on five basic categories: (1) magnetic fusion energy database; (2) computational support and numerical modeling; (3) support for TEXT-upgrade and diagnostics; (4) transport studies; and (5) Alfven waves

  14. Hydrogen isotope exchange and conditioning in graphite limiters used in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    LaMarche, P.H.; Dylla, H.F.; McCarthy, P.J.; Ulrickson, M.

    1986-01-01

    Isotopic exchange experiments performed in the Tokamak Fusion Test Reactor (TFTR) are used to examine the outgassing and diffusive properties of graphite used as the plasma limiter. Changeover from hydrogen to deuterium for different periods ranges from approx.600 to 60 plasma discharges, which appears to be correlated in the limiter temperature. We present a simple analytical model that predicts a fast transient (approx.10 plasma discharges) changeover, where the deuterium fueling dilutes the adsorbed and near surface hydrogen, and a slowly changing term where bulk hydrogen diffuses to the surface. Using this model we can extract an activation energy for diffusion of 0.15 +- 0.02 eV. We hypothesize that interpore diffusion for this porous (approx.15%) material is consistent with our observations

  15. Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design

    International Nuclear Information System (INIS)

    Qiu Qing-Lai; Xiao Bing-Jia; Guo Yong; Liu Lei; Wang Yue-Hang

    2017-01-01

    Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other. (paper)

  16. Ionization balance in EBIT and tokamak plasmas

    Science.gov (United States)

    Peacock, N. J.; Barnsley, R.; O'Mullane, M. G.; Tarbutt, M. R.; Crosby, D.; Silver, J. D.; Rainnie, J. A.

    2001-01-01

    The equilibrium state in tokamak core plasmas has been studied using the relative intensities of resonance x-ray lines, for example Lyα (H-like), "w" (He-like), and "q" (Li-like) from test ions such as Ar+15, Ar+16, and Ar+17. A full spatial analysis involves comparison of the line intensities with ion diffusion calculations, including relevant atomic rates. A zero-dimensional model using a global ion loss rate approximation has also been demonstrated by comparison with the data collected from a Johann configuration spectrometer with a charged coupled device (CCD) detector. Since the lines are nearly monoenergetic, their intensities are independent of the instrument sensitivity and are directly proportional to the ion abundances. This method has recently been applied to Ar in the Oxford electron beam ion trap (EBIT) with a beam energy in the range 3-10 keV. Taking into account the cross sections for monoenergetic electron collisions and polarization effects, model calculations agree with the observed line ratios at 4.1 keV beam energy. This work will be expanded to provide nomograms of ionization state versus line intensity ratios as a function of EBIT beam energy.

  17. Current drive in spherical tokamak plasmas

    International Nuclear Information System (INIS)

    Storer, R.

    1999-01-01

    The early experiments on a spherical rotamak showed that a rotating magnetic field could be used to drive substantial currents and create a compact torus magnetic field configuration. The theoretical analysis of the spherical rotamak has been essentially confined to this class. Recent experiments on the Flinders Rotamak-ST have included a toroidal field, produced by a current-carrying central rod, with encouraging results; for it has been shown that an enhanced current can be driven with this configuration which is the equivalent of a spherical tokamak. This paper will be devoted to a theoretical and computational analysis of this situation. We use a model where the rotating magnetic field is applied to a spherical plasma, with the rotating field oriented parallel to the equatorial plane, taken to be the x-y plane. In our model the ions form a uniform background and the frequency of the rotating Held is very much less than the electron cyclotron frequency (with respect to the rotating field strength) and very much greater than the ion cyclotron frequency. This condition is satisfied by the rotamak experiments

  18. Controlled fusion and plasma heating

    International Nuclear Information System (INIS)

    1990-06-01

    The contributions presented in the 17th European Conference on Controlled Fusion and Plasma Heating were focused on Tore Supra investigations. The following subjects were presented: ohmic discharges, lower hybrid experiments, runaway electrons, Thomson scattering, plasma density measurements, magnetic fluctuations, polarization scattering, plasma currents, plasma fluctuation measurements, evaporation of hydrogen pellets in presence of fast electrons, ripple induced stochastic diffusion of trapped particles, tearing mode stabilization, edge effects on turbulence behavior, electron cyclotron heating, micro-tearing modes, divertors, limiters

  19. Two-dimensional Simulations of Correlation Reflectometry in Fusion Plasmas

    International Nuclear Information System (INIS)

    Valeo, E.J.; Kramer, G.J.; Nazikian, R.

    2001-01-01

    A two-dimensional wave propagation code, developed specifically to simulate correlation reflectometry in large-scale fusion plasmas is described. The code makes use of separate computational methods in the vacuum, underdense and reflection regions of the plasma in order to obtain the high computational efficiency necessary for correlation analysis. Simulations of Tokamak Fusion Test Reactor (TFTR) plasma with internal transport barriers are presented and compared with one-dimensional full-wave simulations. It is shown that the two-dimensional simulations are remarkably similar to the results of the one-dimensional full-wave analysis for a wide range of turbulent correlation lengths. Implications for the interpretation of correlation reflectometer measurements in fusion plasma are discussed

  20. Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

    International Nuclear Information System (INIS)

    Giunta, A.S.; Henderson, S.; O'Mullane, M.; Summers, H.P.; Harrison, J.; Doyle, J.G.

    2016-01-01

    This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.

  1. Calculations of alpha particle loss for reversed magnetic shear in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Redi, M.H.; White, R.B.; Batha, S.H.; Levinton, F.M.; McCune, D.C.

    1997-03-01

    Hamiltonian coordinate, guiding center code calculations of the toroidal field ripple loss of alpha particles from a reversed shear plasma predict both total alpha losses and ripple diffusion losses to be greater than those from a comparable non-reversed magnetic shear plasma in the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. High central q is found to increase alpha ripple losses as well as first orbit losses of alphas in the reversed shear simulations. A simple ripple loss model, benchmarked against the guiding center code, is found to work satisfactorily in transport analysis modelling of reversed and monotonic shear scenarios. Alpha ripple transport on TFTR affects ions within r/a=0.5, not at the plasma edge. The entire plasma is above threshold for stochastic ripple loss of alpha particles at birth energy in the reversed shear case simulated, so that all trapped 3.5 MeV alphas are lost stochastically or through prompt losses. The 40% alpha particle loss predictions for TFTR suggest that reduction of toroidal field ripple will be a critical issue in the design of a reversed shear fusion reactor.

  2. Plasma diagnostics at Aditya Tokamak by two views visible light tomography

    International Nuclear Information System (INIS)

    Goswami, Mayank; Munshi, Prabhat; Saxena, Anupam; Kumar, Manoj; Kumar, Ajai

    2014-01-01

    Graphical abstract: - Highlights: • Improved algorithm works equally well for central as well as for peripherical plasma regions. • Entropy optimized smoothening parameters eliminate user dependencies. • Real time fusion grade plasma diagnostics images. - Abstract: This visible light computerized tomography exercise is a part of a project to establish an auxiliary imaging method to assist other imaging facilities at the Institute of Plasma Research (IPR), India. Space constraints around Aditya Tokamak allow only two orthogonal ports. Each port has one detector array (64 sensors) sensitive to the visual spectrum emitted by H α emission. The objective here is to report the developments on limited view tomography for hot plasma imaging. Spatially filtered entropy maximization algorithm with non-uniform discretization grids is employed. Estimation of unique kernel smoothening parameters (mask size and exponent factor) depends on entropy function and projection data. It removes requirement of any arbitrary/user-based decision for choosing a regularization factor thus minimizes the chance for biasedness or errors. Synthetic projection data is used to analyse the performance of this modification. The error band in the process of recovery remains under acceptable level (less than 15%) irrespective of the origin of the emissions from the core. Reconstructed hot plasma images/profiles from Aditya Tokamak are shown. These profiles may improve the current understanding about (a) plasma–wall interaction or edge plasma turbulence, (b) control and generation of plasma and (c) correlations between theoretical and engineering advancements in Tokamak reactors

  3. Fast wave current drive in reactor scale tokamak plasmas

    International Nuclear Information System (INIS)

    Becoulet, A.; Moreau, D.; Saoutic, B.

    1991-01-01

    The possibility for driving current in large tokamak plasmas using the fast magnetosonic wave is analysed in terms of linear propagation-absorption, and also in terms of quasilinear absorption through an hamiltonian analysis of the wave-particle interaction. The tokamak geometry is shown to strongly influence the capability for the fast wave to sustain a significant part of the toroidal current. Synergetic effects with other scenarios are also discussed

  4. Interaction of candidate plasma facing materials with tokamak plasma in COMPASS

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Macková, Anna; Malinský, Petr; Havránek, Vladimír; Naydenkova, Diana; Klevarová, Veronika; Petersson, P.; Gasior, P.; Hakola, A.; Rubel, M.; Fortuna, E.; Kolehmainen, J.; Tervakangas, S.

    2017-01-01

    Roč. 493, September (2017), s. 102-119 ISSN 0022-3115. [International Conference on Plasma-Facing Materials and Components for Fusion Applications/15./. Aix-en-Provence, 18.05.2015-22.05.2015] R&D Projects: GA ČR(CZ) GA14-12837S; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045; GA MŠk LM2015056 Institutional support: RVO:61389021 ; RVO:61389005 Keywords : erosion * COMPASS tokamak * plasma-material interaction * ion beam analysis Subject RIV: JF - Nuclear Energetics; JF - Nuclear Energetics (UJF-V) OBOR OECD: Nuclear related engineering; Nuclear related engineering (UJF-V) Impact factor: 2.048, year: 2016 http://www.sciencedirect.com/science/article/pii/S0022311517301708

  5. Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

    Science.gov (United States)

    Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; Grierson, B. A.; Staebler, G. M.; Rice, J. E.; Yuan, X.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Sciortino, F.

    2018-02-01

    A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. This Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and time scales of cold-pulse experiments in tokamak plasmas.

  6. MHD equilibrium properties of tokamak fusion reactor designs

    International Nuclear Information System (INIS)

    Todd, A.M.M.; Gralnick, S.L.; Dalhed, H.E.

    1976-01-01

    The equilibrium properties of several Tokamak Reactor Designs are analyzed and compared for varying pressure and current profiles using the Princeton Equilibrium Code. It is found that the UWMAK configuration has a broader range of equilibria than the Princeton Reference Design configuration, but that the safety factor on axis is less than unity for peaked current distributions. The Argonne Experimental Power Reactor has a satisfactory range of equilibria, but a means of limiting or diverting the plasma has not yet been proposed, and this may substantially change the results obtained

  7. Scaling for scrape-off layer plasma in tokamak

    International Nuclear Information System (INIS)

    Shimomura, Yasuo; Maeda, Hikosuke; Kimura, Haruyuki; Azumi, Masashi; Odajima, Kazuo

    1977-12-01

    Scaling for a scrape-off layer plasma in a tokamak is obtained by using DIVA (JFT-2a). The scaling gives the average electron temperature, the width and the mean electron density of the scrape-off layer. The temperature at the edge will be high in a future large tokamak with a small energy-loss by charge-exchange and radiation. The scrape-off layer plasma can easily shield the impurity influx from the wall. The fuel, however, can easily penetrate into the main plasma. (auth.)

  8. Reconstruction of plasma current profile of Tokamaks using genetic algorithm

    International Nuclear Information System (INIS)

    Kishimoto, Maki

    1996-01-01

    A new method to reconstruct plasma shape and plasma current distribution from magnetic measurements using a combinatorial optimization technique are proposed. The reconstruction of plasma current profile from magnetic measurements is regarded as an optimum allocation problem of currents into cross section of the vacuum vessel of the Tokamak. In order to solve this optimization problem, we use a genetic algorithm. The effectiveness of this method is shown by the application of this technique to JT-60U plasmas. (author)

  9. Internal magnetic field measurement in tokamak plasmas using a ...

    Indian Academy of Sciences (India)

    Abstract. In a tokamak plasma, the poloidal magnetic field profile closely depends on the current density profile. We can deduce the internal magnetic field from the analysis of circular polarization of the spectral lines emitted by the plasma. The theory of the measurement and a detailed design of the Zeeman polarimeter ...

  10. The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR

    International Nuclear Information System (INIS)

    Tammen, H.F.

    1995-01-01

    One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics 'Rijnhuizen', was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL)

  11. The ion velocity distribution of tokamak plasmas: Rutherford scattering at TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Tammen, H.F.

    1995-01-10

    One of the most promising ways to gererate electricity in the next century on a large scale is nuclear fusion. In this process two light nuclei fuse and create a new nucleus with a smaller mass than the total mass of the original nuclei, the mass deficit is released in the form of kinetic energy. Research into this field has already been carried out for some decades now, and will have to continue for several more decades before a commercially viable fusion reactor can be build. In order to obtain fusion, fuels of extremely high temperatures are needed to overcome the repulsive force of the nuclei involved. Under these circumstances the fuel is fully ionized: it consists of ions and electrons and is in the plasma state. The problem of confining such a hot substance is solved by using strong magnetic fields. One specific magnetic configuration, in common use, is called the tokamak. The plasma in this machine has a toroidal, i.e. doughnut shaped, configuration. For understanding the physical processes which take place in the plasma, a good temporally and spatially resolved knowledge of both the ion and electron velocity distribution is required. The situation concerning the electrons is favourable, but this is not the case for the ions. To improve the existing knowledge of the ion velocity distribution in tokamak plasmas, a Rutherford scattering diagnostic (RUSC), designed and built by the FOM-Institute for Plasmaphysics `Rijnhuizen`, was installed at the TEXTOR tokamak in Juelich (D). The principle of the diagnostic is as follows. A beam of monoenergetic particles (30 keV, He) is injected vertically into the plasma. A small part of these particles collides elastically with the moving plasma ions. By determining the energy of a scattered beam particle under a certain angle (7 ), the initial velocity of the plasma ion in one direction can be computed. (orig./WL).

  12. Simulation science for fusion plasmas

    International Nuclear Information System (INIS)

    Sudo, S; Skoric, M M; Watanabe, T-H; Todo, Y; Miura, H; Ishizaki, R; Ito, A; Ohtani, H; Ishiguro, S; Tomita, Y; Horiuchi, R; Ishizawa, A; Usami, S; Nakamura, H; Ito, Atsushi; Takayama, A; Sato, M; Yamamoto, T; Den, M; Sakagami, H

    2008-01-01

    The world fusion effort has embarked into a new age with the construction of ITER in Cadarache, France, which will be the first magnetic confinement fusion plasma experiment dominated by the self-heating of fusion reactions. In order to operate and control burning plasmas and next generation demo fusion reactors, an advanced capability for comprehensive integrated computer simulations that are fully verified and validated against experimental data will be necessary. The ultimate goal is to predict reliably the behaviour of plasmas in toroidal magnetic confinement devices on all relevant scales, both in time and space. In addition to developing a sophisticated integrated simulation codes, directed advanced research in fusion physics, applied mathematics, computer science and software is envisaged. In this paper we review the basic strategy and main research efforts at the Department of Simulation Science of the National Institute for Fusion Science (NIFS)- which is the Inter University Institute and the coordinating Center of Excellence for academic fusion research in Japan. We overview a simulation research at NIFS, in particular relation to experiments in the Large Helical Device (LHD), the world's largest superconducting heliotron device, as a National Users' facility (see Motojima et al.). Our main goal is understanding and systemizing the rich hierarchy of physical mechanisms in fusion plasmas, supported by exploring a basic science of complexity of plasma as a highly nonlinear, non-equilibrium, open system. The aim is to establish a simulation science as a new interdisciplinary field by fostering collaborative research in utilizing the large-scale supercomputer simulators. A concept of the hierarchy-renormalized simulation modelling will be invoked en route toward the LHD numerical test reactor.

  13. High-Gain High-Field Fusion Plasma

    Science.gov (United States)

    Li, Ge

    2015-01-01

    A Faraday wheel (FW)—an electric generator of constant electrical polarity that produces huge currents—could be implemented in an existing tokamak to study high-gain high-field (HGHF) fusion plasma, such as the Experimental Advanced Superconducting Tokamak (EAST). HGHF plasma can be realized in EAST by updating its pulsed-power system to compress plasma in two steps by induction fields; high gains of the Lawson trinity parameter and fusion power are both predicted by formulating the HGHF plasma. Both gain rates are faster than the decrease rate of the plasma volume. The formulation is checked by earlier ATC tests. Good agreement between theory and tests indicates that scaling to over 10 T at EAST may be possible by two-step compressions with a compression ratio of the minor radius of up to 3. These results point to a quick new path of fusion plasma study, i.e., simulating the Sun by EAST. PMID:26507314

  14. Tokamak fusion reactors with less than full tritium breeding

    International Nuclear Information System (INIS)

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed

  15. STARFIRE: a commercial tokamak fusion power plant study

    International Nuclear Information System (INIS)

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations

  16. Systems study of tokamak fusion--fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations.

  17. Systems study of tokamak fusion--fission reactors

    International Nuclear Information System (INIS)

    Tenney, F.H.; Bathke, C.G.; Price, W.G. Jr.; Bohlke, W.H.; Mills, R.G.; Johnson, E.F.; Todd, A.M.M.; Buchanan, C.H.; Gralnick, S.L.

    1978-11-01

    This publication reports the results of a two to three year effort at a systematic analysis of a wide variety of tokamak-driven fissioning blanket reactors, i.e., fusion--fission hybrids. It addresses the quantitative problems of determining the economically most desirable mix of the two products: electric power and fissionable fuel and shows how the price of electric power can be minimized when subject to a variety of constraints. An attempt has been made to avoid restricting assumptions, and the result is an optimizing algorithm that operates in a six-dimensional parameter space. Comparisons are made on sets of as many as 100,000 distinct machine models, and the principal results of the study have been derived from the examination of several hundred thousand possible reactor configurations

  18. STARFIRE: a commercial tokamak fusion power plant study

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

  19. High density plasma heating in the Tokamak à configuration variable

    International Nuclear Information System (INIS)

    Curchod, L.

    2011-04-01

    The Tokamak à Configuration Variable (TCV) is a medium size magnetic confinement thermonuclear fusion experiment designed for the study of the plasma performances as a function of its shape. It is equipped with a high power and highly flexible electron cyclotron heating (ECH) and current drive (ECCD) system. Up to 3 MW of 2 nd harmonic EC power in ordinary (O 2 ) or extraordinary (X 2 ) polarization can be injected from TCV low-field side via six independently steerable launchers. In addition, up to 1.5 MW of 3 rd harmonic EC power (X 3 ) can be launched along the EC resonance from the top of TCV vacuum vessel. At high density, standard ECH and ECCD are prevented by the appearance of a cutoff layer screening the access to the EC resonance at the plasma center. As a consequence, less than 50% of TCV density operational domain is accessible to X 2 and X 3 ECH. The electron Bernstein waves (EBW) have been proposed to overcome this limitation. EBW is an electrostatic mode propagating beyond the plasma cutoff without upper density limit. Since it cannot propagate in vacuum, it has to be excited by mode conversion of EC waves in the plasma. Efficient electron Bernstein waves heating (EBH) and current drive (EBCD) were previously performed in several fusion devices, in particular in the W7-AS stellarator and in the MAST spherical tokamak. In TCV, the conditions for an efficient O-X-B mode conversion (i.e. a steep density gradient at the O 2 plasma cutoff) are met at the edge of high confinement (H-mode) plasmas characterized by the appearance of a pedestal in the electron temperature and density profiles. TCV experiments have demonstrated the first EBW coupling to overdense plasmas in a medium aspect-ratio tokamak via O-X-B mode conversion. This thesis work focuses on several aspects of ECH and EBH in low and high density plasmas. Firstly, the experimental optimum angles for the O-X-B mode conversion is successfully compared to the full-wave mode conversion calculation

  20. Ion cyclotron emission from fusion plasmas

    Science.gov (United States)

    Cauffman, Stephen Richard

    1997-11-01

    Ion cyclotron emission (ICE), characterized by a spectrum of discrete harmonics of the ion cyclotron frequency evaluated at a particular magnetic field strength, is commonly observed in fusion plasma devices. In the Joint European Torus (JET), ICE power varies linearly with the fusion alpha particle population in the plasma edge, suggesting that ICE may serve as a useful alpha particle diagnostic. Here, we use a system of wire loop probes to monitor the behavior of ICE in the Tokamak Fusion Test Reactor (TFTR). ICE in TFTR exhibits more complicated behavior than in JET. The harmonic structure varies with plasma species, and typical discharges exhibit two distinct classes of emission-one occurring transiently after the onset of neutral beam injection (NBI), and the other rising more slowly and persisting throughout NBI. Correlations with the fusion rate are observed, but the dependence is less straightforward than in JET. To understand the differences between TFTR and JET observations, we review the theory of the magnetoacoustic cyclotron instability (MCI). We then consider the theory of compressional Alfven eigenmodes (CAEs), which describes the same physics using a global approach. A numerical investigation of CAE behavior demonstrates that in TFTR, ICE behavior should depend sensitively on many quantities, including the Alfven speed near the plasma edge and the shape of the fusion product velocity distribution, while in JET, ICE behavior should depend primarily on the size of the alpha particle population. Dissimilar observations of ICE in TFTR and JET are thus consistent with a single theory, which predicts that the behavior of ICE will depend strongly on the ratio of the alpha particle birth velocity to the local Alfven speed. When alpha particles are super-Alfvenic near the plasma edge, ICE power will vary primarily with the alpha particle density, and will thus provide a simple method of monitoring this population. However, if the alpha particles are sub

  1. Development of large insulator rings for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Brown, T.; Tobin, A.

    1978-01-01

    This paper discusses research and development leading to the manufacture of large ceramic insulator rings for the TFTR (TOKAMAK Fusion Test Reactor). Material applications, fabrication approach and testing activities are highlighted

  2. Proceedings of 1995 the first Taedok international fusion symposium on advanced tokamak researches

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Lee, K. W.; Hwang, C. K.; Hong, B. G.; Hong, G. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-05-01

    This proceeding is from the First Taeduk International Fusion Symposium on advanced tokamak research, which was held at Korea Atomic Energy Research Institute, Taeduk Science Town, Korea on March 28-29, 1995. (Author) .new.

  3. Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST

    International Nuclear Information System (INIS)

    Xu, X Q

    2007-01-01

    We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. With our 4D (ψ, θ, ε, μ) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices

  4. Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

    International Nuclear Information System (INIS)

    Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

    1995-01-01

    The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

  5. Final Report: Spectral Analysis of L-shell Data in the Extreme Ultraviolet from Tokamak Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Lepson, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jernigan, J. Garrett [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Beiersdorfer, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-02-05

    We performed detailed analyses of extreme ultraviolet spectra taken by Lawrence Livermore National Laboratory on the National Spherical Torus Experiment at Princeton Plasma Physics Laboratory and on the Alcator CKmod tokamak at the M.I.T. Plasma Science and Fusion Center. We focused on the emission of iron, carbon, and other elements in several spectral band pass regions covered by the Atmospheric Imaging Assembly on the Solar Dynamics Observatory. We documented emission lines of carbon not found in currently used solar databases and demonstrated that this emission was due to charge exchange.

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

    Czech Academy of Sciences Publication Activity Database

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

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

  7. Kinetic modelling of runaway electron avalanches in tokamak plasmas.

    Czech Academy of Sciences Publication Activity Database

    Nilsson, E.; Decker, J.; Peysson, Y.; Granetz, R.S.; Saint-Laurent, F.; Vlainic, Milos

    2015-01-01

    Roč. 57, č. 9 (2015), č. článku 095006. ISSN 0741-3335 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : plasma physics * runaway electrons * knock-on collisions * tokamak * Fokker-Planck * runaway avalanches Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.404, year: 2015

  8. Negative edge plasma currents in the SINP tokamak

    Indian Academy of Sciences (India)

    conditions. In the ohmic discharges of the SINP tokamak, we observed 'negative edge plasma currents' at the edge region (r/a ∼ 0.7–0.8, where a = 7.5 cm is the ... The horizontal plasma column (figure 1c) is observed to move inwards/ ... (b) 3 by 3 NaI(Tl) output (HX) in MeV, (c) horizontal plasma column motion ( hor).

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

    Czech Academy of Sciences Publication Activity Database

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

    2006-01-01

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

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

    Czech Academy of Sciences Publication Activity Database

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

    2004-01-01

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

  11. 2001 activity report of the development and research line in controlled thermonuclear fusion of the Plasma Associated Laboratory

    International Nuclear Information System (INIS)

    Ludwig, Gerson Otto

    2002-01-01

    The year 2001 activities of the controlled thermonuclear fusion research line of the Plasma Associated Laboratory at the National Institute for Space Research - Brazil are reported. The report approaches the staff, participation in congresses, goals for the year 2002 and papers on Tokamak plasmas, plasma diagnostic, bootstraps, plasma equilibrium and diagnostic

  12. The Numerical Tokamak Project (NTP) simulation of turbulent transport in the core plasma: A grand challenge in plasma physics

    International Nuclear Information System (INIS)

    1993-12-01

    The long-range goal of the Numerical Tokamak Project (NTP) is the reliable prediction of tokamak performance using physics-based numerical tools describing tokamak physics. The NTP is accomplishing the development of the most advanced particle and extended fluid model's on massively parallel processing (MPP) environments as part of a multi-institutional, multi-disciplinary numerical study of tokamak core fluctuations. The NTP is a continuing focus of the Office of Fusion Energy's theory and computation program. Near-term HPCC work concentrates on developing a predictive numerical description of the core plasma transport in tokamaks driven by low-frequency collective fluctuations. This work addresses one of the greatest intellectual challenges to our understanding of the physics of tokamak performance and needs the most advanced computational resources to progress. We are conducting detailed comparisons of kinetic and fluid numerical models of tokamak turbulence. These comparisons are stimulating the improvement of each and the development of hybrid models which embody aspects of both. The combination of emerging massively parallel processing hardware and algorithmic improvements will result in an estimated 10**2--10**6 performance increase. Development of information processing and visualization tools is accelerating our comparison of computational models to one another, to experimental data, and to analytical theory, providing a bootstrap effect in our understanding of the target physics. The measure of success is the degree to which the experimentally observed scaling of fluctuation-driven transport may be predicted numerically. The NTP is advancing the HPCC Initiative through its state-of-the-art computational work. We are pushing the capability of high performance computing through our efforts which are strongly leveraged by OFE support

  13. Fusion plasma experiments on TFTR: A 20 year retrospective*

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R. J.; Batha, S.; Blanchard, W.; Beer, M.; Bell, M. G.; Bell, R. E.; Berk, H.; Bernabei, S.; Bitter, M.; Breizman, B.; Bretz, N. L; Budny, R.; Bush, C. E.; Callen, J.; Camp, R.; Cauffman, S.; Chang, Z.; Cheng, C. Z.; Darrow, D. S.; Dendy, R. O.; Dorland, W.; Duong, H.; Efthimion, P. C.; Ernst, D.; Fisch, N. J.; Fisher, R.; Fonck, R. J.; Fredrickson, E. D.; Fu, G. Y.; Furth, H. P.; Gorelenkov, N. N.; Grek, B.; Grisham, L. R.; Hammett, G. W.; Hanson, G. R.; Herrmann, H. W.; Herrmann, M. C.; Hill, K. W.; Hogan, J.; Hosea, J. C.; Houlberg, W. A.; Hughes, M.; Hulse, R. A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Kaita, R.; Kaye, S.; Kim, J. S.; Kissick, M.; Krasilnikov, A. V.; Kugel, H.; Kumar, A.; Leblanc, B.; Levinton, F. M.; Ludescher, C.; Majeski, R. P.; Manickam, J.; Mansfield, D. K.; Mazzucato, E.; McChesney, J.; McCune, D. C.; McGuire, K. M.; Meade, D. M.; Medley, S. S.; Mika, R.; Mikkelsen, D. R.; Mirnov, S. V.; Mueller, D.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Okabayashi, M.; Park, H. K.; Park, W.; Paul, S. F.; Pearson, G.; Petrov, M. P.; Phillips, C. K.; Phillips, M.; Ramsey, A. T.; Redi, M. H.; Rewoldt, G.; Reznik, S.; Roquemore, A. L.; Rogers, J.; Ruskov, E.; Sabbagh, S. A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G. L.; Scott, S. D.; Semenov, I.; Skinner, C. H.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Thompson, M. E.; Von Goeler, S.; Von Halle, A.; Walters, R. T.; White, R.; Wieland, R. M.; Williams, M.; Wilson, J. R.; Wong, K. L.; Wurden, G. A.; Yamada, M.; Yavorski, V.; Young, K. M.; Zakharov, L.; Zarnstorff, M. C.; Zweben, S. J. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

    1998-01-01

    The Tokamak Fusion Test Reactor (TFTR) (R. J. Hawryluk, to be published in Rev. Mod. Phys.) experiments on high-temperature plasmas, that culminated in the study of deuterium–tritium D–T plasmas containing significant populations of energetic alpha particles, spanned over two decades from conception to completion. During the design of TFTR, the key physics issues were magnetohydrodynamic (MHD) equilibrium and stability, plasma energy transport, impurity effects, and plasma reactivity. Energetic particle physics was given less attention during this phase because, in part, of the necessity to address the issues that would create the conditions for the study of energetic particles and also the lack of diagnostics to study the energetic particles in detail. The worldwide tokamak program including the contributions from TFTR made substantial progress during the past two decades in addressing the fundamental issues affecting the performance of high-temperature plasmas and the behavior of energetic particles. The progress has been the result of the construction of new facilities, which enabled the production of high-temperature well-confined plasmas, development of sophisticated diagnostic techniques to study both the background plasma and the resulting energetic fusion products, and computational techniques to both interpret the experimental results and to predict the outcome of experiments. © 1998 American Institute of Physics.

  14. Remote operation of the vertical plasma stabilization @ the GOLEM tokamak for the plasma physics education

    Energy Technology Data Exchange (ETDEWEB)

    Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Kocman, J.; Grover, O. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Krbec, J.; Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, CZ-182 21 Prague (Czech Republic)

    2015-10-15

    Graphical abstract: * Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes.* Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform.* More than 20% plasma life prolongation with plasma position control in feedback mode. - Highlights: • Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes. • Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform. • More than 20% plasma life prolongation with plasma position control in feedback mode. - Abstract: The GOLEM tokamak at the Czech Technical University has been established as an educational tokamak device for domestic and foreign students. Remote participation in the scope of several laboratory practices, plasma physics schools and workshops has been successfully performed from abroad. A new enhancement allowing understandable remote control of vertical plasma position in two modes (i) predefined and (ii) feedback control is presented. It allows to drive the current in the stabilization coils in any time-dependent scenario, which can include as a parameter the actual plasma position measured by magnetic diagnostics. Arbitrary movement of the plasma column in a vertical direction, stabilization of the plasma column in the center of the tokamak vessel as well as prolongation/shortening of plasma life according to the remotely defined request are demonstrated.

  15. Nonlinear MHD simulations of spherical tokamak and helical plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, T.; Mizuguchi, N.; Miura, H.; Kanno, R.; Nakajima, N.; Okamoto, M. [National Inst. for Fusion Science, Toki, Gifu (Japan)

    2002-11-01

    Nonlinear magnetohydrodynamic (MHD) simulations on relaxation phenomena in a spherical tokamak and a helical plasma, including three-dimensional (3D) equilibrium computations, are executed in full toroidal geometries. For a spherical tokamak, two-step evolution of the medium-n ballooning instabilities and a successive excitation of an internal n=1 crash has been observed. A similar process has been observed in an actual tokamak experiment of TFTR. The process is compared with a previous simulation for another MHD activity that is called the internal reconnection event (IRE). Three dimensional equilibrium code HINT is modified to extend functions, by which an n=1 island structure used for the local island diverter (LID) experiment in the LHD device is analyzed in finite pressure equilibria. Nonlinear simulations are executed for LHD plasma and pressure deformation due to evolution of m=2/n=1 pressure driven mode is observed, which has larger growth rate and saturation level than medium-n ballooning modes. (author)

  16. Review of deuterium--tritium results from the Tokamak Fusion Test Reactor*

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, K. M.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J. L.; Anderson, J W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, C. W.; Barnes, G.; Batha, S.; Bateman, G.; Beer, M; Bell, M. G.; Bell, R.; Bitter, M.; Blanchard, W.; Bretz, N. L.; Brunkhorst, C.; Budny, R.; Bush, C. E.; Camp, R.; Caorlin, M.; Carnevale, H.; Cauffman, S.; Chang, Z.; Chang, C. S.; Cheng, C. Z.; Chrzanowski, J.; Collins, J.; Coward, G.; Cropper, M.; Darrow, D. S; Daugert, R.; DeLooper, J.; Dendy, R.; Dorland, W.; Dudek, L.; Duong, H.; Durst, R.; Efthimion, P. C.; Ernst, D.; Evenson, H.; Fisch, N.; Fisher, R.; Fonck, R. J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G. Y.; Fujita, T.; Furth, H. P.; Garzotto, V.; Gentile, C.; Gilbert, J.; Gioia, J.; Gorelenkov, N.; Grek, B.; Grisham, L. R.; Hammett, G.; Hanson, G. R.; Hawryluk, R. J.; Heidbrink, W.; Herrmann, H. W.; Hill, K. W.; Hosea, J.; Hsuan, H.; Hughes, M.; Hulse, R.; Janos, A.; Jassby, D. L.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kalish, M.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Labik, G.; Lam, N. T.; LaMarche, P. H.; Lawson, E.; LeBlanc, B.; Levine, J.; Levinton, F. M.; Loesser, D.; Long, D.; Loughlin, M. J.; Machuzak, J.; Majeski, R.; Mansfield, D. K.; Marmar, E. S.; Marsala, R.; Martin, A.; Martin, G.; Mazzucato, E.; Mauel, M.; McCarthy, M. P.; McChesney, J.; McCormack, B.; McCune, D. C.; McKee, G.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Mirnov, S. V.; Mueller, D.; Murakami, M.; Murphy, J. A.; Nagy, A.; Navratil, G. A.; Nazikian, R.; Newman, R.; Norris, M.; O`Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D. K.; Park, H.; Park, W.; Parks, P.; Paul, S. F.; Pearson, G.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C. K.; Phillips, M.; Pitcher, S.; Pysher, R.; Qualls, A. L.; Raftopoulos, S.; Ramakrishnan, S.; Ramsey, A.; Rasmussen, D. A.; Redi, M. H.; Renda, G.; Rewoldt, G.; Roberts, D.; Rogers, J.; Rossmassler, R.; Roquemore, A. L.; Ruskov, E.; Sabbagh, S. A.; Sasao, M.; Schilling, G.; Schivell, J.; Schmidt, G.; Scillia, R.; Scott, S. D.; Semenov, I.; Senko, T.; Sesnic, S.; Sissingh, R.; Skinner, C. H.; Snipes, J.; Stencel, J.; Stevens, J.; Stevenson, T.; Stratton, B. C.; Strachan, J. D.; Stodiek, W.; Swanson, J.; Synakowski, E.; Takahashi, H.; Tang, W.; Taylor, G.; Terry, J.; Thompson, M. E.; Tighe, W.; Timberlake, J. R.; Tobita, K.; Towner, H. H.; Tuszewski, M.; Halle, A. Von; Vannoy, C.; Viola, M.; Goeler, S. Von; Voorhees, D.; Walters, R. T.; Wester, R.; White, R.; Wieland, R.; Wilgen, J. B.; Williams, M.; Wilson, J. R.; Winston, J.; Wright, K.; Wong, K. L.; Woskov, P.; Wurden, G. A.; Yamada, M.; Yoshikawa, S.; Young, K. M.; Zarnstorff, M. C.; Zavereev, V.; Zweben, S. J.

    1995-01-01

    The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≈ 2.8 MW m₋3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni (0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

  17. Trapping of gun-injected plasma by a tokamak

    International Nuclear Information System (INIS)

    Leonard, A.W.; Dexter, R.N.; Sprott, J.C.

    1987-01-01

    It has been seen that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. This trapping of a gun-injected plasma is explained in terms of a depolarization current mechanism. A model is developed that describes the slowing of a plasma beam crossing into the magnetic field of a tokamak. The slowing down time is shown to go as tau/sub s/proportionalT/sup 3/2//sub e/L 2 /n/sub b/α 2 0 , where n/sub b/ and T/sub e/ are the density and temperature of the plasma beam and α 0 /L is the pitch of the field lines per unit length in the direction in which the beam is traveling. Experimental tests of this model are consistent with the scaling predictions

  18. Trapping of gun-injected plasma by a tokamak

    International Nuclear Information System (INIS)

    Leonard, A.W.; Dexter, R.N.; Sprott, J.C.

    1986-10-01

    It is shown that a plasma produced by a Marshall gun can be injected into and trapped by a tokamak plasma. Gun injection raises the line-averaged density and peaks the density profile. Trapping of the gun-injected plasma is explainable in terms of a depolarization current mechanism. A model is developed which describes the slowing of a plasma beam crossing into the magnetic field of a tokamak. The slowing down time is shown to go as tau/sub s/ ∞ n -1 /sub b/T 3 /sub e/(α 0 /L) 2 , where n/sub b/ and T/sub e/ are the density and temperature of the plasma beam and α 0 /L is the pitch of the field lines per unit length in the direction in which the beam is traveling. Experimental tests of this model are consistent with the scaling predictions

  19. 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)

  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. Studies on the parametric decay of waves in fusion plasmas

    International Nuclear Information System (INIS)

    Paettikangas, T.

    1992-08-01

    Parametric instabilities of large-amplitude electromagnetic waves are investigated in fusion applications. In laser fusion, the electromegnetic wave reflected from the overdense plasma can act as a secondary pump wave and exite parametric instabilities. In double simulated Brilloun scattering (DSBS), both the incoming and the reflected pump wave scatter from a common ion sound wave. The stationary states and the dynamics of DSBS are investigated by using a simple envelope model. The ion sound wave that is exited in DSBS is shown to have soliton-like properties. The simulated Raman scattering (SRS) of free-electron-laser radiation can be applied to current drive in tokamaks. SRS generates fast longitudinal electron plasma waves which accelerate electrons to relativistic energies. Since the energetic current-carrying electrons are almost collisionless, the current decays very slowly. The feasibility of the Raman current drive in tokamaks is investigated theoretically. The current drive efficiency and the optimum free-electron-laser parameters are determined. The energy transfer to the fast electrons from the electrostatic wave is studied with relativistic Vlasov-Maxwell simulations. The parametric decay of a wave to half-harmonics is investigated. It is shown that the growth rate of the decay vanishes in the limit of a long wavelenght of the pump wave even for general electromagnetic or electrostatic decay models. The results are applied to the decay of a fast magnetosonic waves in tokamak plasmas. (orig.)

  2. Near-wall plasma in a tokamak thermonuclear reactor

    International Nuclear Information System (INIS)

    Pistunovich, V.I.

    1991-01-01

    The development of thermonuclear reactor projects such as the national ones GTRT and OTR (USSR), UWMAK-NUMAK and TIBER (USA), NET (Euratom), FER (Japan), as well as the international ones INTOR and JET allows one to determine in detail the problems arising in the creation of a reactor. One of these problems is extracting heat and particles from the reactor and removing thermonuclear reaction ash, which in the case of a DT reactor is helium. The solution of this problem is connected with the investigation of interaction processes of the plasma with the materials from which the reactor is constructed. Progress in studying the interaction of plasma with a material surface has given rise to a new division of plasma physics, which includes not only reactor problems, but also plasma-physical processes occurring in the region between the hot plasma and the first wall. Experiments on tokamaks have shown that the phenomena occurring in the wall plasma and in the hot zone of the tokamak once strongly correlated. In and number of regimes the confinement of the hot plasma is determined by the conditions in the peripheral region. A survey of the near-well plasma problem in a tokamak thermonuclear reactor is given. The results of recent investigations are presented, and the status of the problem as applied to the ITER reactor project is described

  3. Measurement of plasma current in Tokamaks using an optical fibre reflectometry technique

    Directory of Open Access Journals (Sweden)

    Wuilpart Marc

    2018-01-01

    Full Text Available An optical time-domain reflectometer sensitive to the polarization of light is proposed for the measurement of plasma current in the Tore Supra fusion reactor. The measurement principle relies on the Faraday effect i.e. on the generation of a circular birefringence along an optical fiber subject to an axial magnetic field. The circular birefringence induces a polarization rotation that can be mapped along the fiber thanks to an opticaltime domain reflectometer followed by an linear polarizer. A proper fitting of the measurement trace then allows determining the applied plasma current. The sensor has been experimentally validated on the Tore Supra tokamak fusion reactor for a plasma current range going from 0.6 to 1.5 MA. A maximum error of 13.50% has been observed for the lowest current.

  4. Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

    Science.gov (United States)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion

  5. Power supply requirements for a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Brooks, J.N.; Kustom, R.L.

    1979-02-01

    The power supply requirements for a 7-M major radius commercial tokamak reactor have been examined, using a system approach combining models of the reactor and poloidal coil set, plasma burn cycle and MHD calculations, and power supply characteristics and cost data. A conventional system using an MGF set and solid-state rectifier/inverter power supplies was studied in addition to systems using a homopolar generator, superconducting energy storage inductor, and dump resistors. The requirements and cost of the power supplies depend on several factors but most critically on the ohmic heating ramp time used for startup. Long ramp times (approx. > 8 s) seems to be feasible, from the standpoint of resistive volt-second losses, and would appear to make conventional systems quite competitive with nonconventional ones, which require further research and development

  6. FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Suratia, Pooja, E-mail: poojasuratia@yahoo.com [Electrical Engineering Department, Faculty of Technology and Engineering, The Maharaja Sayajirao University of Baroda, Kalabhavan, Vadodara 390001, Gujarat (India); Patel, Jigneshkumar, E-mail: jjp@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India); Rajpal, Rachana, E-mail: rachana@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India); Kotia, Sorum, E-mail: smkotia-eed@msubaroda.ac.in [Electrical Engineering Department, Faculty of Technology and Engineering, The Maharaja Sayajirao University of Baroda, Kalabhavan, Vadodara 390001, Gujarat (India); Govindarajan, J., E-mail: govindarajan@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat (India)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer Evaluation and comparison of the working performance of FLC is done with that of PID Controller. Black-Right-Pointing-Pointer FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. Black-Right-Pointing-Pointer FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. Black-Right-Pointing-Pointer Developed FLC controller is able to maintain the plasma column within required range of {+-}0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional-Integral-Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).

  7. Fusion Plasma Physics and ITER - An Introduction (1/4)

    CERN Multimedia

    CERN. Geneva

    2011-01-01

    In November 2006, ministers representing the world’s major fusion research communities signed the agreement formally establishing the international project ITER. Sited at Cadarache in France, the project involves China, the European Union (including Switzerland), India, Japan, the Russian Federation, South Korea and the United States. ITER is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. The ITER tokamak is designed to study the “burning plasma” regime in deuterium-tritium (D-T) plasmas by achieving a fusion amplification factor, Q (the ratio of fusion output power to plasma heating input power), of 10 for several hundreds of seconds with a nominal fusion power output of 500MW. It is also intended to allow the study of steady-state plasma operation at Q≥5 by me...

  8. Internal magnetic field measurement in tokamak plasmas using a ...

    Indian Academy of Sciences (India)

    Abstract. In a tokamak plasma, the poloidal magnetic field profile closely depends on the current density profile. ... Transitions between the energy levels that are split by the magnetic field give rise to several components of a spectral ... train of charged pulses having an amplitude proportional to the light intensity sensed by.

  9. Modelling multi-ion plasma gun simulations of Tokamak disruptions

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1995-08-01

    The effect of impurity ions in plasma gun ablation tests of various targets is considered. Inclusion of reasonable amounts of impurity (∼10%) is adequate to explain observed energy transmission and erosion measurements. The gun tests and the computer code calculations are relevant to the parameter range expected for major disruptions on large tokamaks

  10. Controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    1994-07-01

    40 papers are presented at this 21. conference on controlled fusion and plasma physics (JET). Titles are: effects of sawtooth crashes on beams ions and fusion product tritons; beta limits in H-modes and VH-modes; impurity induced neutralization of MeV energy protons in JET plasmas; lost α particle diagnostic for high-yield D-T fusion plasmas; 15-MeV proton emission from ICRF-heated plasmas; pulse compression radar reflectometry for density measurements; gamma-ray emission profile measurements during ICRH discharges; the new JET phase ICRH array; simulation of triton burn-up; parametric dependencies of JET electron temperature profiles; detached divertor plasmas; excitation of global Alfven Eigenmodes by RF heating; mechanisms of toroidal rotation; effect of shear in the radial electric field on confinement; plasma transport properties at the L-H transition; numerical study of plasma detachment conditions in JET divertor plasmas; the SOL width and the MHD interchange instability; non linear magnetic reconnection in low collisionality plasmas; topology and slowing down of high energy ion orbits; sawtooth crashes at high beta; fusion performances and alpha heating in future JET D-T plasmas; a stable route to high-beta plasmas with non-monotonic q-profiles; theory of propagation of changes to confinement; spatial distribution of gamma emissivity and fast ions during ICRF heating; multi-camera soft X-ray diagnostic; radiation phenomena and particle fluxes in the X-event; local measurement of transport parameters for laser injected trace impurities; impurity transport of high performance discharges; negative snakes and negative shear; neural-network charge exchange analysis; ion temperature anisotropy in helium neutral beam fuelling; impurity line emission due to thermal charge exchange in edge plasmas; control of convection by fuelling and pumping; VH mode accessibility and global H-mode properties; ion cyclotron emission by spontaneous emission; LHCD/ICRH synergy

  11. 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

  12. Fusion reactor development: A review

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This paper is a review of the current prospects for fusion reactor development based upon the present status in plasma physics research, fusion technology development and reactor conceptual design for the tokamak magnetic confinement concept. Recent advances in tokamak plasma research and fusion technology development are summarized. The direction and conclusions of tokamak reactor conceptual design are discussed. The status of alternate magnetic confinement concept research is reviewed briefly. A feasible timetable for the development of fusion reactors is presented

  13. 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

  14. Problems with the concept of plasma equilibrium in tokamaks

    International Nuclear Information System (INIS)

    Carreras, B.A.

    1992-01-01

    The equilibrium condition for a magnetically confined plasma in normally formulated in terms of macroscopic equations. In these equations, the plasma pressure is assumed to be a function of the magnetic flux with continuous derivatives. However, in three- dimensional systems this is not necessarily the case. Here, we look at the case of an intrinsically three-dimensional realistic tokamak, and we discuss the possible interconnection between the equilibrium and anomalous transport

  15. Tokamak Fusion Core Experiment: design studies based on superconducting and hybrid toroidal field coils. Design overview

    Energy Technology Data Exchange (ETDEWEB)

    Flanagan, C.A. (ed.)

    1984-10-01

    This document is a design overview that describes the scoping studies and preconceptual design effort performed in FY 1983 on the Tokamak Fusion Core Experiment (TFCX) class of device. These studies focussed on devices with all-superconducting toroidal field (TF) coils and on devices with superconducting TF coils supplemented with copper TF coil inserts located in the bore of the TF coils in the shield region. Each class of device is designed to satisfy the mission of ignition and long pulse equilibrium burn. Typical design parameters are: major radius = 3.75 m, minor radius = 1.0 m, field on axis = 4.5 T, plasma current = 7.0 MA. These designs relay on lower hybrid (LHRH) current rampup and heating to ignition using ion cyclotron range of frequency (ICRF). A pumped limiter has been assumed for impurity control. The present document is a design overview; a more detailed design description is contained in a companion document.

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

    Energy Technology Data Exchange (ETDEWEB)

    Martin, R.; Manero, F.

    1984-07-01

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

  17. Influence of the plasma edge on tokamak performance

    International Nuclear Information System (INIS)

    Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

    2001-01-01

    A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge magneto-hydrodynamic (MHD) instabilities, plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge, to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. This paper examines these phenomena and their interaction. (author)

  18. Influence of the plasma edge on tokamak performance

    International Nuclear Information System (INIS)

    Wilson, H.R.; Connor, J.W.; Field, A.R.; Fielding, S.J.; Hastie, R.J.; Taylor, J.B.; Miller, R.L.

    1999-01-01

    A number of edge plasma physics phenomena are considered to determine tokamak performance: transport barrier, edge magneto-hydrodynamic (MHD) instabilities, plasma flow. These phenomena are thought to be causally related: a spontaneous increase in the plasma flow (actually, its radial variation) suppresses heat and particle fluxes at the plasma edge, to form a transport barrier; the edge pressure gradient steepens until limited by MHD instabilities, resulting in a temperature pedestal at the top of the steep gradient region; a number of core transport models predict enhanced confinement for higher values of the temperature pedestal. This paper examines these phenomena and their interaction. (author)

  19. Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Bruhn, M.L.

    1988-04-01

    Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs.

  20. Time-dependent free boundary equilibrium and resistive diffusion in a tokamak plasma

    International Nuclear Information System (INIS)

    Selig, G.

    2012-12-01

    In a Tokamak, in order to create the necessary conditions for nuclear fusion to occur, a plasma is maintained by applying magnetic fields. Under the hypothesis of an axial symmetry of the tokamak, the study of the magnetic configuration at equilibrium is done in two dimensions, and is deduced from the poloidal flux function. This function is solution of a non linear partial differential equation system, known as equilibrium problem. This thesis presents the time dependent free boundary equilibrium problem, where the circuit equations in the tokamak coils and passive conductors are solved together with the Grad-Shafranov equation to produce a dynamic simulation of the plasma. In this framework, the Finite Element equilibrium code CEDRES has been improved in order to solve the aforementioned dynamic problem. Consistency tests and comparisons with the DINA-CH code on an ITER vertical instability case have validated the results. Then, the resistive diffusion of the plasma current density has been simulated using a coupling between CEDRES and the averaged one-dimensional diffusion equation, and it has been successfully compared with the integrated modeling code CRONOS. (author)

  1. Features of self-organized plasma physics in tokamaks

    Science.gov (United States)

    Razumova, K. A.

    2018-01-01

    The history of investigations the role of self-organization processes in tokamak plasma confinement is presented. It was experimentally shown that the normalized pressure profile is the same for different tokamaks. Instead of the conventional Fick equation, where the thermal flux is proportional to a pressure gradient, processes in the plasma are well described by the Dyabilanin’s energy balance equation, in which the heat flux is proportional to the difference of normalized gradients for self-consistent and real pressure profiles. The transport coefficient depends on the values of heat flux, which compensates distortion of the pressure profile with external impacts. Radiative cooling of the plasma edge decreases the heat flux and improves the confinement.

  2. Pseudo-MHD ballooning modes in tokamak plasmas

    International Nuclear Information System (INIS)

    Callen, J.D.; Hegna, C.C.

    1996-08-01

    The MHD description of a plasma is extended to allow electrons to have both fluid-like and adiabatic-regime responses within an instability eigenmode. In the resultant open-quotes pseudo-MHDclose quotes model, magnetic field line bending is reduced in the adiabatic electron regime. This makes possible a new class of ballooning-type, long parallel extent, MHD-like instabilities in tokamak plasmas for α > s 2 (2 7/3 /9) (r p /R 0 ) or-d√Β/dr > (2 1/6 /3)(s/ R 0q ), which is well below the ideal-MHD stability boundary. The marginally stable pressure profile is similar in both magnitude and shape to that observed in ohmically heated tokamak plasmas

  3. Far infrared fusion plasma diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Luhmann, N.C. Jr.; Peebles, W.A.

    1990-01-01

    Over the last several years, reflectometry has grown in importance as a diagnostic for both steady-state density Profiles as well as for the investigation of density fluctuations and turbulence. As a diagnostic for density profile measurement, it is generally believed to be well understood in the tokamak environment. However, its use as a fluctuation diagnostic is hampered by a lack of quantitative experimental understanding of its wavenumber sensitivity and spatial resolution. Several researchers, have theoretically investigated these questions. However, prior to the UCLA laboratory investigation, no group has experimentally investigated these questions. Because of the reflectometer's importance to the world effort in understanding plasma turbulence and transport, UCLA has, over the last year, made its primary Task IIIA effort the resolution of these questions. UCLA has taken the lead in a quantitative experimental understanding of reflectometer data as applied to the measurement of density fluctuations. In addition to this, work has proceeded on the design, construction, and installation of a reflectometer system on UCLA's CCT tokamak. This effort will allow a comparison between the improved confinement regimes (H-mode) observed on both the DIII-D and CCT machines with the goal of achieving a physics understanding of the phenomena. Preliminary investigation of a new diagnostic technique to measure density profiles as a function of time has been initiated at UCLA. The technique promises to be a valuable addition to the range of available plasma diagnostics. Work on advanced holographic reflectometry technique as applied to fluctuation diagnostics has awaited a better understanding of the reflectometer signal itself as discussed above. Efforts to ensure the transfer of the diagnostic developments have continued with particular attention devoted to the preliminary design of a multichannel FIR interferometer for MST.

  4. Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Ja

    1994-05-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. [bold 21], 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert[sup TM] system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of [approx]10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of [alpha]-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined [alpha] particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed.

  5. Automation of Aditya tokamak plasma position control DC power supply

    Energy Technology Data Exchange (ETDEWEB)

    Arambhadiya, Bharat, E-mail: bharat@ipr.res.in; Raj, Harshita; Tanna, R.L.; Edappala, Praveenlal; Rajpal, Rachana; Ghosh, Joydeep; Chattopadhyay, P.K.; Kalal, M.B.

    2016-11-15

    Highlights: • Plasma position control is very essential for obtaining repeatable high temperature, high-density discharges of longer durations in tokomak. • The present capacitor bank has limitations of maximum current capacity and position control beyond 200 ms. • The installation of a separate set of coils and a DC power supply can control the plasma position beyond 200 ms. • A high power thyristor (T588N1200) triggers for DC current pulse of 300 A fires precisely at required positions to modify plasma position. • The commissioning is done for the automated in-house, quick and reliable solution. - Abstract: Plasma position control is essential for obtaining repeatable high temperature, high-density discharges of longer duration in tokamaks. Recently, a set of external coils is installed in the vertical field mode configuration to control the radial plasma position in ADITYA tokamak. The existing capacitor bank cannot provide the required current pulse beyond 200 ms for position control. This motivated to have a DC power supply of 500 A to provide current pulse beyond 200 ms for the position control. The automatization of the DC power supply mandated interfaces with the plasma control system, Aditya Pulse Power supply, and Data acquisition system for coordinated discharge operation. A high current thyristor circuit and a timer circuit have been developed for controlling the power supply automatically for charging vertical field coils of Aditya tokamak. Key protection interlocks implemented in the development ensure machine and occupational safety. Fiber-optic trans-receiver isolates the power supply with other subsystems, while analog channel is optically isolated. Commissioning and testing established proper synchronization of the power supply with tokamak operation. The paper discusses the automation of the DC power supply with main circuit components, timing control, and testing results.

  6. Thermographic analysis of plasma facing components covered by carbon surface layer in tokamaks

    International Nuclear Information System (INIS)

    Gardarein, Jean-Laurent

    2007-01-01

    Tokamaks are reactors based on the thermonuclear fusion energy with magnetic confinement of the plasma. In theses machines, several MW are coupled to the plasma for about 10 s. A large part of this power is directed towards plasma facing components (PFC). For better understanding and control the heat flux transfer from the plasma to the surrounding wall, it is very important to measure the surface temperature of the PFC and to estimate the imposed heat flux. In most of tokamaks using carbon PFC, the eroded carbon is circulating in the plasma and redeposited elsewhere. During the plasma operations, this leads at some locations to the formation of thin or thick carbon layers usually poorly attached to the PFC. These surface layers with unknown thermal properties complicate the calculation of the heat flux from IR surface temperature measurements. To solve this problem, we develop first, inverse method to estimate the heat flux using thermocouple (not sensitive to the carbon surface layers) temperature measurements. Then, we propose a front face pulsed photothermal method allowing an estimation of layers thermal diffusivity, conductivity, effusivity and the thermal contact resistance between the layer and the tile. The principle is to study with an infrared sensor, the cooling of the layer surface after heating by a short laser pulse, this cooling depending on the thermal properties of the successive layers. (author) [fr

  7. ISTTOK tokamak plasmas influence on a liquid gallium jet dynamic behavior

    International Nuclear Information System (INIS)

    Gomes, R.B.; Silva, C.; Fernandes, H.; Duarte, P.; Nedzelskiy, I.; Lielausis, O.; Klyukin, A.; Platacis, E.

    2011-01-01

    The main concern in using free flowing liquid metals in fusion devices is related to their interaction with magnetic fields. On ISTTOK tokamak, liquid gallium jets are injected deep into the plasma along a vertical direction. The influence of the plasma interaction on the jet has been investigated monitoring the liquid metal behavior using a fast frame camera. A radial shift on its trajectory has been detected and found to depend on the toroidal magnetic field magnitude and principally on the plasma position within the chamber. The analysis performed to understand the dynamics of the jet perturbation by the plasma is presented in this paper. The jet surface temperature increase during this interaction has also been measured, using absolutely calibrated multichannel IR sensors, to evaluate the jet power exhaustion capability.

  8. ISTTOK tokamak plasmas influence on a liquid gallium jet dynamic behavior

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, R.B., E-mail: gomes@ipfn.ist.utl.pt [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, 1049-001 Lisboa (Portugal); Silva, C.; Fernandes, H.; Duarte, P.; Nedzelskiy, I. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, 1049-001 Lisboa (Portugal); Lielausis, O.; Klyukin, A.; Platacis, E. [Association EURATOM/University of Latvia, Institute of Solid State Physics, 8 Kengaraga Str., LV-1063 Riga (Latvia)

    2011-08-01

    The main concern in using free flowing liquid metals in fusion devices is related to their interaction with magnetic fields. On ISTTOK tokamak, liquid gallium jets are injected deep into the plasma along a vertical direction. The influence of the plasma interaction on the jet has been investigated monitoring the liquid metal behavior using a fast frame camera. A radial shift on its trajectory has been detected and found to depend on the toroidal magnetic field magnitude and principally on the plasma position within the chamber. The analysis performed to understand the dynamics of the jet perturbation by the plasma is presented in this paper. The jet surface temperature increase during this interaction has also been measured, using absolutely calibrated multichannel IR sensors, to evaluate the jet power exhaustion capability.

  9. Far-infrared fusion plasma diagnostics. Task IIIA. Final report

    International Nuclear Information System (INIS)

    Luhmann, N.C. Jr.

    1986-01-01

    The Task IIIA program at UCLA has been concerned with the development of innovative yet practical plasma diagnostic systems capable of providing detailed information essential to the success of the fusion program but not presently available within the fusion community. Historically, this has involved an initial development in the laboratory, followed by a test of feasibility on the Microtor tokamak prior to transfer of the technique/instrument to main line fusion devices. Strong emphasis has been placed upon the far-infrared (FIR) spectral region where novel diagnostic systems and technology have been developed and then distributed throughout the fusion program. The major diagnostics under development have been the measurement of plasma microturbulence and coherent modes via multichannel cw collective Thomson scattering, and the application of phase/polarization imaging techniques to provide accurate and detailed (>20 channel) electron density and current profiles not presently available using conventional methods. The eventual transfer of the above techniques to main line fusion devices is, of course, a major goal of the UCLA development program. The multichannel scattering development at UCLA was efficiently transferred to TEXT a few years ago. The apparatus has been employed to investigate the strong spectral and spatial asymmetries in the microturbulence uncovered through the unique multichannel and spatial scanning capabilities of the system. The scattering apparatus has also produced evidence for the ion pressure gradient driven eta/sub i/ modes thought responsible for anomalous transport in the edge regions of tokamak plasmas, as well as providing insight into the wave-wave coupling processes between various plasma modes

  10. Design study of superconducting toroidal magnet for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-10-01

    Design study of the superconducting toroidal magnet for the Tokamak experimental fusion reactor has been carried out. Performed here were conductor design, magnetic field calculation, design of coil support, stress analysis, design of refrigeration system and safety analysis. The maximum toroidal field at the coil is 12.0 T, providing 6.0 T at the plasma center. The coil bore is 7.3 x 11.2 m, and the coil shape is deformed constant-tension D-shape. 16 coil design was chosen, considering the necessity for good access for the evacuation port or the support structure of blanket and shield. The maximum field ripple obtained in the plasma region is 0.47%. The operational current is 25,100 amperes, and the cryogenic stabilization is fulfilled. Nb 3 Sn superconductor was chosen in spite of its brittleness, to attain a toroidal field of 12 T which is higher than a practical limit for NbTi superconductor. The development of large Nb 3 Sn coil technology would be necessary to realize an economic fusion power reactor. (auth.)

  11. Possibilities for breakeven and ignition of D-3He fusion fuel in a near term tokamak

    International Nuclear Information System (INIS)

    Emmert, G.A.; El-Guebaly, L.; Kulcinski, G.L.; Santarius, J.F.; Scharer, J.E.; Sviatoslavsky, I.N.; Walstrom, P.L.; Klinghoefer, R.; Wittenberg, J.L.

    1988-09-01

    The recent realization that the moon contains a large amount of the isotope 3 He has rekindled interest in the D- 3 He fuel cycle. In this study we consider the feasibility of investigating D- 3 He reactor plasma conditions in a tokamak of the NET/INTOR class. We have found that, depending on the energy confinement scaling law, energy breakeven may be achieved without significant modification to the NET design. The best results are for the more optimistic ASDEX H-mode scaling law. Kaye-Goldston scaling with a modest improvement due to the H-mode is more pessimistic and makes achieving breakeven more difficult. Significant improvement in Q (ratio of the fusion power to the injected power), or the ignition margin, can be achieved by taking advantage of the much reduced neutron production of the D- 3 He fuel cycle. Removal of the tritium producing blanket and replacing the inboard neutron shield by a thinner shield optimized for the neutron spectrum in D- 3 He allows the plasma to be increased without changing the magnetic field at the toroidal field magnet. This allows the plasma to achieve higher beta and Q values up to about 3. The implications of D- 3 He operation for fast ion loss, neutron shielding, heat loads on the first wall and divertor, plasma refuelling, changes to the poloidal field coil system, and pumping of the helium from the vacuum chamber are considered in the report. (orig.)

  12. Fusion Canada issue 17

    International Nuclear Information System (INIS)

    1992-05-01

    A short bulletin from the National Fusion Program. Included in this issue is a report on increased funding for the Canadian Fusion Program, news of the compact Toroid fuelling gun, an update on Tokamak de Varennes, the Canada - U.S. fusion meeting, measurements of plasma flow velocity, and replaceable Tokamak divertors. 4 figs

  13. Plasma radiation in tokamak disruption simulation experiments

    International Nuclear Information System (INIS)

    Arkhipov, N.; Bakhtin, V.; Safronov, V.; Toporkov, D.; Vasenin, S.; Zhitlukhin, A.; Wuerz, H.

    1995-01-01

    Plasma impact results in sudden evaporation of divertor plate material and produces a plasma cloud which acts as a protective shield. The incoming energy flux is absorbed in the plasma shield and is converted mainly into radiation. Thus the radiative characteristics of the target plasma determine the dissipation of the incoming energy and the heat load at the target. Radiation of target plasma is studied at the two plasma gun facility 2MK-200 at Troitsk. Space- and time-resolved spectroscopy and time-integrated space-resolved calorimetry are employed as diagnostics. Graphite and tungsten samples are exposed to deuterium plasma streams. It is found that the radiative characteristics depend strongly on the target material. Tungsten plasma arises within 1 micros close to the surface and shows continuum radiation only. Expansion of tungsten plasma is restricted. For a graphite target the plasma shield is a mixture of carbon and deuterium. It expands along the magnetic field lines with a velocity of v = (3--4) 10 6 cm/s. The plasma shield is a two zone plasma with a hot low dense corona and a cold dense layer close to the target. The plasma corona emits intense soft x-ray (SXR) line radiation in the frequency range from 300--380 eV mainly from CV ions. It acts as effective dissipation system and converts volumetrically the incoming energy flux into SXR radiation

  14. Plasma radiation in tokamak disruption simulation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, N.; Bakhtin, V.; Safronov, V.; Toporkov, D.; Vasenin, S.; Zhitlukhin, A. [Troitsk Inst. for Innovation and Fusion Research (Russian Federation); Wuerz, H. [Forschungszentrum Karlsruhe (Germany)

    1995-12-31

    Plasma impact results in sudden evaporation of divertor plate material and produces a plasma cloud which acts as a protective shield. The incoming energy flux is absorbed in the plasma shield and is converted mainly into radiation. Thus the radiative characteristics of the target plasma determine the dissipation of the incoming energy and the heat load at the target. Radiation of target plasma is studied at the two plasma gun facility 2MK-200 at Troitsk. Space- and time-resolved spectroscopy and time-integrated space-resolved calorimetry are employed as diagnostics. Graphite and tungsten samples are exposed to deuterium plasma streams. It is found that the radiative characteristics depend strongly on the target material. Tungsten plasma arises within 1 {micro}s close to the surface and shows continuum radiation only. Expansion of tungsten plasma is restricted. For a graphite target the plasma shield is a mixture of carbon and deuterium. It expands along the magnetic field lines with a velocity of v = (3--4) 10{sub 6} cm/s. The plasma shield is a two zone plasma with a hot low dense corona and a cold dense layer close to the target. The plasma corona emits intense soft x-ray (SXR) line radiation in the frequency range from 300--380 eV mainly from CV ions. It acts as effective dissipation system and converts volumetrically the incoming energy flux into SXR radiation.

  15. Fusion alpha-particle losses in a high-beta rippled tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bunno, M.; Nakamura, Y. [Graduate School of Energy Science, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Suzuki, Y. [National Institute for Fusion Science, Toki 509-5292 (Japan); Shinohara, K.; Matsunaga, G. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Tani, K. [Nippon Advanced Technology, Naka, Ibaraki 311-0102 (Japan)

    2013-08-15

    In tokamak plasmas, the confinement of energetic ions depends on the magnetic field structure. If the plasma pressure is finite, the equilibrium current (i.e., the Pfirsch-Schlüter current and diamagnetic current) flows in the plasma to maintain the magnetohydrodynamic (MHD) equilibrium. These plasma currents generate poloidal and toroidal magnetic field and alter the field structure. Moreover, if we consider the non-axisymmetry of magnetic field structures such as toroidal field (TF) ripples, the non-axisymmetric component of the equilibrium current can alter TF ripples themselves. When the plasma beta becomes high, the changes in the field structure due to the equilibrium current might affect the confinement of energetic ions significantly. We intend to clarify how these currents alter the field structure and affect the confinement of alpha particles in high-beta plasma. The MHD equilibrium is calculated using VMEC and the orbits of fusion alpha particles are followed by using the fully three-dimensional magnetic field orbit-following Monte Carlo code. In relatively low-beta plasma (e.g., the volume-averaged beta value <β>≤2%), the changes in the magnetic field component due to the plasma current negligibly affect the confinement of alpha particles except for the Shafranov shift effect. However, for <β>≥3%, the diamagnetic effect reduces the magnetic field strength and significantly increases alpha-particle losses. In these high-beta cases, the non-axisymmetric field component generated by the equilibrium current also increases these losses, but not as effectively as compared to the diamagnetic effect.

  16. Initial plasma production by induction electric field on QUEST tokamak

    International Nuclear Information System (INIS)

    Hasegawa, Makoto; Nakamura, Kazuo; Sato, Kohnosuke

    2007-01-01

    Induction electric field by center solenoid coil plays a roll to produce initial plasma. According to Townsend avalanche theory, minimum electric field for plasma breakdown depends on neutral gas pressure and connection length. On QUEST spherical tokamak, a connection length is evaluated as 966m on null point neighborhood with coil current ratio I PF26 /I CS =0.1, and induction electric field considering eddy current of vacuum vessel is evaluated as about 0.1 V/m on null point neighborhood. With Townsend avalanche theory, these values manage to produce initial plasma on QUEST. (author)

  17. Plasma position control in SST1 tokamak

    Indian Academy of Sciences (India)

    Radial position control is also necessary for maintaining a predefined gap between plasma and lower hybrid wave launcher antenna for efficient coupling of the RF waves. The radial position of the plasma can change because of change in plasma beta and internal inductance. (henceforth referred to as ¬Ф +Р 2), due to ...

  18. The impact of servicing requirements on Tokamak fusion reactor design

    International Nuclear Information System (INIS)

    Mitchell, J.T.D.

    1978-01-01

    There is one design requirement which appears to be common to all D-T fusion reactors, irrespective of the plasma confinement principle embodied in the design. It must be practicable to repair the structure or 'first wall' nearest the plasma. A repair can include a range of operations between finding and sealing a pinhole leak in a weld to complete replacement of the whole structure of the breeding blanket. The two reasons why these repairs are necessary are: 1) The variability and defects in generally perfect products, such as those caused by errors of human performance or by flaws in raw materials. 2) The extreme severity of the environment of the first wall, which causes the deterioration of the properties of materials situated there. This deterioration is the result of neutron irradiation and transmutation in the materials under conditions of stress and high temperature and of surface corrosion and erosion

  19. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Castracane, J.

    2001-01-04

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.

  20. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    International Nuclear Information System (INIS)

    Castracane, J.

    2001-01-01

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies

  1. Plasma current profile during current reversal in a tokamak

    International Nuclear Information System (INIS)

    Huang Jianguo; Yang Xuanzong; Zheng Shaobai; Feng Chunhua; Zhang Houxian; Wang Long

    1999-01-01

    Alternating current operation with one full cycle and a current level of 2.5 kA have been achieved in the CT-6B tokamak. The poloidal magnetic field in the plasma is measured with two internal magnetic probes in repeated discharges. The current distribution is reconstructed with an inversion algorithm. The inverse current first appears on the weak field side. The existence of magnetic surfaces and rotational transform provide particle confinement in the current reversal phase

  2. Multi-field plasma sandpile model in tokamaks and applications

    Science.gov (United States)

    Peng, X. D.; Xu, J. Q.

    2016-08-01

    A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.

  3. Momentum Injection in Tokamak Plasmas and Transitions to Reduced Transport

    International Nuclear Information System (INIS)

    Parra, F. I.; Highcock, E. G.; Schekochihin, A. A.; Barnes, M.; Cowley, S. C.

    2011-01-01

    The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.

  4. Numerical studies of transport processes in Tokamak plasma

    International Nuclear Information System (INIS)

    Spineanu, F.; Vlad, M.

    1984-09-01

    The paper contains the summary of a set of studies of the transport processes in tokamak plasma, performed with a one-dimensional computer code. The various transport models (which are implemented by the expressions of the transport coefficients) are presented in connection with the regimes of the dynamical development of the discharge. Results of studies concerning the skin effect and the large scale MHD instabilities are also included

  5. 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.

  6. 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

  7. Plasma physics and controlled nuclear fusion research 1988. V.1

    International Nuclear Information System (INIS)

    1989-01-01

    Volume 1 of the Proceedings of the Twelfth International Conference on Plasma Physics and Controlled Nuclear Fusion contains papers given in two of the sessions: A and E. Session A contains, in addition to the Artsimovich Memorial Lecture, the papers on tokamaks, and session E the papers on heating and current drive. Each of these papers and their authors is listed in the Contents; in turn, each paper contains an abstract for more information as to the contents of a specific paper. Refs, figs and tabs

  8. MTX [Microwave Tokamak Experiment] plasma diagnostic system

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. 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)

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

    International Nuclear Information System (INIS)

    Sengoku, Seio

    1985-08-01

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

  11. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    Amiel, Stephane

    2014-01-01

    During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author) [fr

  12. Fusion reactor design studies

    International Nuclear Information System (INIS)

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

    1990-01-01

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

  13. Plasma diagnostics for tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Stott, P. E.; Sanchez, J.

    1994-07-01

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

  14. Plasma diagnostics for tokamaks and stellarators

    International Nuclear Information System (INIS)

    Stott, P.E.; Sanchez, J.

    1994-01-01

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

  15. Aspect Ratio Scaling of Ideal No-wall Stability Limits in High Bootstrap Fraction Tokamak Plasmas

    International Nuclear Information System (INIS)

    Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Maingi, R.; Sabbagh, S.A.; Soukhanovskii, V.; Stutman, D.

    2003-01-01

    Recent experiments in the low aspect ratio National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40 (2000) 557] have achieved normalized beta values twice the conventional tokamak limit at low internal inductance and with significant bootstrap current. These experimental results have motivated a computational re-examination of the plasma aspect ratio dependence of ideal no-wall magnetohydrodynamic stability limits. These calculations find that the profile-optimized no-wall stability limit in high bootstrap fraction regimes is well described by a nearly aspect ratio invariant normalized beta parameter utilizing the total magnetic field energy density inside the plasma. However, the scaling of normalized beta with internal inductance is found to be strongly aspect ratio dependent at sufficiently low aspect ratio. These calculations and detailed stability analyses of experimental equilibria indicate that the nonrotating plasma no-wall stability limit has been exceeded by as much as 30% in NSTX in a high bootstrap fraction regime

  16. Atomic and plasma-material interaction data for fusion. V. 5

    International Nuclear Information System (INIS)

    1994-01-01

    Volume 5 of the supplements on ''atomic and plasma-material interaction data for fusion'' to the journal ''Nuclear Fusion'' is devoted to a critical assessment of the physical and thermo-mechanical properties of presently considered candidate plasma-facing and structural materials for next-generation thermonuclear fusion devices. It contains 9 papers. The subjects are: (i) requirements and selection criteria for plasma-facing materials and components in the ITER EDA (Engineering Design Activities) design; (ii) thermomechanical properties of Beryllium; (iii) material properties data for fusion reactor plasma-facing carbon-carbon composites; (iv) high-Z candidate plasma facing materials; (v) recommended property data for Molybdenum, Niobium and Vanadium alloys; (vi) copper alloys for high heat flux structure applications; (vii) erosion of plasma-facing materials during a tokamak disruption; (viii) runaway electron effects; and (ix) data bases for thermo-hydrodynamic coupling with coolants. Refs, figs, tabs

  17. Pellet-plasma interactions in tokamaks

    DEFF Research Database (Denmark)

    Chang, C.T.

    1991-01-01

    The ablation of a refuelling pellet of solid hydrogen isotopes is governed by the plasma state, especially the density and energy distribution of the electrons. On the other hand, the cryogenic pellet gives rise to perturbations of the plasma temperature and density. Based on extensive experimental...... data, the interaction between the pellet and the plasma is reviewed. Among the subjects discussed are the MHD activity, evolution of temperature and density profiles, and the behaviour of impurities following the injection of a pellet (or pellets). The beneficial effect of density peaking on the energy...... of the plasma. The appearance of striations and the curving of the pellet trajectory are discussed in detail. The possibility is described for using these observations to study the plasma current-density distribution as well as the existence of suprathermal electrons....

  18. 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

  19. The National Spherical Tokamak Experiment at the Princeton Plasma Physics Laboratory

    International Nuclear Information System (INIS)

    1995-12-01

    The Department of Energy (DOE) has prepared an Environmental Assessment (EA), DOE/EA-1108, evaluating the environmental effects of the proposed construction and operation of the National Spherical Tokamak Experiment (NSTX) within the existing C-Stellarator (CS) Building at the Princeton Plasma Physics Laboratory, Princeton, New Jersey. The purpose of the NSTX is to investigate the physics of spherically shaped plasmas as an alternative path to conventional tokamaks for development of fusion energy. Fusion energy has the potential to help compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Construction of the NSTX in the CS Building would require the dismantling and removal of the existing unused Princeton Large Torus (PLT) device, part of which would be reused to construct the NSTX. Based on the analyses in the EA, the DOE has determined that the proposed action does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, 42 U.S.C. 4,321 et seq. The preparation of an Environmental Impact Statement is not required. Thus, the DOE is issuing a FONSI pursuant to the Council on Environmental Quality regulations implementing NEPA (40 CFR Parts 1500--1508) and the DOE NEPA implementing regulations (10 CFR Part 1021)

  20. Impurity screening of scrape-off plasma in a tokamak

    International Nuclear Information System (INIS)

    Kishimoto, Hiroshi; Tani, Keiji; Nakamura, Hiroo

    1981-11-01

    Impurity screening effect of a scrape-off layer has been studied in a tokamak, based on a simple model of wall-released impurity behavior. Wall-sputtered impurities are stopped effectively by the scrape-off plasma for a medium-Z or high-Z wall system while major part of impurities enters the main plasma in a low-Z wall system. The screening becomes inefficient with increase of scrape-off plasma temperature. Successive multiplication of recycling impurities in the scrape-off layer is large for a high-Z wall and is enhanced by a rise of scrape-off plasma temperature. The stability of plasma-wall interaction is determined by a multiplication factor of recycling impurities. (author)

  1. Evaluation of the plasma parameters in COMPASS tokamak divertor area

    Czech Academy of Sciences Publication Activity Database

    Dimitrova, M.; Ivanova, P.; Kotseva, I.; Popov, Tsv.K.; Benova, E.; Bogdanov, T.; Stöckel, Jan; Dejarnac, Renaud

    2012-01-01

    Roč. 356, č. 1 (2012), s. 012007 ISSN 1742-6588. [InternationalSummerSchoolonVacuum,Electron,andIonTechnologies(VEIT2011)/17./. Sunny Beach , 19.09.2011-23.09.2011] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostics * electric probe * magnetic-field * Langmuir probe * intermediate * pressures Subject RIV: BL - Plasma and Gas Discharge Physics http://iopscience.iop.org/1742-6596/356/1/012007/pdf/1742-6596_356_1_012007.pdf

  2. Plasma facing components design of KT-2 tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

    1997-04-01

    The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

  3. Electron temperature gradient driven instability in the tokamak boundary plasma

    International Nuclear Information System (INIS)

    Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.

    1992-01-01

    A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t -1/2 e γmt

  4. Continuum kinetic modeling of the tokamak plasma edge

    Energy Technology Data Exchange (ETDEWEB)

    Dorf, M. A.; Dorr, M. R.; Hittinger, J. A.; Rognlien, T. D. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Cohen, R. H. [CompX, P.O. Box 2672, Del Mar, California 92014 (United States)

    2016-05-15

    The first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalous radial transport.

  5. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    1999-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  6. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    2001-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  7. Preparations for deuterium tritium experiments on the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G. [and others

    1994-04-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR). These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinet{sup {trademark}} system, modification of the vacuum system to handle tritium, preparation and testing of the neutral beam system for tritium operation and a final deuterium-deuterium (D-D) run to simulate expected deuterium-tritium (D-T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D-T experiments using D-D have been performed. The physics objectives of D-T operation are production of {approximately} 10 megawatts (MW) of fusion power, evaluation of confinement and heating in deuterium-tritium plasmas, evaluation of {alpha}-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined {alpha}-particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed.

  8. Size scaling of turbulent transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Lin Zhihong

    2002-01-01

    Transport scaling with respect to tokamak device size is critically examined for electrostatic ion temperature gradient (ITG) turbulence with adiabatic electrons using first-principles gyrokinetic particle simulations, which use up to one billion particles to address realistic parameters of reactor-grade plasmas. Results of these large scale simulations, varying ρ* (ion gyroradius normalized by tokamak minor radius) while keeping other dimensionless plasma parameters fixed, show that the fluctuation scale length is microscopic and transport is diffusive in the presence of zonal flows. The local transport coefficient exhibits a gradual transition from a Bohm-like scaling for device sizes corresponding to present-day tokamak experiments to a gyro-Bohm scaling for future larger devices. The device size where this transition occurs is much larger than that expected from linear ITG theory for profile variations. Our simulations include a heat bath/source to prevent profile relaxation and are in the strong turbulence regime far away from ITG marginality. The effects of kinetic electrons on electrostatic ITG-TEM (trapped electron mode) driven turbulence will also be presented. (author)

  9. Emissive probe measurements of plasma potential fluctuations in the edge plasma regions of tokamaks

    Czech Academy of Sciences Publication Activity Database

    Balan, P.; Schrittweiser, R.; Ionita, C.; Cabral, J. A.; Figueiredo, H. F. C.; Fernandes, H.; Varandas, C.; Adámek, Jiří; Hron, Martin; Stöckel, Jan; Martines, E.; Tichý, M.; Van Oost, G.

    2003-01-01

    Roč. 74, č. 3 (2003), s. 1583-1587 ISSN 0034-6748 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : plasma physics, tokamaks, probes Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.343, year: 2003

  10. Thermal stability of the tokamak plasma edge

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1997-01-01

    The general linear, fluid, thermal instability theory for the plasma edge has been extended. An analysis of a two-dimensional fluid model of the plasma edge has identified the importance of many previously unappreciated phenomena associated with parallel and gyroviscous forces in the presence of large radial gradients, with large radial or parallel flows, with the temperature dependence of transport coefficients, and with the coupling of temperature, flow and density perturbations. The radiative condensation effect is generalized to include a further destabilizing condensation effect associated with radial heat conduction. Representative plasma edge neutral and impurity densities are found to be capable of driving thermal instabilities in the edge transport barrier and radiative mantle, respectively. (author)

  11. The COMPASS Tokamak Plasma Control Software Performance

    Czech Academy of Sciences Publication Activity Database

    Valcárcel, D.F.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Janky, F.; Havlíček, Josef; Beňo, R.; Horáček, Jan; Hron, Martin; Pánek, Radomír

    2011-01-01

    Roč. 58, č. 4 (2011), s. 1490-1496 ISSN 0018-9499. [Real Time Conference, RT10/17th./. Lisboa, 24.05.2010-28.05.2010] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Real-Time * ATCA * Data Acquisition * Plasma Control Software Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.447, year: 2011 http://dx.doi.org/10.1109/TNS.2011.2143726

  12. Numerical Simulation of Neoclassical Currents, Parallel Viscosity, and Radial Current Balance in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Kiviniemi, T.

    2001-01-01

    One of the principal problems en route to a fusion reactor is that of insufficient plasma confinement, which has lead to both theoretical and experimental research into transport processes in the parameter range relevant for fusion energy production. The neoclassical theory of tokamak transport is well-established unlike the theory of turbulence driven anomalous transport in which extensive progress has been made during last few years. So far, anomalous transport has been dominant in experiments, but transport may be reduced to the neoclassical level in advanced tokamak scenarios. This thesis reports a numerical study of neoclassical fluxes, parallel viscosity, and neoclassical radial current balance in tokamaks. Neoclassical parallel viscosity and particle fluxes are simulated over a wide range of collisionalities, using the fully kinetic five-dimensional neoclassical orbit-following Monte Carlo code ASCOT. The qualitative behavior of parallel viscosity derived in earlier analytic models is shown to be incorrect for high poloidal Mach numbers. This is because the poloidal dependence of density was neglected. However, in high Mach number regime, it is the convection and compression terms, rather than the parallel viscosity term, that are shown to dominate the momentum balance. For fluxes, a reasonable agreement between numerical and analytical results is found in the collisional parameter regime. Neoclassical particle fluxes are additionally studied in the banana regime using the three-dimensional Fokker-Planck code DEPORA, which solves the drift-kinetic equation with finite differencing. Limitations of the small inverse aspect ratio approximation adopted in the analytic theory are addressed. Assuming that the anomalous transport is ambipolar, the radial electric field and its shear at the tokamak plasma edge can be solved from the neoclassical radial current balance. This is performed both for JET and ASDEX Upgrade tokamaks using the ASCOT code. It is shown that

  13. Reversal of plasma flow in tokamak divertors

    International Nuclear Information System (INIS)

    Maddison, G.P.; Reiter, D.; Stangeby, P.C.; Prinja, A.K.

    1993-01-01

    In a magnetic divertor, retention of impurity ions is expected to be dependent on an expulsive thermal force directed up the gradient of ion temperature being opposed by frictional entrainment in a plasma flow towards the target. Preferred conditions of high recycling, however, can induce a reversal of usual plasma flow, with consequent reinforcement of thermal forces potentially leading to damaging contamination of the core. Backflow in diverted plasmas was first anticipated theoretically by Nedospasov and Tokar', subsequently observed experimentally in DITE and JET, and has been seen in a number of numerical studies. We report briefly on a systematic investigation of steady-state divertor flow reversal for ITER-relevant conditions, by detailed numerical modelling. The BRAAMS 'B2' edge plasma transport code is used, with both analytic approximations and EIRENE Monte Carlo simulation of neutral particle recycling. The flexibility of numerical models regarding physics admitted is exploited to expose the key role of redistribution of recycling sources across magnetic surfaces in flow reversal. Concomitant amplification of cross-field ion diffusion in the SOL is also examined. (author) 10 refs., 4 figs

  14. Investigation of plasma equilibrium in HL-1 tokamak

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  15. Plasma vertical instability in a tokamak with rail limiters

    International Nuclear Information System (INIS)

    Belashob, V.I.; Brevnov, N.N.; Gribov, Yu.V.; Putvinskij, S.V.

    1989-01-01

    An effect of currents between rail limiters on plasma equilibrium in the tokamak is studied theoretically and experimentally. Limiter currents can emerge at fast changes of plasma position along rail limiters for example when compression along major radius takes place and result in additional electrodynamic loadings on to the chamber and limiters. It is shown that at high currents between the limiters, the behaviour of discharge depends on limiter voltage polarity. When the plasma - limiter contact points are asymmetrically located respective to an equatorial plane a radial component of the limiter current emerges. The interaction of the component with the toroidal magnetic field can result in a vertical plasma filament instability. 9 refs.; 10 figs

  16. A Michelson interferometer/polarimeter on the Tokamak Fusion Test Reactor (TFTR)

    International Nuclear Information System (INIS)

    Park, H.K.; Mansfield, D.K.; Johnson, L.C.; Ma, C.H.

    1987-01-01

    A multichannel interferometer/polarimeter for the Tokamak Fusion Test Reactor (TFTR) has been developed in order to study the time dependent plasma current density (J/sub p/) and electron density (n/sub e/) profile simultaneously. The goal of the TFTR is demonstration of breakeven via dueuterium and tritium (DT) plasma. In order to be operated and maintained during DT operation phase, the system is designed based on the Michelson geometry which possesses intrinsic standing wave problems. So far, there has been no observable signals due to these standing waves. However, a standing wave resulted from the beam path design to achieve a optimum use of the laser power was found. This standing wave has not prevented initial 10 channel interferometer operation. However, a single channel polarimeter test indicated this standing wave was fatal for Faraday notation measurements. Techniques employing 1/2 wave plates and polarizers have been applied to eliminate this standing wave problem. The completion of 10 channel Faraday rotation measurements may be feasible in the near future

  17. Runaway electrons in disruptions and perturbed magnetic topologies of tokamak plasmas

    International Nuclear Information System (INIS)

    Forster, Michael

    2012-01-01

    Nuclear fusion represents a valuable perspective for a safe and reliable energy supply from the middle of the 21st century on. Currently, the tokamak is the most advanced principle of confining a man-made fusion plasma. The operation of future, reactor sized tokamaks like ITER faces a crucial difficulty in the generation of runaway electrons. The runaway of electrons is a free fall acceleration into the relativistic regime which is known in various kinds of plasmas including astrophysical ones, thunderbolts and fusion plasmas. The tokamak disruption instability can include the conversion of a substantial part of the plasma current into a runaway electron current. When the high energetic runaways are lost, they can strike the plasma facing components at localised spots. Due to their high energies up to a few tens of MeV, the runaways carry the potential to reduce the lifetimes of wall components and even to destroy sensitive, i.e. actively cooled parts. The research for effective ways to suppress the generation of runaway electrons is hampered by the lack of a complete understanding of the physics of the runaways in disruptions. As it is practically impossible to use standard electron detectors in the challenging environment of a tokamak, the experimental knowledge about runaways is limited and it relies on rather indirect techniques of measurement. The main diagnostics used for this PhD work are three reciprocating probes which measure the runaway electrons directly at the plasma edge of the tokamak TEXTOR. A calorimetric probe and a material probe which exploits the signature that a runaway beam impact leaves in the probe were developed in the course of the PhD work. Novel observations of the burst-like runaway electron losses in tokamak disruptions are reported. The runaway bursts are temporally resolved and first-time measurements of the corresponding runaway energy spectra are presented. A characteristic shape and typical burst to burst variations of the

  18. Lifetime evaluation for thermal fatigue: application at the first wall of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Merola, M.; Biggio, M.

    1989-01-01

    Thermal fatigue seems to be the most lifetime limiting phenomenon for the first wall of the next generation Tokamak fusion reactors. This work deals with the problem of the thermal fatigue in relation to the lifetime prediction of the fusion reactor first wall. The aim is to compare different lifetime methodologies among them and with experimental results. To fulfil this purpose, it has been necessary to develop a new numerical methodology, called reduced-3D, especially suitable for thermal fatigue problems

  19. Plasma facing materials for fusion reactor applications

    OpenAIRE

    Gonzalez Arrabal, Raquel; Gordillo Garcia, Nuria; Rivera de Mena, Antonio; Alvarez Ruiz, Jesus; Garoz, D.; Perlado Martín, José Manuel

    2012-01-01

    The lack of plasma facing materials (PFM) able to withstand the severe magnetiicffusiion radiation conditions expected in fusion reactors is the actual bottle In both fusions approaches energy is released in the form of kinetic energy of neck for fusion to becomes a reality.

  20. Achievement of solid-state plasma fusion ('Cold-Fusion')

    International Nuclear Information System (INIS)

    Arata, Yoshiaki; Zhang, Yue-Chang

    1995-01-01

    Using a 'QMS' (Quadrupole Mass Spectrometer), the authors detected a significantly large amount (10 20 -10 21 [cm -3 ]) of helium ( 2 4 He), which was concluded to have been produced by a deuterium nuclear reaction within a host solid. These results were found to be fully repeatable and supported the authors' proposition that solid state plasma fusion ('Cold Fusion') can be generated in energetic deuterium Strongly Coupled Plasma ('SC-plasma'). This fusion reaction is thought to be sustained by localized 'Latticequake' in a solid-state media with the deuterium density equivalent to that of the host solid. While exploring this basic proposition, the characteristic differences when compared with ultra high temperature-state plasma fusion ('Hot Fusion') are clarified. In general, the most essential reaction product in both types of the deuterium plasma fusion is considered to be helium, irrespective of the 'well-known and/or unknown reactions', which is stored within the solid-state medium in abundance as a 'Residual Product', but which generally can not enter into nor be released from host-solid at a room temperature. Even measuring instruments with relatively poor sensitivity should be able to easily detect such residual helium. An absence of residual helium means that no nuclear fusion reaction has occurred, whereas its presence provides crucial evidence that nuclear fusion has, in fact, occurred in the solid. (author)

  1. Measurement of the effective plasma ion mass in large tokamaks

    International Nuclear Information System (INIS)

    Lister, J.B.; Villard, L.; Ridder, G. de

    1997-01-01

    There is not yet a straightforward method for the measurement of the D-T ratio in the centre of a tokamak plasma. One of the simpler measurements put forward in the past is the interpretation of the MHD spectrum in the frequency range of the Global Alfven Eigenmodes (GAE). However, the frequencies of these modes do not only depend on the plasma mass, but are also quite strongly dependent on the details of the current and density profiles, creating a problem of deconvolution of the estimate of the plasma mass from an implicit relationship between several measurable plasma parameters and the detected eigenmode frequencies. This method has been revised to assess its likely precision for the JET tokamak. The low n GAE modes are sometimes too close to the continuum edge to be detectable and the interpretation of the GAE spectrum is rendered less direct than had been hoped. We present a statistical study on the precision with which the D-T ratio could be estimated from the GAE spectrum on JET. (author) 4 figs., 8 refs

  2. Plasma Physics An Introduction to Laboratory, Space, and Fusion Plasmas

    CERN Document Server

    Piel, Alexander

    2010-01-01

    Plasma Physics gives a comprehensive introduction to the basic processes in plasmas and demonstrates that the same fundamental concepts describe cold gas-discharge plasmas, space plasmas, and hot fusion plasmas. Starting from particle drifts in magnetic fields, the principles of magnetic confinement fusion are explained and compared with laser fusion. Collective processes are discussed in terms of plasma waves and instabilities. The concepts of plasma description by magnetohydrodynamics, kinetic theory, and particle simulation are stepwise introduced. Space charge effects in sheath regions, double layers and plasma diodes are given the necessary attention. The new fundamental mechanisms of dusty plasmas are explored and integrated into the framework of conventional plasmas. The book concludes with a brief introduction to plasma discharges. Written by an internationally renowned researcher in experimental plasma physics, the text keeps the mathematical apparatus simple and emphasizes the underlying concepts. T...

  3. Study of plasma turbulence by ultrafast sweeping reflectometry on the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Hornung, Gregoire

    2013-01-01

    The performance of a fusion reactor is closely related to the turbulence present in the plasma. The latter is responsible for anomalous transport of heat and particles that degrades the confinement. The measure and characterization of turbulence in tokamak plasma is therefore essential to the understanding and control of this phenomenon. Among the available diagnostics, the sweeping reflectometer installed on Tore Supra allows to access the plasma density fluctuations from the edge to the centre of the plasma discharge with a fine spatial (mm) and temporal resolution (μs), that is of the order of the characteristic turbulence scales.This thesis consisted in the characterization of plasma turbulence in Tore Supra by ultrafast sweeping reflectometry measurements. Correlation analyses are used to quantify the spatial and temporal scales of turbulence as well as their radial velocity. In the first part, the characterization of turbulence properties from the reconstructed plasma density profiles is discussed, in particular through a comparative study with Langmuir probe data. Then, a parametric study is presented, highlighting the effect of collisionality on turbulence, an interpretation of which is proposed in terms of the stabilization of trapped electron turbulence in the confined plasma. Finally, it is shown how additional heating at ion cyclotron frequency produces a significant though local modification of the turbulence in the plasma near the walls, resulting in a strong increase of the structure velocity and a decrease of the correlation time. The supposed effect of rectified potentials generated by the antenna is investigated via numerical simulations. (author) [fr

  4. Development of DEMO-FNS tokamak for fusion and hybrid technologies

    Science.gov (United States)

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

    2015-07-01

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

  5. Fusion. A voyage through the plasma universe

    Energy Technology Data Exchange (ETDEWEB)

    Wilhelmsson, H. [Chalmers University of Technology, Goeteborg (Sweden)

    2000-07-01

    This book adopts a novel approach to fusion plasmas, covering fusion both in the laboratory and in the cosmos. It discusses the evolution of the field from early plasma research to the gigantic scientific efforts of today, covering basic fusion plasma science and laboratory fusion experiments as well as geocosmophysical and astrophysical projects. The description is interspersed with passages which suggest relationships between science and art or poetry. The author also relates meetings which he has had with famous scientists like Niels Bohr, Hannes Alfven, Piotr Kapitza, Subrahmanyan Chandrasekhar, Lyman Spitzer and others.

  6. Fusion plasma physics during half a century

    International Nuclear Information System (INIS)

    Lehnert, Bo

    1999-08-01

    A review is given on the potentialities of fusion energy with respect to energy production and related environmental problems, the various approaches to controlled thermonuclear fusion, the main problem areas of research, the historical development, the present state of investigations, and future perspectives. This article also presents a personal memorandum of the author. Thereby special reference will be given to part of the research conducted at the Royal Institute of Technology in Stockholm, merely to identify its place within the general historical development. Considerable progress has been made in fusion research during the last decades. In large tokamak experiments temperatures above the ignition limit of about 10 8 K have been reached under break-even conditions where the fusion power generation is comparable to the energy loss. A power producing fusion reactor could in principle be realized already today, but it would not become technically and economically efficient. The future international research programme has therefore to be conducted along broad lines, with necessary ingredients of basis research and new ideas, and also within lines of magnetic confinement being alternative to that of tokamaks

  7. Fusion plasma physics during half a century

    Energy Technology Data Exchange (ETDEWEB)

    Lehnert, Bo

    1999-08-01

    A review is given on the potentialities of fusion energy with respect to energy production and related environmental problems, the various approaches to controlled thermonuclear fusion, the main problem areas of research, the historical development, the present state of investigations, and future perspectives. This article also presents a personal memorandum of the author. Thereby special reference will be given to part of the research conducted at the Royal Institute of Technology in Stockholm, merely to identify its place within the general historical development. Considerable progress has been made in fusion research during the last decades. In large tokamak experiments temperatures above the ignition limit of about 10{sup 8} K have been reached under break-even conditions where the fusion power generation is comparable to the energy loss. A power producing fusion reactor could in principle be realized already today, but it would not become technically and economically efficient. The future international research programme has therefore to be conducted along broad lines, with necessary ingredients of basis research and new ideas, and also within lines of magnetic confinement being alternative to that of tokamaks.

  8. TGV, hutě a tokamak ITER

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    Leden (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * TGV * Pulse Power Electrical Network * Steady State Electrical Network 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/1954-tgv-hute-a-tokamak-iter

  9. Tomography of laser fusion plasmas

    International Nuclear Information System (INIS)

    Ceglio, N.M.

    1977-01-01

    Experimental programs exist in a number of laboratories throughout the world to test the feasibility of using powerful laser systems to drive the implosion of hydrogen isotope fuel to thermonuclear burn conditions. In a typical experiment multiple laser beams are focused onto a glass microshell (typically 50 μm to 200 μm diameter) filled with an equimolar D-T gas mixture. X-ray and particle emissions from the target provide important information about the hydrodynamic implosion of the glass shell and the associated compression and heating of the D-T fuel. Standard diagnostics for imaging such emissions are the grazing incidence reflection (GIR) x-ray microscope and the pinhole camera. Recently, a particular coded imaging technique, Zone Plate Coded Imaging (ZPCI), has been successfully used for x-ray and particle microscopy of laser fusion plasmas. ZPCI is highly attractive for investigating laser produced plasmas because it possesses a tomographic capability not shared by either the GIR or pinhole imaging techniques. This presentation provides a brief discussion of the tomographic potential of ZPCI. In addition, the first tomographic x-ray images (tomographic resolution approximately 74 μm) of a laser produced plasma are presented

  10. Tokamak D-T neutron source models for different plasma physics confinement modes

    Energy Technology Data Exchange (ETDEWEB)

    Fausser, Clement, E-mail: clement.fausser@cea.fr [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Puma, Antonella Li; Gabriel, Franck [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer HCLL DEMO neutronics is based on plasma physics L-mode, but may use H or A mode. Black-Right-Pointing-Pointer Based on Plasma Physics 0D code, H and A-mode D-T neutron sources formulae are proposed. Black-Right-Pointing-Pointer TRANSGEN code is built to create 2D source maps as input for Monte-Carlo codes. Black-Right-Pointing-Pointer A-mode neutronic impact is compared to L-mode at same power on a HCLL DEMO design. Black-Right-Pointing-Pointer Results show TBR and Me slight changes, contrary to NWL profile: from -22% to +11%. - Abstract: Neutronic studies of European demonstration fusion power plant (DEMO) have been so far based on plasma physics low confinement mode (L-mode). Future tokamaks, nevertheless, may likely use alternative confinement modes such as high or advanced confinement modes (H and A-mode). Based on analytical formulae used in plasma physics, H and A-modes D-T neutron sources formulae are proposed in this paper. For that purpose, a tokamak random neutron source generator, TRANSGEN, has been built generating bidimensional (radial and poloidal) neutron source maps to be used as input for neutronics Monte-Carlo codes (TRIPOLI-4 and MCNP5). The impact of such a source on the neutronic behavior of the European DEMO-2007 Helium-cooled lithium-lead reactor concept has been assessed and compared with previous results obtained using a L-mode neutron source. An A-mode neutron source map from TRANSGEN has been used with the code TRIPOLI-4. Assuming the same fusion power, results show that main reactor global neutronic parameters, e.g. tritium breeding ratio and neutron multiplication factor, evolved slightly when compared to present uncertainties margin. However, local parameters, such as the neutron wall loading (NWL), change significantly compared to L-mode shape: from -22% to +11% for NWL.

  11. Plasma physics an introduction to laboratory, space, and fusion plasmas

    CERN Document Server

    Piel, Alexander

    2017-01-01

    The enlarged new edition of this textbook provides a comprehensive introduction to the basic processes in plasmas and demonstrates that the same fundamental concepts describe cold gas-discharge plasmas, space plasmas, and hot fusion plasmas. Starting from particle drifts in magnetic fields, the principles of magnetic confinement fusion are explained and compared with laser fusion. Collective processes are discussed in terms of plasma waves and instabilities. The concepts of plasma description by magnetohydrodynamics, kinetic theory, and particle simulation are stepwise introduced. Space charge effects in sheath regions, double layers and plasma diodes are given the necessary attention. The novel fundamental mechanisms of dusty plasmas are explored and integrated into the framework of conventional plasmas. The book concludes with a concise description of modern plasma discharges. Written by an internationally renowned researcher in experimental plasma physics, the text keeps the mathematical apparatus simple a...

  12. Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

    Energy Technology Data Exchange (ETDEWEB)

    S.C. Jardin; C.E. Kessel; T.K. Mau; R.L. Miller; F. Najmabadi; V.S. Chan; M.S. Chu; R. LaHaye; L.L. Lao; T.W. Petrie; P. Politzer; H.E. St. John; P. Snyder; G.M. Staebler; A.D. Turnbull; W.P. West

    2003-10-07

    The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads to a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.

  13. Tokamak plasma self-organization and the possibility to have the peaked density profile in ITER

    NARCIS (Netherlands)

    Razumova, K. A.; Andreev, V. F.; Kislov, A. Y.; Kirneva, N. A.; Lysenko, S. E.; Pavlov, Y. D.; Shafranov, T. V.; Donne, A. J. H.; Hogeweij, G. M. D.; Spakman, G. W.; R. Jaspers,; Kantor, M.; Walsh, M.

    2009-01-01

    The self-organization of a tokamak plasma is a fundamental turbulent plasma phenomenon, which leads to the formation of a self-consistent pressure profile. This phenomenon has been investigated in several tokamaks with different methods of heating. It is shown that the normalized pressure profile

  14. 2001 activity report of the development and research line in controlled thermonuclear fusion of the Plasma Associated Laboratory; Relatorio de atividades de 2001 da linha de pesquisa e desenvolvimento em fusao termonuclear controlada (fusao), do Laboratorio Associado de Plasma (LAP)

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto

    2002-07-01

    The year 2001 activities of the controlled thermonuclear fusion research line of the Plasma Associated Laboratory at the National Institute for Space Research - Brazil are reported. The report approaches the staff, participation in congresses, goals for the year 2002 and papers on Tokamak plasmas, plasma diagnostic, bootstraps, plasma equilibrium and diagnostic.

  15. Response of a tokamak plasma to particle and momentum sources

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Sigmar, D.J.

    1978-12-01

    The response of an axisymmetric toroidal tokamak plasma to first-order particle and momentum sources is investigated. The momentum sources drive coupled poloidal and toroidal mass flows and electrostatic field evolution which relax to asymptotic values on a time scale that is characteristic of the dominant viscous or external drag mechanism. The asymptotic steady-state momentum balance provides the necessary condition to completely determine the particle fluxes and currents in the flux surfaces, and, hence, to determine transport fluxes across flux surfaces. Transport fluxes are driven across flux surfaces both by interspecies collisional momentum exchange, the usual case, and by momentum exchange between the plasma and external sources and/or drags. A generalized Ohm's law is obtained and used to determine the manner in which particle and momentum sources can drive parallel currents and can alter the evolution of the q-profile. The theory is formulated for arbitrary plasma cross sections, beta, and collision regimes

  16. Neoclassical Physics for Current Drive in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Duthoit, F.X.

    2012-03-01

    The Lie transform formalism is applied to charged particle dynamics in tokamak magnetic topologies, in order to build a Fokker-Planck type operator for Coulomb collisions usable for current drive. This approach makes it possible to reduce the problem to three dimensions (two in velocity space, one in real space) while keeping the wealth of phase-space cross-term coupling effects resulting from conservation of the toroidal canonical momentum (axisymmetry). This kinetic approach makes it possible to describe physical phenomena related to the presence of strong pressure gradients in plasmas of an unspecified form, like the bootstrap current which role will be paramount for the future ITER machine. The choice of coordinates and the method used are particularly adapted to the numerical resolution of the drift kinetic equation making it possible to calculate the particle distributions, which may present a strong variation with respect to the Maxwellian under the effect of an electric field (static or produced by a radio-frequency wave). This work, mainly dedicated to plasma physics of tokamaks, was extended to those of space plasmas with a magnetic dipole configuration. (author)

  17. Self-consistent treatment of transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Wilhelmsson, H.

    1993-01-01

    A theory is developed for the dynamics of tokamak plasmas considering the influence of combinations of simultaneous heating processes (alpha particle, auxiliary and ohmic), thermal conduction and particle diffusion, thermal and particle pinches, thermalization of alpha particles as well as the effects of boundary conditions. The analysis is based on a generalization of the central expansion technique which transforms the partial differential equations to a set of nonlinear coupled equations in time for the dynamic variables. Oscillatory solutions are found, but only in the presence of alpha particle heating. Examples of extensive computer simulations are included which support and complete the analytic results. (26 refs.)

  18. Magnetohydrodynamic stability of tokamak plasmas with poloidal mode coupling

    International Nuclear Information System (INIS)

    Shigueoka, H.; Sakanaka, P.H.

    1987-01-01

    The stability behavior with respect to internal modes is examined for a class of tokamak equilibria with non-circular cross sections. The surfaces of the constant poloidal magnetic flux ψ (R,Z) are obtained numerically by solving the Grad-Shafranov's equation with a specified shape for the outmost plasma surface. The equation of motion for ideal MHD stability is written in a ortogonal coordinate system (ψ, χ, φ). Th e stability analysis is performance numerically in a truncated set of coupled m (poloidal wave number) equations. The calculations involve no approximations, and so all parameters of the equilibrium solution can be arbitrarily varied. (author) [pt

  19. Information content of transient synchrotron radiation in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1989-04-01

    A brief, deliberate, perturbation of hot tokamak electrons produces a transient, synchrotron radiation signal, in frequency-time space, with impressive informative potential on plasma parameters; for example, the dc toroidal electric field, not available by other means, may be measurably. Very fast algorithms have been developed, making tractable a statistical analysis that compares essentially all parameter sets that might possibly explain the transient signal. By simulating data numerically, we can estimate the informative worth of data prior to obtaining it. 20 refs., 2 figs

  20. Real-Time Software for the Compass Tokamak Plasma Control

    Energy Technology Data Exchange (ETDEWEB)

    Valcarcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J. [Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Lisboa (Portugal); Sartori, F. [Euratom-UKAEA, Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Janky, F.; Cahyna, P.; Hron, M.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM / IPP.CR, Prague (Costa Rica)

    2009-07-01

    This poster presents the flexible and high-performance real time system that guarantees the desired time cycles for plasma control on the COMPASS tokamak: 500 {mu}s for toroidal field, current, equilibrium and shaping; 50 {mu}s for fast control of the equilibrium and vertical instability. This system was developed on top of a high-performance processor and a software framework (MARTe) tailored for real-time. The preliminary measurements indicate that the time constraints will be met on the final solution. The system allows the making of modifications in the future to improve software components. (A.C.)

  1. Ion thermal conductivity for a pure tokamak plasma

    International Nuclear Information System (INIS)

    Bolton, C.; Ware, A.

    1981-10-01

    The neoclassical and Pfirsch-Schlueter component of the ion heat conduction for a pure hydrogen Tokamak plasma are recalculated without assuming large aspect ratio and without neglecting energy scattering collisions. Using a model collision operator the conductivity is determined numerically for various collision frequencies and aspect ratios. An approximate algebraic expression is fitted to the results. Even for comparatively large aspect ratios(approx. 10) energy scattering increases the conductivity in the banana regime by about 50% and for small aspect ratios(approx. 3.3) the Pfirsch-Schlueter contribution causes a further increase of approximately 60%

  2. Study of edge turbulence in tokamak plasmas

    International Nuclear Information System (INIS)

    Sarazin, Y.

    1997-01-01

    The aim of this work is to propose a new frame to study turbulent transport in plasmas. In order to avoid the restraint of scale separability the forcing by flux is used. A critical one-dimension self-organized cellular model is developed. In keeping with experience the average transport can be described by means of diffusion and convection terms whereas the local transport could not. The instability due to interchanging process is thoroughly studied and some simplified equations are derived. The proposed model agrees with the following experimental results: the relative fluctuations of density are maximized on the edge, the profile shows an exponential behaviour and the amplitude of density fluctuations depends on ionization source strongly. (A.C.)

  3. Plasma physics and nuclear fusion research

    CERN Document Server

    Gill, Richard D

    1981-01-01

    Plasma Physics and Nuclear Fusion Research covers the theoretical and experimental aspects of plasma physics and nuclear fusion. The book starts by providing an overview and survey of plasma physics; the theory of the electrodynamics of deformable media and magnetohydrodynamics; and the particle orbit theory. The text also describes the plasma waves; the kinetic theory; the transport theory; and the MHD stability theory. Advanced theories such as microinstabilities, plasma turbulence, anomalous transport theory, and nonlinear laser plasma interaction theory are also considered. The book furthe

  4. Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

    Energy Technology Data Exchange (ETDEWEB)

    Mitrishkin, Yuri V., E-mail: y_mitrishkin@hotmail.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: janerigoler@mail.ru [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: ea.kuznetsov@mail.ru [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: k.gaydamaka@gmail.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)

    2016-10-15

    Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

  5. Shielding study of a fusion machine. Elaboration of a global shielding calculation scheme for the Tokamak tore Supra

    International Nuclear Information System (INIS)

    Diop, C.M'B.

    1984-01-01

    This thesis presents a global shielding calculation scheme for neutron and gamma rays arising from the Tokamak TORE SUPRA fusion device, in which a deuterium plasma is used. To study the shield parameters we have elabored a important chaining of neutron and gamma transport codes, TRIPOLI, ANISN, MERCURE 4, allowing to evaluate the radial and skyshine components of the dose rate behind the concrete shield. The study of thermonuclear neutron activation is fundamental to define a tokamak exploitation strategy. For this, two formalisme have been developed. They are based on a modelization of the activation reaction rates according to TRIPOLI, ANISN, and MERCURE 4 codes capabilities. The first one calculates, in one dimensional geometry, the desactivation gamma dose rate inside the vacuum chamber. The second one is a tridimensional model which determines the spatial variation of the gamma dose rate in the machine room. The problem of the existence of runaway electrons and associated secondaries radiations, bremsstrahlung gamma rays particularly, is approched. The results which are presented have contributed to define the parameters of the concrete shield and a strategy for TORE SUPRA Tokamak exploitation [fr

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

    International Nuclear Information System (INIS)

    Jeronimo, Leonardo Cunha

    2013-01-01

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

  7. Real-time digital control of plasma position and shape on the TEXTOR tokamak

    International Nuclear Information System (INIS)

    Mitri, Mikhael

    2009-01-01

    Beside the objective of contributing to the controlled thermonuclear fusion research and ultimately the development of a fusion based power plant, the main objectives of the thesis are a substantial improvement of plasma vertical position control and plasma shape control as well as a better understanding of formerly unexplained effects, e.g. disturbance fields. As for the vertical position control, a deep analysis has to be undertaken to identify the problem sources. Accurate control of the plasma position is very difficult to achieve. This is mainly due to the complexity of the tokamak and the difficulty in measuring or modelling all relevant discharge variables. Any models would be highly nonlinear and time varying. Thus, for simulation and controller design, a simplified, but nevertheless accurate model has to be developed, based on physics and measured data of the process. Furthermore, the quality of the measured position has to be improved by using new inductive sensors, integrators, and hardware. The integration drift problem has to be analysed and resolved by developing a drift-free integration method. Concerning the shape control, a better understanding of the relation between the stray fields and the iron core saturation is required. Furthermore, the influence on the plasma elongation has to be determined. Upon this, a shape compensation algorithm has to be developed accordingly. The accuracy of the shape control has to be better than 1%. (orig.)

  8. 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

  9. Full Tokamak discharge simulation and kinetic plasma profile control for ITER

    International Nuclear Information System (INIS)

    Hee Kim, S.

    2009-10-01

    transitions were not fully achievable due to a vertical displacement event (VDE) caused by a strong inward plasma movement. In the part dedicated to full tokamak discharge simulations, firstly, we have introduced the combined DINA-CH/CRONOS tokamak discharge simulator. DINA-CH self-consistently calculates the non-linear evolution of the free-boundary plasma equilibrium with the plasma current diffusion, in response to both controlled poloidal field (PF) coil currents and inductively driven currents in the surrounding conducting system. CRONOS provides the evolution of the plasma profiles by self-consistently solving heat and particle transport with source profiles. Secondly, we have successfully simulated ITER operation scenario 2 as a demonstration of the capabilities of the combined simulator, as well as being a design study in itself. The fusion power ratio to the total auxiliary power Q was about 10 with the application of 53 MW of auxiliary heating and current drive (H and CD) power. We have investigated several specific issues related to the tokamak operation, such as the vertical instability, PF coil current limits and poloidal flux consumption during the current ramp-up. Lower hybrid (LH) applied from the initial phase of the plasma current ramp-up increased the safety margins in operating the superconducting PF coils both by reducing resistive ohmic flux consumption and by providing non-inductively driven plasma current. Lastly, we have studied ITER hybrid mode operation, focusing on the operational capability of obtaining a stationary flat safety factor (q) profile at the start of at-top (SOF) phase and sustaining it as long as possible by combining various non-inductively driven current sources. Application of a near on-axis electron cyclotron current drive (ECCD) appears to be effective compared to the far off-axis lower hybrid current drive (LHCD), at least on short time scales. In the active plasma profile control part, we have developed a robust control

  10. Fusion-product ash buildup in tokamak with radial electric field

    International Nuclear Information System (INIS)

    Downum, W.B.; Choi, C.K.; Miley, G.H.

    1979-01-01

    The buildup of thermalized fusion products (ash) in a tokamak can seriously limit burn times. Prior studies have concentrated on deposition profile effects on alpha particle transport in tokamaks but have not considered the effect on ash of radial electric fields (either created internally, e.g. due to high-energy alpha leakage, or generated externally). The present study focuses on this issue since it appears that electric fields might offer one approach to control of the ash. Approximate field and source profiles are used, based on prior calculations

  11. The role of alpha particles in magnetically confined fusion plasmas

    International Nuclear Information System (INIS)

    Lisak, M.; Wilhelmsson, H.

    1986-01-01

    Recent progress in the confinement of hot plasmas in magnetic fusion experiments throughout the world has intensified interest and research in the physics of D-T burning plasmas especially in the wide range of unresolved theoretical as well as experimental questions associated with the role of alpha particles in such devices. In order to review the state-of-the- art in this field, and to identify new issues and problems for further research, the Symposium on the Role of Alpha Particles in Magnetically Confined Fusion Plasmas was held from 24 to 26 June 1986 at Aspenaesgaarden near Goeteborg, Sweden. About 25 leading experts from nine countries attended the Symposium and gave invited talks. The major part of the programme was devoted to alpha-particle effects in tokamaks but some aspects of open systems were also discussed. The possibilities of obtaining ignition in JET and TFTR as well as physics issues for the compact ignition experiments were considered in particular. A special session was devoted to the diagnostics of alpha particles and other fusion products. In this report are summarised some of the highlights of the symposium. (authors)

  12. Electron and current density measurements on tokamak plasmas

    International Nuclear Information System (INIS)

    Lammeren, A.C.A.P. van.

    1991-01-01

    The first part of this thesis describes the Thomson-scattering diagnostic as it was present at the TORTUR tokamak. For the first time with this diagnostic a complete tangential scattering spectrum was recorded during one single laser pulse. From this scattering spectrum the local current density was derived. Small deviations from the expected gaussian scattering spectrum were observed indicating the non-Maxwellian character of the electron-velocity distribution. The second part of this thesis describes the multi-channel interferometer/ polarimeter diagnostic which was constructed, build and operated on the Rijnhuizen Tokamak Project (RTP) tokamak. The diagnostic was operated routinely, yielding the development of the density profiles for every discharge. When ECRH (Electron Cyclotron Resonance Heating) is switched on the density profile broadens, the central density decreases and the total density increases, the opposite takes place when ECRH is switched off. The influence of MHD (magnetohydrodynamics) activity on the density was clearly observable. In the central region of the plasma it was measured that in hydrogen discharges the so-called sawtooth collapse is preceded by an m=1 instability which grows rapidly. An increase in radius of this m=1 mode of 1.5 cm just before the crash is observed. In hydrogen discharges the sawtooth induced density pulse shows an asymmetry for the high- and low-field side propagation. This asymmetry disappeared for helium discharges. From the location of the maximum density variations during an m=2 mode the position of the q=2 surface is derived. The density profiles are measured during the energy quench phase of a plasma disruption. A fast flattening and broadening of the density profile is observed. (author). 95 refs.; 66 figs.; 7 tabs

  13. New D3-D tokamak plasma control system

    Science.gov (United States)

    Campbell, G. L.; Ferron, J. R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C. M.; Pinsker, R. I.; Lazarus, E. A.

    1992-09-01

    A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and divertor power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the U.S. Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter.

  14. Annual report of the Division of Thermonuclear Fusion Research and the Division of Large Tokamak Development for the period of April 1, 1977 to March 31, 1978

    International Nuclear Information System (INIS)

    1979-02-01

    Research and development works in fiscal year 1977 of the Division of Thermonuclear Fusion Research and the Division of Large Tokamak Development are described. 1) Theoretical studies on tokamak confinement have continued with more emphasis on computations. A task was started of developing a computer code system for mhd behavior of tokamak plasmas. 2) Experimental studies of lower hybrid heating up to 140 kW were made in JFT-2. The ion temperature was increased by 50% -- 60% near the plasma center. Plasma-wall interactions (particle and thermal fluxes to the wall, and titanium gettering) were studied. In JFT-2a (DIVA) ion sputtering, arcing and evaporation were identified, and the impurity ion sputtering was found to be a dominant origin of metal impurities in the present tokamaks. High temperature and high-density plasma divertor actions were demonstrated; i.e. the divertor decreases the radiation power loss by a factor of 3 and increases the energy confinement time by a factor of 2.5. Various diagnostic instruments operated sufficiently to provide useful information for the research with JFT-2 and JFT-2a(DIVA). 3) JFT-2 and JFT-2a(DIVA) operated as scheduled. Technological improvements were made such as titanium coating of the chamber wall, discharge cleaning and pre-ionization. 4) Detailed design of the prototype JT-60 neutral beam injector was made. A 200 kW, 650 MHz radiofrequency heating system for JFT-2 was completed; a lower hybrid heating experiment in JFT-2 was successful 5) In particle-surface interactions, the sputtering and surface erosion were studied. 6) Improvement designs of a superconducting cluster test facility and a test module coil were made in the toroidal coil development. 7) Second preliminary design of the tokamak experimental fusion reactor JXFR started in April 1977. Safety analyses were made of the main components and system of JXFR on the basis of the first preliminary design. (J.P.N.)

  15. Transport analysis of high radiation and high density plasmas in the ASDEX Upgrade tokamak

    Directory of Open Access Journals (Sweden)

    Casali L.

    2014-01-01

    Full Text Available Future fusion reactors, foreseen in the “European road map” such as DEMO, will operate under more demanding conditions compared to present devices. They will require high divertor and core radiation by impurity seeding to reduce heat loads on divertor target plates. In addition, DEMO will have to work at high core densities to reach adequate fusion performance. The performance of fusion reactors depends on three essential parameters: temperature, density and energy confinement time. The latter characterizes the loss rate due to both radiation and transport processes. The DEMO foreseen scenarios described above were not investigated so far, but are now addressed at the ASDEX Upgrade tokamak. In this work we present the transport analysis of such scenarios. Plasma with high radiation by impurity seeding: transport analysis taking into account the radiation distribution shows no change in transport during impurity seeding. The observed confinement improvement is an effect of higher pedestal temperatures which extend to the core via stiffness. A non coronal radiation model was developed and compared to the bolometric measurements in order to provide a reliable radiation profile for transport calculations. High density plasmas with pellets: the analysis of kinetic profiles reveals a transient phase at the start of the pellet fuelling due to a slower density build up compared to the temperature decrease. The low particle diffusion can explain the confinement behaviour.

  16. Development of burning plasma and advanced scenarios in the DIII-D tokamak

    International Nuclear Information System (INIS)

    Luce, T.C.

    2005-01-01

    Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q ∼ 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque. (author)

  17. Fully non-inductive second harmonic electron cyclotron plasma ramp-up in the QUEST spherical tokamak

    Science.gov (United States)

    Idei, H.; Kariya, T.; Imai, T.; Mishra, K.; Onchi, T.; Watanabe, O.; Zushi, H.; Hanada, K.; Qian, J.; Ejiri, A.; Alam, M. M.; Nakamura, K.; Fujisawa, A.; Nagashima, Y.; Hasegawa, M.; Matsuoka, K.; Fukuyama, A.; Kubo, S.; Shimozuma, T.; Yoshikawa, M.; Sakamoto, M.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Ide, S.; Maekawa, T.; Takase, Y.; Toi, K.

    2017-12-01

    Fully non-inductive second (2nd) harmonic electron cyclotron (EC) plasma current ramp-up was demonstrated with a newlly developed 28 GHz system in the QUEST spherical tokamak. A high plasma current of 54 kA was non-inductively ramped up and sustained stably for 0.9 s with a 270 kW 28 GHz wave. A higher plasma current of 66 kA was also non-inductively achieved with a slow ramp-up of the vertical field. We have achieved a significantly higher plasma current than those achieved previously with the 2nd harmonic EC waves. This fully non-inductive 2nd harmonic EC plasma ramp-up method might be useful for future burning plasma devices and fusion reactors, in particular for operations at half magnetic field with the same EC heating equipment.

  18. 2D full-wave simulation of waves in space and tokamak plasmas

    Directory of Open Access Journals (Sweden)

    Kim Eun-Hwa

    2017-01-01

    Full Text Available Simulation results using a 2D full-wave code (FW2D for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF waves in the scape-off layer (SOL of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

  19. 2D full-wave simulation of waves in space and tokamak plasmas

    Science.gov (United States)

    Kim, Eun-Hwa; Bertelli, Nicola; Johnson, Jay; Valeo, Ernest; Hosea, Joel

    2017-10-01

    Simulation results using a 2D full-wave code (FW2D) for space and NSTX fusion plasmas are presented. The FW2D code solves the cold plasma wave equations using the finite element method. The wave code has been successfully applied to describe low frequency waves in planetary magnetospheres (i.e., dipole geometry) and the results include generation and propagation of externally driven ultra-low frequency waves via mode conversion at Mercury and mode coupling, refraction and reflection of internally driven field-aligned propagating left-handed electromagnetic ion cyclotron (EMIC) waves at Earth. In this paper, global structure of linearly polarized EMIC waves is examined and the result shows such resonant wave modes can be localized near the equatorial plane. We also adopt the FW2D code to tokamak geometry and examine radio frequency (RF) waves in the scape-off layer (SOL) of tokamaks. By adopting the rectangular and limiter boundary, we compare the results with existing AORSA simulations. The FW2D code results for the high harmonic fast wave heating case on NSTX with a rectangular vessel boundary shows excellent agreement with the AORSA code.

  20. Modelling the turbulent transport of angular momentum in tokamak plasmas - A quasi-linear gyrokinetic approach

    International Nuclear Information System (INIS)

    Cottier, Pierre

    2013-01-01

    The magnetic confinement in tokamaks is for now the most advanced way towards energy production by nuclear fusion. Both theoretical and experimental studies showed that rotation generation can increase its performance by reducing the turbulent transport in tokamak plasmas. The rotation influence on the heat and particle fluxes is studied along with the angular momentum transport with the quasi-linear gyro-kinetic eigenvalue code QuaLiKiz. For this purpose, the QuaLiKiz code is modified in order to take the plasma rotation into account and compute the angular momentum flux. It is shown that QuaLiKiz framework is able to correctly predict the angular momentum flux including the E*B shear induced residual stress as well as the influence of rotation on the heat and particle fluxes. The major approximations of QuaLiKiz formalisms are reviewed, in particular the ballooning representation at its lowest order and the eigenfunctions calculated in the hydrodynamic limit. The construction of the quasi-linear fluxes is also reviewed in details and the quasi-linear angular momentum flux is derived. The different contributions to the turbulent momentum flux are studied and successfully compared both against non-linear gyro-kinetic simulations and experimental data. (author) [fr

  1. Anomalous energy transport in hot plasmas: solar corona and Tokamak

    International Nuclear Information System (INIS)

    Beaufume, P.

    1992-04-01

    Anomalous energy transport is studied in two hot plasmas and appears to be associated with a heating of the solar corona and with a plasma deconfining process in tokamaks. The magnetic structure is shown to play a fundamental role in this phenomenon through small scale instabilities which are modelized by means of a nonlinear dynamical system: the Beasts' Model. Four behavior classes are found for this system, which are automatically classified in the parameter space thanks to a neural network. We use a compilation of experimental results relative to the solar corona to discuss current-based heating processes. We find that a simple Joule effect cannot provide the required heating rates, and therefore propose a dimensional model involving a resistive reconnective instability which leads to an efficient and discontinuous heating mechanism. Results are in good agreement with the observations. We give an analytical expression for a diffusion coefficient in tokamaks when magnetic turbulence is perturbing the topology, which we validate thanks to the standard mapping. A realistic version of the Beasts' Model allows to test a candidate to anomalous transport: the thermal filamentation instability

  2. End points in discharge cleaning on TFTR (Tokamak Fusion Test Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, D.; Dylla, H.F.; Bell, M.G.; Blanchard, W.R.; Bush, C.E.; Gettelfinger, G.; Hawryluk, R.J.; Hill, K.W.; Janos, A.C.; Jobes, F.C.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) /approx/ 3 at /approx/ 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H/sub 2/O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after /approx/ 10/sup 5/ pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150/degree/C to about 250/degree/C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H/sub 2/O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab.

  3. End points in discharge cleaning on TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Mueller, D.; Dylla, H.F.; Bell, M.G.

    1989-07-01

    It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) approx 3 at approx 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H 2 O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after approx 10 5 pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150 degree C to about 250 degree C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H 2 O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab

  4. Fusion materials studies and technological experiments in test thermonuclear reactor-tokamak

    International Nuclear Information System (INIS)

    Altovskij, I.V.; Gornostaev, B.D.; Blinov, I.A.; Gur'ev, V.V.; Zujkov, G.A.; Osadchij, N.A.; Tochenyj, L.V.

    1987-01-01

    Extensive studies of materials (strength, radiation resistance, reliability and safety) should be made to substantiate designs during the initial stage of R and D and the choice of the final version for the design of the blanket and energy conversion system. It is evident that those experimental devices which are available or planned are not capable of simulating effects on the OTR blanket units (high neutron flux, strong electromagnetic fields, stresses due to coolant pressure at operating temperature and own weight, plasma effects, etc.). Therefore, in OTR design, constructive decisions should be taken on the use of the tokamak OTR as a unique experimental device for real full-scale tests of construction, uranium, lithium and protective materials (antidiffusion coatings for radiation safety, etc.) intended for long term reliable and safe operation in future commercial fusion reactors. In the paper, consideration is given to some experimental devices intended for the study of the physical and mechanical properties of materials under loads typical of blanket operation; research into the mutual compatibility of construction materials, with coolant and operating materials; the study of the efficiency and reliability of material coatings; and the determination of optimal operating temperature ranges for materials, etc. (author). 2 refs, 5 figs

  5. Measurements of energy and flux of neutrals at the wall in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Wampler, W.R.; Kilpatrick, S.J.

    1989-01-01

    The energy and flux of charge-exchange neutral deuterium incident on the wall in the Tokamak Fusion Test Reactor (TFTR) were measured using a carbon resistance probe. This device utilizes the increase in electrical resistance of a thin carbon film produced by the impinging particles. Particle energies and fluxes were determined from the stopping effect of thin layers of silicon over the carbon. Film thicknesses were chosen to give a useful energy range from about 50 to 5000 eV. Time-resolved data were obtained for the first time by using a transient digitizer to record the signals every millisecond during the discharge. The deduced flux and Maxwellian temperature were 0.7 x 10 14 /cm 2 s and kT = 350 eV, respectively, during Ohmic heating. Neutral beam heating (10 MW) increased the temperature by a factor of 4 and increased the flux by a factor of 2. The observed flux of neutrals to the wall is large enough for changes in the recycling of these particles to affect plasma fueling significantly

  6. Design study of superconducting magnets for tokamak experimental fusion reactor, (1)

    International Nuclear Information System (INIS)

    1980-03-01

    Design study has been made of superconducting magnets for a Tokamak experimental fusion reactor: toroidal field magnet design, poloidal field magnet design, refrigeration system design, magnet safety analysis, and magnet assembling and disassembling system design. A maximum toroidal field in the coil is 11.0 T, providing 5.5 T at plasma center. Nb 3 Sn superconducting cable is used to attain the toroidal field of 11 T. The coil bore is 7.3 x 11.2 m, and the coil shape is deformed constant-tension D-shape. The magnetomotive force is 185.6 MAT, and the operational current is 25.9 kA. In poloidal field magnet design, the coil is pancake-wound Nb 3 Sn conductor. The conductor is enclosed in Ti-alloy sheath, which serves also as helium containment vessel. The conductor is cooled by forced flow supercritical helium of 7 atm and 4.6 K, and the operational current is 25 -- 27 kA. (author)

  7. The Fusion Science Research Plan for the Major U.S. Tokamaks. Advisory report

    International Nuclear Information System (INIS)

    1996-01-01

    In summary, the community has developed a research plan for the major tokamak facilities that will produce impressive scientific benefits over the next two years. The plan is well aligned with the new mission and goals of the restructured fusion energy sciences program recommended by FEAC. Budget increases for all three facilities will allow their programs to move forward in FY 1997, increasing their rate of scientific progress. With a shutdown deadline now established, the TFTR will forego all but a few critical upgrades and maximize operation to achieve a set of high-priority scientific objectives with deuterium-tritium plasmas. The DIII-D and Alcator C-Mod facilities will still fall well short of full utilization. Increasing the run time in vii DIII-D is recommended to increase the scientific output using its existing capabilities, even if scheduled upgrades must be further delayed. An increase in the Alcator C-Mod budget is recommended, at the expense of equal and modest reductions (~1%) in the other two facilities if necessary, to develop its capabilities for the long-term and increase its near-term scientific output.

  8. Comparative study of cost models for tokamak DEMO fusion reactors

    International Nuclear Information System (INIS)

    Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Ban, Kanae; Kondo, Takuya; Tobita, Kenji; Goto, Takuya

    2012-01-01

    Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study. (author)

  9. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Steady-state tokamak reactors: SSTR and A-SSTR

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru

    1998-01-01

    The SSTR is a steady-state tokamak reactor based on a bootstrap current with near term technologies. The A-SSTR is an aggressive fusion reactor concept designed to address issues of economical competitiveness. (author)

  10. Heat flow during sawtooth collapse in tokamak plasmas

    International Nuclear Information System (INIS)

    Hanada, Kazuaki

    1994-01-01

    Heat flow during sawtooth collapse was studied on the WT-3 tokamak by using temporal evolution of soft X-ray intensity profile in the poloidal cross section in a lower hybrid current driven plasma as well as an electron cyclotron heated plasma. Two phase in sawtooth collapses were observed. In the first phases, the hottest spot that is the peak of the soft X-ray distribution approaches the inversion surface and heat flows out through a narrow gate on the inversion surface. In the second phase, the hottest spot stays on the inversion surface, and heat flows out through the whole inversion surface. This suggests that magnetic reconnection as predicted by Kadomtsev's model occurs in the first phase, but in the second phase, a different mechanism dominates heat flow. (author)

  11. Electron temperature gradient driven instability in the tokamak boundary plasma

    Energy Technology Data Exchange (ETDEWEB)

    Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.

    1992-12-15

    A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t{sup {minus}1/2}e{sup {gamma}mt}.

  12. Plasma-neutral gas interaction in a tokamak divertor: effects of hydrogen molecules and plasma recombination

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Pigarov, A.Yu.; Soboleva, T.K.; Sigmar, D.J.

    1997-01-01

    We investigate the influence of hydrogen molecules on plasma recombination using a collisional-radiative model for multispecies hydrogen plasmas and tokamak detached divertor parameters. The rate constant found for molecular activated recombination of a plasma can be as high as 2 x 10 -10 cm 3 /s, confirming our pervious estimates. We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a 'gas box' divertor geometry. For the model of plasma-neutral interactions we employ we find: (a) molecular activated recombination is a dominant channel of divertor plasma recombination; and (b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes. (orig.)

  13. Plasma physics and controlled nuclear fusion

    International Nuclear Information System (INIS)

    Sato, Tetsuya

    1993-05-01

    The report contains the proceedings of a conference on plasma physics. A fraction of topics included MHD instabilities, magnetic confinement and plasma heating in the field of fusion plasmas, in 8 papers falling in the INIS scope have been abstracted and indexed for the INIS database. (K.A.)

  14. Massachusetts Institute of Technology Plasma Fusion Center 1987--1988 report to the President

    International Nuclear Information System (INIS)

    1988-06-01

    During the past year, technical progress has been made in all Plasma Fusion Center (PFC) research programs. The Plasma Fusion Center is recognized as one of the leading university research laboratories in the physics and engineering aspects of magnetic confinement fusion. Its research programs have produced significant results on several fronts: the basic physics of high-temperature plasmas (plasmas theory, RF heating, free electron lasers, development of advanced diagnostics, and intermediate-scale experiments on the Versator tokamak and Constance mirror devices), major confinement results on the Alcator C tokamak, including pioneering investigations of the stability, heating, and confinement properties of plasmas at high densities, temperatures and magnetic fields, experiments on the medium-scale TARA tandem mirror, including the development of novel MHD stabilization techniques in axisymmetric geometry, and a broad program of fusion technology and engineering development that addresses problems in several critical subsystem areas (e.g., magnet systems, superconducting materials development, environmental and safety studies, advanced millimeter-wave source development, and system studies of fusion reactor design, operation, and technology requirements

  15. Massachusetts Institute of Technology, Plasma Fusion Center, 1984-1985. Report to the President

    International Nuclear Information System (INIS)

    1985-07-01

    During the past year, technical progress has been made in all Plasma Fusion Center (PFC) research programs. The Plasma Fusion Center is recognized as one of the leading university research laboratories in the physics and engineering aspects of magnetic confinement fusion. Its research programs have produced significant results on four fronts: (1) the basic physics of high-temperature plasmas (plasma theory, rf heating, free electron lasers, development of advanced diagnostics and small-scale experiments on the Versator tokamak and Constance mirror devices); (2) major confinement results on the Alcator C tokamak, including pioneering investigations of the stability, heating, and confinement properties of plasmas at high densities, temperatures and magnetic fields; (3) development of an innovative design for axisymmetric tandem mirrors with inboard thermal barriers, with initial operation of the TARA tandem mirror experiment beginning in 1984; and (4) a broad program of fusion technology and engineering development that addresses problems in several critical subsystem areas (e.g., magnet systems, superconducting materials development, environmental and safety studies, advanced millimeter wave source development, and system studies of fusion reactor design, operation, and technology requirements). A review of these programs is given

  16. Probe diagnostics in the far scrape-off layer plasma of Korea Superconducting Tokamak Advanced Research tokamak using a sideband harmonic method

    International Nuclear Information System (INIS)

    Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook

    2015-01-01

    Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method

  17. Measurements of fusion product emission profiles in tokamaks

    International Nuclear Information System (INIS)

    Strachan, J.D.; Heidbrink, W.W.; Hendel, H.W.; Lovberg, J.; Murphy, T.J.; Nieschmidt, E.B.; Tait, G.D.; Zweben, S.J.

    1986-11-01

    The techniques and results of fusion product emission profile measurements are reviewed. While neutron source strength profile measurements have been attempted by several methods, neutron scattering is a limitation to the results. Profile measurements using charged fusion products have recently provided an alternative since collimation is much easier for the charged particles

  18. Impact of an integrated core/SOL description on the R and B T optimization of tokamak fusion reactors

    Science.gov (United States)

    Siccinio, M.; Fable, E.; Angioni, C.; Saarelma, S.; Scarabosio, A.; Zohm, H.

    2018-01-01

    An updated and improved version of the 0D divertor and scrape-off layer (SOL) model published in Siccinio et al (2016 Plasma Phys. Control. Fusion 58 125011) was coupled with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Pereverzev and Yushmanov 2002 IPP Report 5/98 and Fable et al 2013 Plasma Phys. Control. Fusion 55 124028). The resulting numerical tool was employed for various scans in the major radius R and in the toroidal magnetic field B T—for different safety factors q, allowable loop voltages V loop and H factors—in order to identify the most convenient choices for an electricity producing tokamak. Such a scenario analysis was carried out evaluating self-consistently, and simultaneously, the core profile and transport effects, which significantly impact on the fusion power outcome, and the divertor heat loads, which represent one of the most critical issues in view of the realization of fusion power plants (Zohm et al 2013 Nucl. Fusion 53 073019 and Wenninger et al 2017 Nucl. Fusion 57 046002). The main result is that, when divertor limits are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, and a maximum achievable power exists—i.e. no benefits would be obtained from a further increase in R and B T once the optimum is reached. This result appears as an intrinsic physical limit for all those devices where a radiative SOL is needed to deal with the power exhaust, and where a lower limit on the power crossing the separatrix (e.g. because of the L–H transition) is present.

  19. Plasma confinement using biased electrode in the TCABR tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Kuznetsov, Y.K.; Severo, J.H.F.; Fonseca, A.M.M.; Elfimov, A.; Bellintani, V.; Machida, M.; Heller, M.V.A.P.; Galvao, R.M.O.; Sanada, E.K.; Elizondo, J.I.

    2005-01-01

    Experimental data obtained on the TCABR tokamak (R = 0.61 m, a = 0.18 m) with an electrically polarized electrode, placed at r = 0.16 m, is reported in this paper. The experiment was performed with plasma current of 90 kA (q 3.1) and hydrogen gas injection adjusted for keeping the electron density at 1.0 x 10 19 m -3 without bias. Time evolution and radial profiles of plasma parameters with and without bias were measured. The comparison of the profiles shows an increase of the central line-averaged density, up to a maximum factor of 2.6, while H α hydrogen spectral line intensity decreases and the C III impurity stays on the same level. The analysis of temporal behaviour and radial profiles of plasma parameters indicates that the confined plasma enters the H-mode regime. The data analysis shows a maximum enhanced energy confinement factor of 1.95, decaying to 1.5 at the maximum of the density, in comparison with predicted Neo-Alcator scaling law values. Indications of transient increase of the density gradient near the plasma edge were obtained with measurements of density profiles. Calculations of turbulence and transport at the Scrape-Off-Layer, using measured floating potentials and ion saturation currents, show a strong decrease in the power spectra and transport. Bifurcation was not observed and the decrease in the saturation current occurs in 50 μs

  20. 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

  1. Alpha particle effects in burning tokamak plasmas: overview and specific examples

    International Nuclear Information System (INIS)

    Sigmar, D.J.

    1986-07-01

    Using the total power balance of an ignited tokamak plasma as a guideline, a range of alpha driven effects is surveyed regarding their impact on achieving and maintaining fusion burn. Specific examples of MHD and kinetic modes and multi species transport dynamics are discussed, including the possible interaction of these categories of effects. This power balance approach rather than a straightforward enumeration of possible effects serves to reveal their non-linear dependence and the ensuing fragility of our understanding of the approach to and maintenance of ignition. Specific examples are given of the interaction between α-power driven sawtoothing and ideal MHD stability, and direct α-effects on MHD modes including kinetic corrections. Anomalous ion heat transport and central impurity peaking mechanisms and anomalous and collisional α-transport including the ambipolar electric field are discussed

  2. Finite beta effects on turbulent transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Hein, Tobias

    2011-01-01

    The research on the transport properties of magnetically confined plasmas plays an essential role towards the achievement of practical nuclear fusion energy. An economically viable fusion reactor is expected to operate at high plasma pressure. This implies that the detailed study of the impact of electromagnetic effects, whose strength increases with increasing pressure, is of critical importance. In the present work, the electromagnetic effects on the particle, momentum and heat transport channels have been investigated, with both analytical and numerical calculations. Transport processes due to a finite plasma pressure have been identified, their physical mechanisms have been explained, and their contributions have been quantified, showing that they can be significant under experimentally relevant conditions.

  3. TECHNOLOGIES TO OPTIMIZE ADVANCED TOKAMAK

    Energy Technology Data Exchange (ETDEWEB)

    SIMONEN, TC

    2004-01-01

    OAK-B135 Commercial fusion power systems must operate near the limits of the engineering systems and plasma parameters. Achieving these objectives will require real time feedback control of the plasma. This paper describes plasma control systems being used in the national DIII-D advanced tokamak research program.

  4. High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

    International Nuclear Information System (INIS)

    Goodall, D.H.J.

    1982-01-01

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs. (orig.)

  5. High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goodall, D.H.J. (Euratom/UKAEA Fusion Association, Abingdon (UK). Culham Lab.)

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs.

  6. Analysis of equilibrium and topology of tokamak plasmas

    International Nuclear Information System (INIS)

    Milligen, B.P. van.

    1991-01-01

    In a tokamak, the plasma is confined by means of a magnetic field. There exists an equilibrium between outward forces due to the pressure gradient in plasma and inward forces due to the interaction between currents flowing inside the plasma and the magnetic field. The equilibrium magnetic field is characterized by helical field lines that lie on nested toroidal surfaces of constant flux. The equilibrium yields values for global and local plasma parameters (e.g. plasma position, total current, local pressure). Thus, precise knowledge of the equilibrium is essential for plasma control, for the understanding of many phenomena occurring in the plasma (in particular departures from the ideal equilibrium involving current filamentation on the flux surfaces that lead to the formation of islands, i.e. nested helical flux surfaces), and for the interpretation of many different types of measurements (e.g. the translation of line integrated electron density measurements made by laser beams probing the plasma into a local electron density on a flux surface). The problem of determining the equilibrium magnetic field from external magnetic field measurements has been studied extensively in literature. The problem is 'ill-posed', which means that the solution is unstable to small changes in the measurement data, and the solution has to be constrained in order to stabilize it. Various techniques for handling this problem have been suggested in literature. Usually ad-hoc restrictions are imposed on the equilibrium solution in order to stabilize it. More equilibrium solvers are not able to handle very dissimilar measurement data which means information on the equilibrium is lost. The generally do not allow a straightforward error estimate of the obtained results to be made, and they require large amounts of computing time. This problems are addressed in this thesis. (author). 104 refs.; 42 figs.; 6 tabs

  7. Experimental study of magnetohydrodynamic instabilities in a tokamak plasma

    International Nuclear Information System (INIS)

    Maeno, Masaki

    1981-10-01

    The JFT-2 device is a circular tokamak with a major radius of 90 cm and a shell radius of 36 cm. The maximum toroidal magnetic field and plasma current are 1.8 Wb/m 2 and 160 kA, respectively. Some devices are developed to measure the magnetohydrohynamic instability and the radiation loss. Experimental study of the magnetohydrodynamic instability is made to clear the relation between the instability and the radiation loss. The periphery cooling of the main plasma by the radiation loss plays an important role to the growth of the instability. The reduction of the radiation loss power results in the increase of the plasma density and the decrease of the safety factor q sub(a). Plasma disruption limits the increase of the plasma density and the decrease of the safety factor. In high density discharges, enhanced radiation loss power due to the increased plasma-wall interaction plays an important role in the growth of the instability. The m = 2 tearing mode seems to be the most probable cause of the disruption. In low q sub(a) (q sub(a) < 2) discharges, no inhancement of the radiation loss power was observed until disruption. The plasma disrupts at q sub(a) asymptotically equals 2. The m = 2 kink mode is the probable cause of the disruption. Attainable q sub(a) is reduced with decreasing the ratio of shell radius to plasma radius. The possibility of the shell stabilization against m = 2 kink mode is proposed. (author)

  8. DEALS: a maintainable superconducting magnet system for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Hseih, S.Y.; Danby, G.; Powell, J.R.

    1979-01-01

    The feasibility of demountable superconducting magnet systems has been examined in a design study of a DEALS [Demountable Externally Anchored Low Stress] TF magnet for an HFITR [High Field Ignition Test Reactor] Tokamak device. All parts of the system appear feasible, including the demountable superconducting joints. Measurements on small scale prototype joints indicate that movable pressure contact joints exhibit acceptable electrical, mechanical, and cryogenic performance. Such joints permit a relatively simple support structure and are readily demountable. Assembly and disassembly sequences are described whereby any failed portion of the magnet, or any part of the reactor inside the TF coils can be removed and replaced if necessary

  9. Track-mounted remote handling system for the Tokamak Fusion Engineering Test

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Daubert, R.L.; Yount, J.A.

    1982-01-01

    Concepts for remote handling machines (IVM) designed to transverse the interior of toroidal vessels with guidance and support from track systems have been developed for the proposed Tokamak Fusion Engineering Test (TFET). TFET has been proposed as an upgrade for the Tokamak Fusion Test Reactor (TFTR), currently nearing completion. The track-mounted IVMs were conceived to perform in-vessel remote maintenance for TFET, including removal and replacement of pump limiter blades and protective tiles as well as other maintenance-related tasks such as vessel wall inspection leak testing and interior cleanup. The conceptual IVMs consist of three manipulator arms mounted on a common frame member: a single power manipulator arm with high load carrying capacity and two lower-capacity servomanipulator arms. Descriptions of the IVM concepts, in-vessel track systems, and ex-vessel handling systems are presented

  10. Status of Tokamak Fusion Test Reactor neutron activation

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Rule, K. [Princeton Plasma Physics Lab., NJ (United States)

    1996-12-31

    Measurements have been made following TFTR D-T campaigns to characterize the behavior of D-T fusion reactor neutron activation using Ionization Chamber, Geiger Mueller, and Ge detector gamma-ray spectroscopy measurements. The results exhibit decay rates characteristic of the materials and geometries of the Test Cell hardware, and allow extrapolation to higher fusion power yields. The results can be used for benchmarking D-T fusion reactor activation simulations for accurate determinations of low activation long-lived cooling. 5 refs., 8 figs.

  11. Material Challenges For Plasma Facing Components in Future Fusion Reactors

    International Nuclear Information System (INIS)

    Linke, J; Pintsuk, G.; Rödig, M.

    2013-01-01

    Increasing attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO2-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible materials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PFMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm-2; the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm-2 for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm-2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions

  12. Hamiltonian analysis of fast wave current drive in tokamak plasmas

    International Nuclear Information System (INIS)

    Becoulet, A.; Fraboulet, D.; Giruzzi, G.; Moreau, D.; Saoutic, B.

    1993-12-01

    The Hamiltonian formalism is used to analyze the direct resonant interaction between the fast magnetosonic wave and the electrons in a tokamak plasma. The intrinsic stochasticity of the electron phase space trajectories is derived, and together with extrinsic de-correlation processes, assesses the validity of the quasilinear approximation for the kinetic studies of fast wave current drive (FWCD). A full-wave resolution of the Maxwell-Vlasov set of equations provides the exact pattern of the wave fields in a complete tokamak geometry, for a realistic antenna spectrum. The local quasilinear diffusion tensor is derived from the wave fields, and is used for a computation of the driven current and deposited power profiles, the current drive efficiency, including possible non-linear effects in the kinetic equation. Several applications of FWCD on existing and future machines are given, as well as results concerning combination of FWCD with other non inductive current drive methods. An analytical expression for the current drive efficiency is given in the high single-pass absorption regimes. (authors). 20 figs., 1 tab., 26 refs

  13. Resistive fluid turbulence and tokamak edge plasma dynamics

    International Nuclear Information System (INIS)

    Thayer, D.R.; Diamond, P.H.; Ritz, C.P.

    1988-01-01

    Electrostatic and electromagnetic turbulence has been linked to particle and heat transport in tokamaks. Here we report on several related theoretical and experimental investigations of edge plasma dynamics. The theory of thermally-driven convective cell edge turbulence has been developed to treat the coupling of the radiative-condensation instability to the resistivity-gradient expansion free energy. This model of edge turbulence has led to theoretical understanding of several anomalies in electrostatic edge turbulence found from experiment: that fluctuation levels and transport coefficients are larger than naively expected, that potential fluctuations are significantly larger than the density. Impurity gas-puffing experiments on the TEXT tokamak have been performed to test this theory, and have indicated favorable results. Resistive fluid turbulence models have also been explored and applied in the hope of understanding the extensive edge magnetic fluctuation studies. We discuss models of electromagnetic microtearing turbulence, resistive-pressure-gradient-driven turbulence, and ion temperature gradient driven turbulence. In particular we study the role of resistive fluid turbulence with separatrix effects in the L /yield/ H mode transition. 36 refs., 2 figs

  14. Resistive Magnetohydrodynamics Simulation of Fusion Plasmas

    International Nuclear Information System (INIS)

    Tang, X.Z.; Fu, G.Y.; Jardin, S.C.; Lowe, L.L.; Park, W.; Strauss, H.R.

    2001-01-01

    Although high-temperature plasmas in laboratory magnetic fusion confinements are sufficiently collisionless that formal fluid closures are difficult to attain, the resistive MHD model has proven, by comparison with experimental data, to be useful for describing the large scale dynamics of magnetized plasmas. Resistive MHD model consists of Faraday's law for the evolution of the magnetic field and Navier-Stokes equation for the plasma flow. These equations are closed by the Ohm's law and an equation of state for the plasma

  15. Tokamak Fusion Core Experiment (TFCX) special-purpose remote maintenance systems

    International Nuclear Information System (INIS)

    Masson, L.S.; Welland, H.J.

    1985-01-01

    A key element in the preconceptual design of the Tokamak Fusion Core Experiment (TFCX) was the development of design concepts for special-purpose remote maintenance systems. Included were systems for shield sector replacement, vacuum vessel sector and toroidal field coil replacement, limiter blade replacement, protective tile replacement, and general-purpose maintenance. This paper addresses these systems as they apply to the copper toroidal field (TF) coil version of the TFCX

  16. 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)

  17. Safety analysis of superconducting toroidal field magnet for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-02-01

    Safety analysis of the superconducting toroidal field magnet for a Tokamak experimental fusion reactor has been carried out. Works were accident classification, FMEA and FTA analyses, coil stability and quench behavior calculations, failure detection and coil protection system designs, structure analysis, fracture and fatigue studies, and earthquake response analysis. Accident analysis of cryostat and refrigeration system was also performed. The objective of this work is to reveal technological problems of the toroidal field magnet by safety analysis. (author)

  18. Tokamak-7 operation in experiments with a plasma

    International Nuclear Information System (INIS)

    Buzanki, V.V.; Bychkov, A.V.; Denisov, V.F.

    1982-01-01

    The results of experiments with plasma at the Tokamak-7 (T-7) device are presented. The experiments have been carried out with a constant diaphragm of 31,5 cm radius and two movable graphite diaphragms at the 26-28 cm plasma filament radius and 1,6-1,9 T magnetic field. Two stable regimes with 150 and 200 kA and 250 ms discharge current length have been investigated. It is shown that the strongest poloidal filed perturhations have been observed at the beginning of the discharge. Electron plasma temperature Tsub(e) has been determined from the spectrum analysis of soft X radiation by the foil method. Stable plasma regimes with current up to 200 kA, bypass voltage being equal 1,58V electron density -0,5-5,0 x 10 13 cm -3 , Tsub(e)=1,1-1,3 keV ion temperature-490 eV. The range between discharge pulses has reached 3 min. at the discharge current-240 kA. No considerable effect of magnetic field variables on the superconducting magnetic system has been observed

  19. Development of the 'JFT-2' tokamak plasma position control system

    International Nuclear Information System (INIS)

    Fujisawa, Noboru; Matsuzaki, Yoshimi; Suzuki, Norio; Murai, Katsuji; Suzuki, Satoshi.

    1980-01-01

    Digital control technique was applied to control the plasma position in the JFT-2 tokamak experiment device. The detail of the JFT-2 is described elsewhere. The plasma position control system consists of a Hitachi control computer, HIDIC 80, and a Hitachi micro-computer, HIDIC 08E. The plasma position is detected by the position control computer, and compared with a preset value. Then, a reference signal is supplied to the micro-computer controlling power source, and the phase control of the thyristor controlling power source is performed. Since the behavior of plasma is very fast, the fast control is required. The control of the thyristor controlling power source is made by direct digital control (DDC). The main component of the hardware of the present system is the micro-computer HIDIC 08E. The software is the direct task system without the operating system (OS). The results of experiments showed that the feedback control of the system worked well. (Kato, T.)

  20. Modelisation of synchrotron radiation losses in realistic tokamak plasmas

    International Nuclear Information System (INIS)

    Albajar, F.; Johner, J.; Granata, G.

    2000-08-01

    Synchrotron radiation losses become significant in the power balance of high-temperature plasmas envisaged for next step tokamaks. Due to the complexity of the exact calculation, these losses are usually roughly estimated with expressions derived from a plasma description using simplifying assumptions on the geometry, radiation absorption, and density and temperature profiles. In the present article, the complete formulation of the transport of synchrotron radiation is performed for realistic conditions of toroidal plasma geometry with elongated cross-section, using an exact method for the calculation of the absorption coefficient, and for arbitrary shapes of density and temperature profiles. The effects of toroidicity and temperature profile on synchrotron radiation losses are analyzed in detail. In particular, when the electron temperature profile is almost flat in the plasma center, as for example in ITB confinement regimes, synchrotron losses are found to be much stronger than in the case where the profile is represented by its best generalized parabolic approximation, though both cases give approximately the same thermal energy contents. Such an effect is not included in present approximate expressions. Finally, we propose a seven-variable fit for the fast calculation of synchrotron radiation losses. This fit is derived from a large database, which has been generated using a code implementing the complete formulation and optimized for massively parallel computing. (author)

  1. Plasma features and alpha particle transport in low-aspect ratio tokamak reactor

    International Nuclear Information System (INIS)

    Xu Qiang; Wang Shaojie

    1997-06-01

    The results of the experiment and theory from low-aspect ratio tokamak devices have proved that the MHD stability will be improved. Based on present plasma physics and extrapolation to reduced aspect ratio, the feature of physics of low-aspect ratio tokamak reactor is discussed primarily. Alpha particle confinement and loss in the self-justified low-aspect ratio tokamak reactor parameters and the effect of alpha particle confinement and loss for different aspect ratio are calculated. The results provide a reference for the feasible research of compact tokamak reactor. (9 refs., 2 figs., 3 tabs.)

  2. 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

  3. 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

  4. Simulations of Neon Pellets for Plasma Disruption Mitigation in Tokamaks

    Science.gov (United States)

    Bosviel, Nicolas; Samulyak, Roman; Parks, Paul

    2017-10-01

    Numerical studies of the ablation of neon pellets in tokamaks in the plasma disruption mitigation parameter space have been performed using a time-dependent pellet ablation model based on the front tracking code FronTier-MHD. The main features of the model include the explicit tracking of the solid pellet/ablated gas interface, a self-consistent evolving potential distribution in the ablation cloud, JxB forces, atomic processes, and an improved electrical conductivity model. The equation of state model accounts for atomic processes in the ablation cloud as well as deviations from the ideal gas law in the dense, cold layers of neon gas near the pellet surface. Simulations predict processes in the ablation cloud and pellet ablation rates and address the sensitivity of pellet ablation processes to details of physics models, in particular the equation of state.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  6. EURATOM-CEA Association Contributions to the 16. European Conference on Controlled Fusion and Plasma Physics

    International Nuclear Information System (INIS)

    1989-01-01

    The contributions to the 16th European Conference on controlled fusion and Plasma Physics are presented. The following subjects, concerning Tore Supra, are discussed: runaway electrons dynamics and confinement; spectroscopic studies of plasma surface interactions; ergodic divertor experiments; magnetic field structure and transport induced by the ergodic divertor; fast ions losses during neutral beam injection; current profile control by electron-cyclotron and lower-hybrid waves; and electromagnetic analysis of the lower hybrid system. The report also includes studies on: a possible explanation for the runaway energy limit (resonant interaction with the ripple field); thermal equilibrium of the edge plasma with an ergodic divertor; neutral confinement in pump limiter with a throat; microtearing turbulence and heat transport; toroidal coupling and frequency spectrum of tearing modes; collisionless fast ion dynamics in tokamaks; variational description of lower hybrid wave propagation and absorption in tokamaks; magnetodrift turbulence and disruptions; specific turbulence associated with sawtooth relaxations in TFR plasmas; detailed structure of the q profile around q = 1 in JET; turbulence propagation during pellet injection; tokamak reactor concept with 100% bootstrap current; optimization of a steady state tokamak driven by lower hybrid waves; and thermodesorption of graphite exposed to a deuterium plasma

  7. Application of fusion plasma technology. Final report

    International Nuclear Information System (INIS)

    Sabri, Z.A.

    1976-06-01

    This report presents principal findings of studies conducted at Iowa State on Applications of Fusion Plasma Technology. Two tasks were considered. The first was to identify and investigate plasma processes for near term industrial applications of already developed plasma technology. The second was to explore the potential of reprocessing the fuel for fusion test facilities in a closed-cycle system. For the first task, two applications were considered. One was alumina reduction in magnetically confined plasmas, and the other was uranium enrichment using plasma centrifuges. For the second task, in-core and ex-core plasma purification were considered. Separation techniques that are compatible with the plasma state were identified and preliminary analysis of their effectiveness were carried out. The effects of differential ionization of impurities on the separation effectiveness are considered. Possible technical difficulties in both tasks are identified and recommendations for future work are given

  8. 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

  9. On the origin, properties, and implications of asymmetries in the tungsten impurity density in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Odstrcil, Tomas

    2017-07-03

    In this thesis, the transport of tungsten ions is studied in the plasma of ASDEX Upgrade tokamak. The plasma facing components of the fusion reactors are expected to be built from high-Z materials such as W, Mo or Fe. These materials provide advantages like a high melting point, small erosion rates, and low tritium retention. However, due to the interaction of the plasma with the wall, ions of this material will be inevitably present also in the main plasma. These ions are not entirely stripped even at fusion plasma temperatures, and therefore emit strong line radiation, which can significantly degrade the performance of the fusion plasma. Thus the understanding and control of impurity transport are of critical importance to the success of fusion. The high mass and charge of the heavy impurities make them susceptible to some of the forces acting upon the plasma, resulting in a poloidal variation of their density. The most prominent are the centrifugal force arising from the plasma rotation and the electric force caused by magnetically trapped non-thermal ions. Furthermore, the poloidal asymmetries should have a significant impact on the radial transport of heavy ions, which was widely ignored up to date. In the present work, the poloidal asymmetries in the heavy impurity density were inferred from the soft X-ray radiation using a newly developed tomographic method. The high accuracy of the tomography and of the model for the centrifugal force allowed to identify for the first time in an experiment the effect of the fast ion distribution produced by neutral beam injection on the poloidal asymmetry of the tungsten density. The measured asymmetry was compared to several fast ion models, and the best match was found with the Monte Carlo code in the TRANSP code suite that includes finite orbits effects of the fast ions. Similarly, fast ions accelerated by ion cyclotron heating and localized mainly in the outboard side of the plasma due to a magnetic trapping and produce

  10. Determination of the plasma position for its real-time control in the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Havlicek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, CZ-18000 Prague (Czech Republic); Valcarcel, D. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal); Hron, M.; Horacek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Kudlacek, O. [Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering, Technicka 2, 166 27 Prague (Czech Republic); Panek, R. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Prague (Czech Republic); Carvalho, B.B. [Associacao EURATOM/IST, Instituto de Plasmas e Fusao Nuclear - Laboratorio Associado, Instituto Superior Tecnico, P1049-001 Lisboa (Portugal)

    2011-10-15

    An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

  11. Modeling of noble gas injection into tokamak plasmas

    International Nuclear Information System (INIS)

    Morozov, D.Kh.; Yurchenko, E.I.; Lukash, V.E.; Baronova, E.O.; Rozhansky, V.A.; Senichenkov, I.Yu.; Veselova, I.Yu.; Schneider, R.

    2005-01-01

    Noble gas injection for mitigation of the disruption in DIII-D is simulated. The simulation of the first two stages is performed: of the neutral gas jet penetration through the background plasmas, and of the thermal quench. In order to simulate the first stage the 1.5-dimensional numerical code LLP with improved radiation model for noble gas is used. It is demonstrated that the jet remains mainly neutral and thus is able to penetrate to the central region of the tokamak in accordance with experimental observations. Plasma cooling at this stage is provided by the energy exchange with the jet. The radiation is relatively small, and the plasma thermal energy is spent mainly on the jet expansion. The magnetic surfaces in contact with the jet are cooled significantly. The cooling front propagates towards the plasma center. The simulations of the plasma column dynamics in the presence of moving jet is performed by means of the free boundary transport modeling DINA code. It has been shown that the cooling front is accompanied by strongly localized 'shark fin-like' perturbation in toroidal current density profile. After few milliseconds the jet (together with the current perturbation) achieves the region where safety factor is slightly higher than unity and a new type of the non-local kink mode develops. The unstable kink perturbation is non-resonant for any magnetic surface, both inside the plasma column, and in the vacuum space. The mode disturbs mainly the core region. The growth time of the 'shark fin-like' mode is higher than the Alfven time by a factor of 100 for DIII-D parameters. Hence, the simulation describes the DIII-D experimental results, at least, qualitatively. (author)

  12. Toothbrush probe for instantaneous measurement of radial profile in tokamak boundary plasma

    Energy Technology Data Exchange (ETDEWEB)

    Uehara, Kazuya; Sengoku, Seio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Amemiya, Hiroshi

    1997-04-01

    A new probe for the instantaneous measurement of radial profiles of the boundary scrape-off layer (SOL) plasma has been developed in a tokamak. Five asymmetric double-probe chips are aligned in parallel to a strong magnetic field in the boundary plasma in a tokamak. This probe is named the `toothbrush probe` and can measure the ion temperature as well as the electron temperature and the plasma density in the SOL plasma within only one tokamak plasma shot. First, only one asymmetric probe is mounted on the divertor plate and it is tried to determine the ion temperature. Then, a manufactured toothbrush probe is mounted in the SOL plasma and the radial plasma profiles are simultaneously obtained. Data on the e-folding length of the plasma profile obtained by the toothbrush probe can determine the information on the transport properties such as the diffusion coefficient and the thermal conductivity of electrons and ions. (author)

  13. Analytical modelling of resistive wall mode stabilization by rotation in toroidal tokamak plasmas

    International Nuclear Information System (INIS)

    Ham, C J; Gimblett, C G; Hastie, R J

    2011-01-01

    Stabilization of the resitive wall mode (RWM) may allow fusion power to be doubled for a given magnetic field in advanced tokamak operation. Experimental evidence from DIII-D and other machines suggests that plasma rotation can stabilize the RWM. Several authors (Finn 1995 Phys. Plasmas 2 3782, Bondeson and Xie 1997 Phys. Plasmas 4 2081) have constructed analytical cylindrical models for the RWM, but these do not deal with toroidal effects. The framework of Connor et al (1988 Phys. Fluids 31 577) is used to develop ideal plasma analytic models with toroidicity included. Stepped pressure profiles and careful ordering of terms are used to simplify the analysis. First, a current driven kink mode model is developed and a dispersion relation for arbitrary current profile is calculated. Second, the external pressure driven kink mode is similarly investigated as the most important RWM arises from this mode. Using this latter model it is found that the RWM is stabilized by Alfven continuum damping with rotation levels similar to those seen in experiments. An expression for the stability of the external kink mode for more general current profiles and a resistive wall is derived in the appendix.

  14. 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)

  15. Langmuir Probe Evaluation of the Plasma Potential in Tokamak Edge Plasma for Non-Maxwellian EEDF

    Czech Academy of Sciences Publication Activity Database

    Popov, Tsv.K.; Dimitrova, Miglena; Ivanova, P.; Hasan, E.; Horáček, Jan; Dejarnac, Renaud; Stöckel, Jan; Weinzettl, Vladimír; Kovačič, J.

    2014-01-01

    Roč. 54, č. 3 (2014), s. 267-272 ISSN 0863-1042. [International Workshop on Electric Probes in Magnetized Plasmas/10./. Madrid, 09.07.2013-12.07.2013] R&D Projects: GA MŠk(CZ) LM2011021; GA ČR(CZ) GAP205/12/2327 Institutional support: RVO:61389021 Keywords : Langmuir probe * Plasma potential * First derivative probe technique * Scrape-off-layer * COMPASS tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.838, year: 2014 http://onlinelibrary.wiley.com/doi/10.1002/ctpp.201410076/abstract

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

    International Nuclear Information System (INIS)

    Minato, Akio; Tone, Tatsuzo

    1985-12-01

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

  17. Plasma pressure in the discharge column of the Novillo Tokamak

    International Nuclear Information System (INIS)

    Gaytan G, E.

    1995-01-01

    The design and construction of an acquisition system for the measurement of the plasma pressure in the Novillo Tokamak is described in detail. The system includes a high voltage ramp generator, a hardware and a software interface with a personal computer. It is used to determine experimentally the variations of the pressure in the plasma column in the cleaning and main discharges. The measurement of the pressure is made with a Pirani sensor adapted to the acquisition hardware and synchronized with the discharge in the plasma. The software is made in object oriented programming as a graphic interface designed to be used easily. It controls the acquisition, records the data, displays in graphic form the results and save the measurements. The graphic interface is a building block that can be used in different acquisition tasks. The ramp generator can deliver a signal of 200 V peak to peak with a current of 200 m A and offset control. The acquisition time is 2.5 μ s for every measurement, 8192 measurements can be stored in the acquisition board for every discharge. (Author)

  18. Plasma assessments for the fusion engineering device (FED)

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Rutherford, P.H.; Lyon, J.F.

    1981-01-01

    An initial range of plasma assumptions and scenarios has been examined for the US tokamak FED concept. The results suggest that the current FED baseline parameters of R = 4.8 m, B/sub t/ = 3.6 T, a = 1.3 m, b = 2.1 m (D-shape), and I/sub p/ = 4.8 to 5.4 MA are appropriate for achieving its nominal goals of P(fusion) approx. = 180 MW and a plasma Q greater than or equal to to 5 for a pulse length greater than 100 s. However, large uncertainty still exists in the areas of current startup, ion-cyclotron wave launching, influence of plasma shape on achievable beta, impurity control, plasma edge transport, and plasma disruption. Various options and remedies have been suggested to alleviate the impact of the uncertainty on the FED design concept. They appear promising because they can be studied experimentally and are not expected to lead to fundamental design modifications of FED

  19. Radiation control in fusion plasmas by magnetic confinement

    International Nuclear Information System (INIS)

    Dachicourt, R.

    2012-10-01

    The present work addresses two important issues for the industrial use of fusion: plasma radiation control, as a part of the more general power handling issue, and high density tokamak operation. These two issues will be most critical in the demonstration reactor, called DEMO, intermediate step between ITER and a future commercial reactor. For DEMO, the need to radiate a large fraction of the power so as to limit the peak power load on the divertor will be a key constraint. High confinement will have to be combined with high radiated power fraction, and the required level of plasma purity. The main achievement of this thesis is to have shown experimental evidence of the existence of a stable plasma regime meeting the most critical requirements of a DEMO scenario: an electron density up to 40% above the Greenwald value, together with a fraction of radiated power close to 80%, with a good energy confinement and limited dilution. The plasma is additionally heated with ion cyclotron waves in a central electron heating scenario, featuring alpha particle heating. The original observations reported in this work bring highly valuable new pieces of information both to the physics of the tokamak edge layer and to the construction of an 'integrated operational scenario' required to successfully operate fusion devices. In the way for getting high density plasmas, the new observations involve the following topics. First, the formation of a poloidal asymmetry in the edge electron density profile, with a maximum density located close to toroidal pumped limiter. This asymmetry occurs inside the separatrix, with a constant plasma pressure on magnetic surfaces. Secondly, a correlative decrease of the electron temperature in the same edge region. Thirdly, the excellent coupling capabilities of the ICRH waves, up to a central line averaged electron density of 1.4 times the Greenwald density. Fourthly, a poloidally asymmetric edge radiation region, providing the dissipation of 80% of

  20. Monitoring and Hardware Management for Critical Fusion Plasma Instrumentation

    Directory of Open Access Journals (Sweden)

    Carvalho Paulo F.

    2018-01-01

    Full Text Available Controlled nuclear fusion aims to obtain energy by particles collision confined inside a nuclear reactor (Tokamak. These ionized particles, heavier isotopes of hydrogen, are the main elements inside of plasma that is kept at high temperatures (millions of Celsius degrees. Due to high temperatures and magnetic confinement, plasma is exposed to several sources of instabilities which require a set of procedures by the control and data acquisition systems throughout fusion experiments processes. Control and data acquisition systems often used in nuclear fusion experiments are based on the Advanced Telecommunication Computer Architecture (AdvancedTCA® standard introduced by the Peripheral Component Interconnect Industrial Manufacturers Group (PICMG®, to meet the demands of telecommunications that require large amount of data (TB transportation at high transfer rates (Gb/s, to ensure high availability including features such as reliability, serviceability and redundancy. For efficient plasma control, systems are required to collect large amounts of data, process it, store for later analysis, make critical decisions in real time and provide status reports either from the experience itself or the electronic instrumentation involved. Moreover, systems should also ensure the correct handling of detected anomalies and identified faults, notify the system operator of occurred events, decisions taken to acknowledge and implemented changes. Therefore, for everything to work in compliance with specifications it is required that the instrumentation includes hardware management and monitoring mechanisms for both hardware and software. These mechanisms should check the system status by reading sensors, manage events, update inventory databases with hardware system components in use and maintenance, store collected information, update firmware and installed software modules, configure and handle alarms to detect possible system failures and prevent emergency

  1. Monitoring and Hardware Management for Critical Fusion Plasma Instrumentation

    Science.gov (United States)

    Carvalho, Paulo F.; Santos, Bruno; Correia, Miguel; Combo, Álvaro M.; Rodrigues, AntÓnio P.; Pereira, Rita C.; Fernandes, Ana; Cruz, Nuno; Sousa, Jorge; Carvalho, Bernardo B.; Batista, AntÓnio J. N.; Correia, Carlos M. B. A.; Gonçalves, Bruno

    2018-01-01

    Controlled nuclear fusion aims to obtain energy by particles collision confined inside a nuclear reactor (Tokamak). These ionized particles, heavier isotopes of hydrogen, are the main elements inside of plasma that is kept at high temperatures (millions of Celsius degrees). Due to high temperatures and magnetic confinement, plasma is exposed to several sources of instabilities which require a set of procedures by the control and data acquisition systems throughout fusion experiments processes. Control and data acquisition systems often used in nuclear fusion experiments are based on the Advanced Telecommunication Computer Architecture (AdvancedTCA®) standard introduced by the Peripheral Component Interconnect Industrial Manufacturers Group (PICMG®), to meet the demands of telecommunications that require large amount of data (TB) transportation at high transfer rates (Gb/s), to ensure high availability including features such as reliability, serviceability and redundancy. For efficient plasma control, systems are required to collect large amounts of data, process it, store for later analysis, make critical decisions in real time and provide status reports either from the experience itself or the electronic instrumentation involved. Moreover, systems should also ensure the correct handling of detected anomalies and identified faults, notify the system operator of occurred events, decisions taken to acknowledge and implemented changes. Therefore, for everything to work in compliance with specifications it is required that the instrumentation includes hardware management and monitoring mechanisms for both hardware and software. These mechanisms should check the system status by reading sensors, manage events, update inventory databases with hardware system components in use and maintenance, store collected information, update firmware and installed software modules, configure and handle alarms to detect possible system failures and prevent emergency scenarios

  2. Internal transport barrier in tokamak and helical plasmas

    Science.gov (United States)

    Ida, K.; Fujita, T.

    2018-03-01

    The differences and similarities between the internal transport barriers (ITBs) of tokamak and helical plasmas are reviewed. By comparing the characteristics of the ITBs in tokamak and helical plasmas, the mechanisms of the physics for the formation and dynamics of the ITB are clarified. The ITB is defined as the appearance of discontinuity of temperature, flow velocity, or density gradient in the radius. From the radial profiles of temperature, flow velocity, and density the ITB is characterized by the three parameters of normalized temperature gradient, R/{L}T, the location, {ρ }{ITB}, and the width, W/a, and can be expressed by ‘weak’ ITB (small R/{L}T) or ‘strong’ (large R/{L}T), ‘small’ ITB (small {ρ }{ITB}) or ‘large’ ITB (large {ρ }{ITB}), and ‘narrow’ (small W/a) or ‘wide’ (large W/a). Three key physics elements for the ITB formation, radial electric field shear, magnetic shear, and rational surface (and/or magnetic island) are described. The characteristics of electron and ion heat transport and electron and impurity transport are reviewed. There are significant differences in ion heat transport and electron heat transport. The dynamics of ITB formation and termination is also discussed. The emergence of the location of the ITB is sometimes far inside the ITB foot in the steady-state phase and the ITB region shows radial propagation during the formation of the ITB. The non-diffusive terms in momentum transport and impurity transport become more dominant in the plasma with the ITB. The reversal of the sign of non-diffusive terms in momentum transport and impurity transport associated with the formation of the ITB reported in helical plasma is described. Non-local transport plays an important role in determining the radial profile of temperature and density. The spontaneous change in temperature curvature (second radial derivative of temperature) in the ITB region is described. In addition, the key parameters of the control of the

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

    International Nuclear Information System (INIS)

    Perez-Navarro, A.

    1977-01-01

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

  4. (Fusion energy research)

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, C.A. (ed.)

    1988-01-01

    This report discusses the following topics: principal parameters achieved in experimental devices (FY88); tokamak fusion test reactor; Princeton beta Experiment-Modification; S-1 Spheromak; current drive experiment; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical plasma; tokamak modeling; compact ignition tokamak; international thermonuclear experimental reactor; Engineering Department; Project Planning and Safety Office; quality assurance and reliability; and technology transfer.

  5. [Fusion energy research

    International Nuclear Information System (INIS)

    Phillips, C.A.

    1988-01-01

    This report discusses the following topics: principal parameters achieved in experimental devices (FY88); tokamak fusion test reactor; Princeton beta Experiment-Modification; S-1 Spheromak; current drive experiment; x-ray laser studies; spacecraft glow experiment; plasma deposition and etching of thin films; theoretical plasma; tokamak modeling; compact ignition tokamak; international thermonuclear experimental reactor; Engineering Department; Project Planning and Safety Office; quality assurance and reliability; and technology transfer

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

    International Nuclear Information System (INIS)

    Pachtman, A.

    1986-09-01

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

  7. Engineering analysis of new Brazilian Tokamak

    International Nuclear Information System (INIS)

    Tuszel, A.G.

    1990-01-01

    The engineering basic headlines are described. A project for the construction of a new tokamak is being developed at the Institute of Physics, University of Sao Paulo. The tokamak named TBR-II will be a medium size tokamak using two high power generators of 15 MW each and concepted as a versatile device for plasma physics research of interest for thermonuclear fusion studies. (Author)

  8. Anomalous electron streaming due to electrostatic modes in tokamak plasmas

    International Nuclear Information System (INIS)

    Schultz, S.D.; Bers, A.; Ram, A.K.

    1993-01-01

    The motion of circulating electrons in a tokamak interacting with electrostatic waves (such as lower-hybrid waves) is given by a guiding center Hamiltonian and studied by numerical integration. The unperturbed motion of electron guiding centers is first shown to be integrable, and, in a manner similar to that used in previous works, a set of action-angle coordinates for the orbits are derived which take into account finite aspect ratio and noncircular plasma cross section. Electrostatic modes in the low-frequency, long-wavelength limit are treated as a perturbation to the guiding center Hamiltonian. The waves are generated with low integral values of the toroidal and poloidal mode numbers n and m and satisfy the approximate lower-hybrid dispersion relation k perpendicular /k parallel ∼ ω pe /ω ∼ 10 1.5 . If the number of modes is greater than three, the electron motion parallel to the magnetic field is observed to be stochastic in the phase-space region where v parallel is near the wave parallel phase velocity. On surfaces with rational values of the safety factor q, superposition of modes with degenerate values of the parallel mode number n + (m/q) is shown to result in electron streaming perpendicular to the magnetic field. The speed and direction of this radial motion are observed to have sinusoidal dependence on the poloidal angle. For models including finite magnetic-field shear, the authors find a limit to the extent of the radial streaming of the electrons. Results for the speed of the electron radial motion for typical tokamak parameters are presented

  9. Effects of orbit squeezing on neoclassical toroidal plasma viscosity in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Shaing, K.C.; Sabbagh, S.A.; Chu, M.S.; Bécoulet, M.; Cahyna, Pavel

    2008-01-01

    Roč. 15, č. 8 (2008), 082505-1-082505-8 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma instability * plasma magnetohydrodynamics * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2965146

  10. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.

    2001-01-01

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

  11. Study of plasma discharge evolution and edge turbulence with fast visible imaging in the Aditya tokamak

    International Nuclear Information System (INIS)

    Banerjee, Santanu; Manchanda, R.; Chowdhuri, M.B.

    2015-01-01

    Study of discharge evolution through the different phases of a tokamak plasma shot viz., the discharge initiation, current ramp-up, current flat-top and discharge termination, is essential to address many inherent issues of the operation of a Tokamak. Fast visible imaging of the tokamak plasma can provide valuable insight in this regard. Further, edge turbulence is considered to be one of the quintessential areas of tokamak research as the edge plasma is at the immediate vicinity of the plasma core and plays vital role in the core plasma confinement. The edge plasma also bridges the core and the scrape off layer (SOL) of the tokamak and hence has a bearing on the particle and heat flux escaping the plasma column. Two fast visible imaging systems are installed on the Aditya tokamak. One of the system is for imaging the plasma evolution with a wide angle lens covering a major portion of the vacuum vessel. The imaging fiber bundle along with the objective lens is installed inside a radial re-entrant viewport, specially designed for the purpose. Another system is intended for tangential imaging of the plasma column. Formation of the plasma column and its evolution are studied with the fast visible imaging in Aditya. Features of the ECRH and LHCD operations on Aditya will be discussed. 3D filaments can, be seen at the plasma edge all along the discharge and they get amplified in intensity at the plasma termination phase. Statistical analysis of these filaments, which are essentially plasma blobs will be presented. (author)

  12. Fast ions and momentum transport in JET tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Salmi, A.

    2012-07-01

    Fast ions are an inseparable part of fusion plasmas. They can be generated using electromagnetic waves or injected into plasmas as neutrals to heat the bulk plasma and to drive toroidal rotation and current. In future power plants fusion born fast ions deliver the main heating into the plasma. Understanding and controlling the fast ions is of crucial importance for the operation of a power plant. Furthermore, fast ions provide ways to probe the properties of the thermal plasma and get insight of its confinement properties. In this thesis, numerical code packages are used and developed to simulate JET experiments for a range of physics issues related to fast ions. Namely, the clamping fast ion distribution at high energies with RF heating, fast ion ripple torque generation and the toroidal momentum transport properties using NBI modulation technique are investigated. Through a comparison of numerical simulations and the JET experimental data it is shown that the finite Larmor radius effects in ion cyclotron resonance heating are important and that they can prevent fast ion tail formation beyond certain energy. The identified mechanism could be used for tailoring the fast ion distribution in future experiments. Secondly, ASCOT simulations of NBI ions in a ripple field showed that most of the reduction of the toroidal rotation that has been observed in the JET enhanced ripple experiments could be attributed to fast ion ripple torque. Finally, fast ion torque calculations together with momentum transport analysis have led to the conclusion that momentum transport in not purely diffusive but that a convective component, which increases monotonically in radius, exists in a wide range of JET plasmas. Using parameter scans, the convective transport has been shown to be insensitive to collisionality and q-profile but to increase strongly against density gradient. (orig.)

  13. Interaction between fast ions and ion cyclotron heating in a tokamak plasma; Interaction des ions rapides avec les ondes a la frequence cyclotronique ionique dans un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bergeaud, V

    2001-11-01

    In an ignited fusion reactor, the plasma temperature is sustained by the fusion reactions. However, before this regime is reached, it is necessary to bring an additional power to the plasma. One of the methods that enables the coupling of power is the use of an electromagnetic wave in the ion cyclotron range of frequencies (ICRF). This thesis deals with the interaction between ICRF heating and the fast ions. The thesis contains a theoretical study of the influence of ICRF heating on the ion distribution function. A particular emphasis is put on the importance of the toroidal spectrum of the modes of propagation of the wave in the tokamak. It is necessary to take into account all these modes in order to correctly assess the strength of the wave particle interaction, especially for high energy particles (of the order of hundreds of keV). The classical treatment of the wave particle interaction is based on the hypothesis that the cyclotron phase of the particle and the wave phase are de-correlated between successive resonant interactions. One is therefore led to consider ICRF heating as a diffusive process. This hypothesis is reconsidered in this thesis and it is shown that strong correlations exist in a large part of the velocity space. For this study, a numerical code that computes the full trajectory of particles interacting with a complete electromagnetic field has been developed. The thesis also deals with the problem of fast ion losses due to the breaking of the toroidal symmetry of the confinement magnetic field (called the ripple modulation). Between two toroidal coils, local magnetic wells exist, and particles can be trapped there. When trapped they undergo a vertical drift that makes them quit the plasma rapidly. The ripple modulation also causes an enhancement of the radial diffusion, thereby increasing the losses. A Monte Carlo model describing these mechanisms is presented. This model is validated thanks to a comparison with an experimental database from

  14. Interaction of fast ions with ion cyclotron electromagnetic waves in tokamak plasma; Interaction des ions rapides avec les ondes a la frequence cyclotronique ionique dans un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bergeaud, V

    2000-12-01

    In an ignited fusion reactor, the plasma temperature is sustained by the fusion reactions. However, before this regime is reached, it is necessary to bring an additional power to the plasma. One of the methods that enables the coupling of power is the use of an electromagnetic wave in the ion cyclotron range of frequencies (ICRF). This thesis deals with the interaction between ICRF heating and the fast ions. The thesis contains a theoretical study of the influence of ICRF heating on the ion distribution function. A particular emphasis is put on the importance of the toroidal spectrum of the modes of propagation of the wave in the tokamak. It is necessary to take into account all these modes in order to correctly assess the strength of the wave particle interaction, especially for high energy particles (of the order of hundreds of keV). The classical treatment of the wave particle interaction is based on the hypothesis that the cyclotron phase of the particle and the wave phase are de-correlated between successive resonant interactions. One is therefore led to consider ICRF heating as a diffusive process. This hypothesis is reconsidered in this thesis and it is shown that strong correlations exist in a large part of the velocity space. For this study, a numerical code that computes the full trajectory of particles interacting with a complete electromagnetic field has been developed. The thesis also deals with the problem of fast ion losses due to the breaking of the toroidal symmetry of the confinement magnetic field (called the ripple modulation). Between two toroidal coils, local magnetic wells exist, and particles can be trapped there. When trapped they undergo a vertical drift that makes them quit the plasma rapidly. The ripple modulation also causes an enhancement of the radial diffusion, thereby increasing the losses. A Monte Carlo model describing these mechanisms is presented. This model is validated thanks to a comparison with an experimental database from

  15. TMX: a new fusion plasma experiment

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The primary goal of the magnetic fusion energy program at LLL is the development of a technically and economically feasible approach to the generation of fusion energy. Results from our earlier 2XIIB experiment lead us to believe that a fusion power plant based on a mirror system is technically feasible, assuming a favorable extrapolation to plasmas of reactor size. Achieving economic feasibility is more difficult. For power-producing applications, a reactor needs a large Q, the ratio of fusion power output to the power injected to sustain the system. In a conventional mirror reactor, the fusion power is only about equal to the power injected by the neutral beams--that is, Q is only about unity. A new idea, the tandem mirror concept described in this article, promises to increase this gain, enhancing Q by at least a factor of 5

  16. Some stress-related issues in tokamak fusion reactor first walls

    International Nuclear Information System (INIS)

    Majumdar, S.; Pai, B.; Ryder, R.H.

    1987-01-01

    Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m 2 , and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems

  17. Collisional boundary layer analysis for neoclassical toroidal plasma viscosity in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Shaing, K.C.; Cahyna, Pavel; Bécoulet, M.; Park, J.-K.; Sabbagh, S.A.; Chu, M.S.

    2008-01-01

    Roč. 15, č. 8 (2008), 082506-1-7 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2969434

  18. A DC probe diagnostics for fast electron temperature measurements in tokamak edge plasmas

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Devynck, P.; Pascal, J. Y.; Adámek, Jiří; Ďuran, Ivan; Hron, Martin; Stöckel, Jan; Žáček, František; Bařina, O.; Hrach, R.; Vicher, M.

    2002-01-01

    Roč. 52, č. 10 (2002), s. 1107-1114 ISSN 0011-4626. [Workshop"Role of Electric Fields in Plasma Confinement and Exhaust"/5th./. Montreux, 23.06.2002-24.06.2002] Institutional research plan: CEZ:AV0Z2043910 Keywords : plasma, tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.311, year: 2002

  19. Advanced Probe Measurements of Electron Energy Distribution Functions in CASTOR Tokamak Plasma

    Czech Academy of Sciences Publication Activity Database

    Popov, T.; Stöckel, Jan; Dejarnac, Renaud; Dimitrova, M.; Ivanova, P.; Naydenova, Tsv.

    2006-01-01

    Roč. 63, bez (2006), 012002-012003 E-ISSN 1742-6596. [SECOND INTERNATIONAL WORKSHOP AND SUMMER SCHOOL ON PLASMA PHYSICS. Kiten, 03.07.2006-09.07.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics

  20. Role of turbulence and electric fields in the establishment of improved confinement in tokamak plasmas

    Czech Academy of Sciences Publication Activity Database

    Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Krämer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Peleman, P.; Razumova, K.; Stöckel, Jan; Vershkov, V.; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Dnestrovskij, A.Yu.; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Jachmin, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Soldatov, S.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Zimmermann, O.

    2006-01-01

    Roč. 12, č. 6 (2006), s. 14-19 ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * improved confinement * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics http:// vant .kipt.kharkov.ua/TABFRAME.html

  1. Dynamics of the edge transport barrier at plasma biasing on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Spolaore, M.; Peleman, P.; Brotánková, Jana; Horáček, Jan; Dejarnac, Renaud; Devynck, P.; Ďuran, Ivan; Gunn, J. P.; Hron, Martin; Kocan, M.; Martines, E.; Pánek, Radomír; Sharma, A.; Van Oost, G.

    2006-01-01

    Roč. 12, č. 6 (2006), s. 19-23 ISSN 1562-6016. [International Conference on Plasma Physics and Technology/11th./. Alushta, 11.9.2006-16.9.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * transport barrier * relaxations Subject RIV: BL - Plasma and Gas Discharge Physics http:// vant .kipt.kharkov.ua/TABFRAME.html

  2. 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

  3. 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

  4. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. DREAM (DRastically EAsy Maintenance) tokamak

    International Nuclear Information System (INIS)

    Nishio, Satoshi

    1998-01-01

    If the major part of the electric power demand will be supplied by tokamak fusion power plants, a suitable tokamak reactor must be an ultimate goal, i.e., the reactor must be excellent both in terms of construction cost and safety aspects including operation availability (maintainability and reliability). In attaining this goal, an approach focusing on both safety and availability (including reliability and maintainability) issues is the most promising strategy. The tokamak reactor concept with a very high aspect ratio configuration and SiC/SiC composite structural materials is compatible with this approach, which is called the DREAM (DRastically EAsy Maintenance) approach. The SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to good accessibility for the maintenance of machines. As an intermediate steps between an experimental reactor such as ITER and the ultimate goal, the development of prototype reactor which demonstrates electric power generation and an initial-phase commercial reactor which demonstrates for COE (cost of electricity) competitiveness has been investigated. Especially for the prototype reactor, material and technological immaturity must be considered. (J.P.N.)

  5. Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y.

    1996-12-01

    This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage

  6. Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A.; Stevens, P.L.; Hino, T.; Hirohata, Y. [eds.

    1996-12-01

    This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage.

  7. Tokamak

    International Nuclear Information System (INIS)

    Meglicki, Z.

    1995-01-01

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs

  8. Fluctuations and transport in fusion plasmas. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Gould, R.W.; Liewer, P.C.

    1995-02-01

    The energy confinement in tokamaks in thought to be limited by transport caused by plasma turbulence. Three dimensional plasma particle-in-cell (PIC) codes are used to model the turbulent transport in tokamaks to attempt to understand this phenomena so that tokamaks can be made more efficient. Presently, hundreds of hours of Cray time are used to model these experiments and much bigger and longer runs are desired, to model a large tokamak with realistic parameters is beyond the capability of existing sequential supercomputers. Parallel supercomputers might be a cost effect tool for performing such large scale 3D tokamak simulations. The goal of the work was to develop algorithms for performing PIC codes on coarse-grained message passing parallel computers and to evaluate the performance of such parallel computers on PIC codes. This algorithm would be used in a large scale PIC production code such as the UCLA 3D gyrokinetic code.

  9. Fluctuations and transport in fusion plasmas. Final report

    International Nuclear Information System (INIS)

    Gould, R.W.; Liewer, P.C.

    1995-01-01

    The energy confinement in tokamaks in thought to be limited by transport caused by plasma turbulence. Three dimensional plasma particle-in-cell (PIC) codes are used to model the turbulent transport in tokamaks to attempt to understand this phenomena so that tokamaks can be made more efficient. Presently, hundreds of hours of Cray time are used to model these experiments and much bigger and longer runs are desired, to model a large tokamak with realistic parameters is beyond the capability of existing sequential supercomputers. Parallel supercomputers might be a cost effect tool for performing such large scale 3D tokamak simulations. The goal of the work was to develop algorithms for performing PIC codes on coarse-grained message passing parallel computers and to evaluate the performance of such parallel computers on PIC codes. This algorithm would be used in a large scale PIC production code such as the UCLA 3D gyrokinetic code

  10. BOOK REVIEW: Fundamentals of Plasma Physics and Controlled Fusion

    Science.gov (United States)

    Brambilla, Marco

    1998-04-01

    Professor Kenro Miyamoto, already well known for his textbook Plasma Physics for Nuclear Fusion (MIT Press, Cambridge, MA, 1976; revised edition 1989), has now published a new book entitled Fundamentals of Plasma Physics and Controlled Fusion (Iwanami Book Service Center, Tokyo, 1997). To a large extent, the new book is a somewhat shortened and well reorganized version of its predecessor. The style, concise and matter of fact, clearly shows the origin of the text in lectures given by the author to graduate students. As announced by the title, the book is divided into two parts: the first part (about 250 pages) is a general introduction to the physics of plasmas, while the second, somewhat shorter, part (about 150 pages), is devoted to a description of the most important experimental approaches to achieving controlled thermonuclear fusion. Even in the first part, moreover, the choice of subjects is consistently oriented towards the needs of fusion research. Thus, the introduction to the behaviour of charged particles (particle motion, collisions, etc.) and to the collective description of plasmas is quite short, although the reader will get a flavour of all the most important topics and will find a number of examples chosen for their relevance to fusion applications (only the presentation of the Vlasov equation, in the second section of Chapter 4, might be criticized as so concise as to be almost misleading, since the difference between microscopic and macroscopic fields is not even mentioned). Considerably more space is devoted to the magnetohydrodynamic (MHD) description of equilibrium and stability. This part includes the solution of the Grad-Shafranov equation for circular tokamaks, a brief discussion of Pfirsch-Schlüter, neoclassical and anomalous diffusion, and two relatively long chapters on the most important ideal and resistive MHD instabilities of toroidal plasmas; drift and ion temperature gradient driven instabilities are also briefly presented. The

  11. Science Court on ICRH [ion cyclotron resonance heating] modeling of tokamak plasmas

    International Nuclear Information System (INIS)

    Hively, L.M.; Sadowski, W.L.

    1987-10-01

    The Applied Plasma Physics (APP) Theory program in the Office of Fusion Energy is charged with supporting the development of advanced physics models for fusion research. One such effort is ion cyclotron resonance heating (ICRH), which has seen substantial progress recently. However, due to serious questions about the adequacy of present models for CIT (Compact Ignition Tokamak), a Science Court was formed to assess ICRH models, including: validity of theoretical and computational approximations; underlying physics assumptions and corresponding limits on the results; self-consistency; any subsidiary issues needing resolution (e.g., new computer tools); adequacy of the models in simulating experiments (especially CIT); and new or improved experiments to validate and refine the models. The Court did not review work by specific individuals, institutions, or programs, thereby avoiding any biases along these lines. Rather, the Science Court was carefully structured as a technical review of ICRH theory and modeling in the US. This paper discusses the Science Court process, findings, and conclusions

  12. ICRF fundamental minority heating in inhomogeneous tokamak plasmas

    International Nuclear Information System (INIS)

    Romero, H.; Scharer, J.

    1987-01-01

    A theoretical model for the investigation of the ICRF fundamental minority plasma heating scheme in tokamak configurations is developed. The wave differential operator is obtained by including in a self-consistent manner the effects of strong wave damping, linear mode conversion and a one-dimensional non-uniform equilibrium configuration. It is found that the use of a self-consistent equilibrium distribution function yields important modifications of the ICRF wave differential operator applicable to this heating regime. In particular, the paper presents a set of new terms which are resonant at the fundamental cyclotron frequency and which ensure the self-adjointness of the resulting wave operator in the limit k parallel -> 0. A numerical scheme is developed with which solutions for the ICRF electromagnetic field and the corresponding power deposition and energy flux profiles can be obtained. An extensive parametric study is carried out for a range of wave and plasma parameters illustrative of current and proposed JET operating regimes. The results are considerably different from those obtained using a WKB fast wave model. In particular, the 'full wave' model presented in this paper yields a percentage for the wave power absorbed by the ionic species which is much larger than the one predicted by the WKB theory. The model presented also shows that the majority species can absorb a much higher proportion of the incident wave power than previously reported. Finally, the results obtained for JET indicate that in the case of low magnetic field incidence a sizeable percentage of the launched wave energy can be reflected on the fast wave branch for values of k parallel -1 and that at higher plasma temperatures electron heating becomes appreciable. (author)

  13. Microwave reflectometry for fusion plasma diagnostics

    International Nuclear Information System (INIS)

    1992-01-01

    This document contains a collection of 26 papers on ''Microwave Reflectometry for Fusion Plasma Diagnostics'', presented at the IAEA Technical Committee Meeting of the same name held at the JET Joint Undertaking, Abingdon, United Kingdom, March 4-6, 1992. It contains five papers on the measurement of plasma density profiles, six papers on theory and simulations in support of the development and application of this type of plasma diagnostics, eight papers on the measurement of density transients and fluctuations, and seven on new approaches to reflectometry-based plasma diagnostics. Refs, figs and tabs

  14. Mirror fusion test facility plasma diagnostics system

    International Nuclear Information System (INIS)

    Thomas, S.R. Jr.; Coffield, F.E.; Davis, G.E.; Felker, B.

    1979-01-01

    During the past 25 years, experiments with several magnetic mirror machines were performed as part of the Magnetic Fusion Energy (MFE) Program at LLL. The latest MFE experiment, the Mirror Fusion Test Facility (MFTF), builds on the advances of earlier machines in initiating, stabilizing, heating, and sustaining plasmas formed with deuterium. The goals of this machine are to increase ion and electron temperatures and show a corresponding increase in containment time, to test theoretical scaling laws of plasma instabilities with increased physical dimensions, and to sustain high-beta plasmas for times that are long compared to the energy containment time. This paper describes the diagnostic system being developed to characterize these plasma parameters

  15. Study of lower hybrid current drive system in tokamak fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Maebara, Sunao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    This report describes R and D of a high-power klystron, RF vacuum window, low-outgassing antenna and a front module for a plasma-facing antenna aiming the 5 GHz Lower Hybrid Current Drive (LHCD) system for the next Tokamak Fusion Device. 5 GHz klystron with a low-perveances of 0.7 {mu}P is designed for a high-power and a high-efficiency, the output-power of 715 kW and the efficiency of 63%, which are beyond the conventional design scaling of 450 kW-45%, are performed using the prototype klystron which operates at the pulse duration of 15 {mu}sec. A new pillbox window, which has an oversized length in both the axial and the radial direction, are designed to reduce the RF power density and the electric field strength at the ceramics. It is evaluated that the power capability by cooling edge of ceramics is 1 MW with continuous-wave operation. The antenna module using Dispersion Strengthened Copper which combines high mechanical property up to 500degC with high thermal conductivity, are developed for a low-outgassing antenna in a steady state operation. It is found that the outgassing rate is in the lower range of 4x10{sup -6} Pam{sup 3}/sm{sup 2} at the module temperature of 300degC, which requires no active vacuum pumping of the LHCD antenna. A front module using Carbon Fiber Composite (CFC) are fabricated and tested for a plasma facing antenna which has a high heat-resistive. Stationary operation of the CFC module with water cooling is performed at the RF power of 46 MWm{sup -2} (about 2 times higher than the design value) during 1000 sec, it is found that the outgassing rate is less than 10{sup -5} Pam{sup 3}/sm{sup 2} which is low enough for an antenna material. (author)

  16. 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.

  17. Real-time software for the COMPASS tokamak plasma control

    Czech Academy of Sciences Publication Activity Database

    Valcárcel, D.F.; Duarte, A.S.; Neto, A.; Carvalho, I.S.; Carvalho, B.B.; Fernandes, H.; Sousa, J.; Sartori, F.; Janky, Filip; Cahyna, Pavel; Hron, Martin; Pánek, Radomír

    2010-01-01

    Roč. 85, 3-4 (2010), s. 470-473 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition and Remote Participation for Fusion Research/7th./. Aix – en – Provence, 15.06.2009-19.06.2009] Institutional research plan: CEZ:AV0Z20430508 Keywords : Real-time * ATCA * Data acquisition * Plasma control software Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.143, year: 2010 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6V3C-4YXMP1Y-3&_user=6542793&_coverDate=07%2F31%2F2010&_rdoc=1&_fmt=high&_orig=search&_origin=search&_sort=d&_docanchor=&view=c&_acct=C000070123&_version=1&_urlVersion=0&_userid=6542793&md5=9005df0735c0dbb3a93a9c154b0d09c1&searchtype=a

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

    International Nuclear Information System (INIS)

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

    1991-12-01

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

  19. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    CERN Document Server

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  20. Neutron Activation Cool-down of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ascione, G.; Kugel, H.W.; Kumar, A.; Tilson, Jr, C.

    1998-06-10

    Tokamak Fusion Test Reactor (TFTR) final operations and post-shutdown neutron activation measurements were made. Ionization chambers were used to follow TFTR activation during operations and after shutdown. Gamma-ray energy spectroscopy measurements were performed to characterize TFTR activation at accessible vessel-bays and on sample hardware removed from structures at various distances from the vessel. The results demonstrate long-lived activations from common, commercially available materials used in the fabrication and field engineering of TFTR. The measurements allow characterization of residual TFTR neutron activation, the projection of residual activation decay, and benchmarking of low activation issues simulations.

  1. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    M.E. Lumia; C.A. Gentile

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed

  2. Overview of the STARFIRE reference commercial tokamak fusion power reactor design

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Barry, K.

    1980-01-01

    The purpose of the STARFIRE study is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup, superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield

  3. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    M.E. Lumia; C.A. Gentile

    2002-01-18

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  4. Fundamental studies of fusion plasmas

    Science.gov (United States)

    Aamodt, R. E.; Catto, P. J.; Dippolito, D. A.; Myra, J. R.; Russell, D. A.

    1992-05-01

    The major portion of this program is devoted to critical ICH phenomena. The topics include edge physics, fast wave propagation, ICH induced high frequency instabilities, and a preliminary antenna design for Ignitor. This research was strongly coordinated with the world's experimental and design teams at JET, Culham, ORNL, and Ignitor. The results have been widely publicized at both general scientific meetings and topical workshops including the specialty workshop on ICRF design and physics sponsored by Lodestar in April 1992. The combination of theory, empirical modeling, and engineering design in this program makes this research particularly important for the design of future devices and for the understanding and performance projections of present tokamak devices. Additionally, the development of a diagnostic of runaway electrons on TEXT has proven particularly useful for the fundamental understanding of energetic electron confinement. This work has led to a better quantitative basis for quasilinear theory and the role of magnetic vs. electrostatic field fluctuations on electron transport. An APS invited talk was given on this subject and collaboration with PPPL personnel was also initiated. Ongoing research on these topics will continue for the remainder of the contract period and the strong collaborations are expected to continue, enhancing both the relevance of the work and its immediate impact on areas needing critical understanding.

  5. 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)

  6. 10. LAWPP: Latin American workshop on plasma physics; 7. EBFP: Brazilian meeting on plasma physics. Abstracts

    International Nuclear Information System (INIS)

    2003-01-01

    Theoretical and experimental short communications are presented on plasma and fusion covering the following subjects: plasma production, confinement, plasma waves, diagnostics, heating, tokamak, impurities, astrophysics plasma and technological applications

  7. 3D field induced collisionless zonal flow decay in tokamak plasmas

    Science.gov (United States)

    Choi, G. J.; Hahm, T. S.

    2018-02-01

    We perform a gyrokinetic study of 3D resonant magnetic perturbation (RMP) field effect on zonal flows in tokamak plasmas. The result indicates that energy-dependent secular radial drift of particles due to the tangential component (δ B_\\Vert ) of the RMP field induces a further long term algebraic decay of the Rosenbluth–Hinton (1998 Phys. Rev. Lett. 80 724) residual flow expected in an axisymmetric tokamak. This mechanism could be related to an increase of H-mode transition power threshold in the presence of RMP observed in many tokamak experiments.

  8. Plasma transport in stochastic magnetic field caused by vacuum resonant magnetic perturbations at diverted tokamak edge

    International Nuclear Information System (INIS)

    Park, G.; Chang, C. S.; Joseph, I.; Moyer, R. A.

    2010-01-01

    A kinetic transport simulation for the first 4 ms of the vacuum resonant magnetic perturbations (RMPs) application has been performed for the first time in realistic diverted DIII-D tokamak geometry [J. Luxon, Nucl. Fusion 42, 614 (2002)], with the self-consistent evaluation of the radial electric field and the plasma rotation. It is found that, due to the kinetic effects, the stochastic parallel thermal transport is significantly reduced when compared to the standard analytic model [A. B. Rechester and M. N. Rosenbluth, Phys. Rev. Lett. 40, 38 (1978)] and the nonaxisymmetric perpendicular radial particle transport is significantly enhanced from the axisymmetric level. These trends agree with recent experimental result trends [T. E. Evans, R. A. Moyer, K. H. Burrell et al., Nat. Phys. 2, 419 (2006)]. It is also found, as a side product, that an artificial local reduction of the vacuum RMP fields in the vicinity of the magnetic separatrix can bring the kinetic simulation results to a more detailed agreement with experimental plasma profiles.

  9. Radiation-hardened x-ray imaging for burning-plasma tokamaks

    Science.gov (United States)

    Séguin, Fredrick H.; Petrasso, Richard D.; Li, Chi Kang

    1997-01-01

    A special type of vacuum-photodiode detector is being developed for x-ray imaging of plasma in fusion-producing tokamaks such as the international thermonuclear experimental reactor (ITER), where the radiation environment will be too hostile for conventional x-ray detectors. The vacuum photodiode has modest efficiency, but it is intrinsically immune to radiation damage if built in such a manner as to expose only metal components to radiation. A design based on appropriately chosen materials (including high-Z cathodes) and geometries (including a small angle between cathode surface and incident x rays) can provide good signals from the 1-100 keV x rays that are of particular importance for imaging the plasmas in the Joint European Torus (JET) and ITER. It should also provide better rejection of signal distortion and noise due to unwanted detection of neutrons and hard gamma rays than conventional detectors. A prototype design is described, along with performance parameters predicted for JET and ITER. In addition, we show results of laboratory experiments that confirm some of the predicted behaviors of the design.

  10. Report of the 1990 workshop on plasma-materials interactions for fusion research

    International Nuclear Information System (INIS)

    1991-03-01

    The 1990 Workshop of the Working Group in Research Committee on A and M Data, on plasma-materials interactions was held at the Headquarters of JAERI, Tokyo, on July 10-11, 1990. The aim of the Workshop was to obtain future prospects for the activities of the Working Group, by discussing current problems in plasma-materials interactions relevant to fusion research. This report contains all the 16 papers presented at hte Workshop, which are mainly concerned with problems in the first wall of large Tokamak devices, such as retention and release of hydrogen in carbon materials, sputtering and problems in material data system. (author)

  11. THE GENERAL ATOMICS FUSION THEORY PROGRAM ANNUAL REPORT FOR GRANT YEAR 2004

    International Nuclear Information System (INIS)

    PROJECT STAFF

    2004-01-01

    The dual objective of the fusion theory program at General Atomics (GA) is to significantly advance our scientific understanding of the physics of fusion plasmas and to support the DIII-D and other tokamak experiments. The program plan is aimed at contributing significantly to the Fusion Energy Science and the Tokamak Concept Improvement goals of the Office of Fusion Energy Sciences (OFES)

  12. THE GENERAL ATOMICS FUSION THEORY PROGRAM ANNUAL REPORT FOR GRANT YEAR 2004

    Energy Technology Data Exchange (ETDEWEB)

    PROJECT STAFF

    2004-12-01

    The dual objective of the fusion theory program at General Atomics (GA) is to significantly advance our scientific understanding of the physics of fusion plasmas and to support the DIII-D and other tokamak experiments. The program plan is aimed at contributing significantly to the Fusion Energy Science and the Tokamak Concept Improvement goals of the Office of Fusion Energy Sciences (OFES).

  13. Electron transport in the plasma edge with rotating resonant magnetic perturbations at the TEXTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Stoschus, Henning

    2011-10-13

    Small three-dimensional (3D) magnetic perturbations can be used as a tool to control the edge plasma parameters in magnetically confined plasmas in high confinement mode (''H-mode'') to suppress edge instabilities inherent to this regime, the Edge Localized Modes (ELMs). In this work, the impact of rotating 3D resonant magnetic perturbation (RMP) fields on the edge plasma structure characterized by electron density and temperature fields is investigated. We study a low confinement (L-mode) edge plasma (r/a>0.9) with high resistivity (edge electron collisionality {nu}{sup *}{sub e}>4) at the TEXTOR tokamak. The plasma structure in the plasma edge is measured by a set of high resolution diagnostics: a fast CCD camera ({delta}t=20 {mu}s) is set up in order to visualize the plasma structure in terms of electron density variations. A supersonic helium beam diagnostic is established as standard diagnostic at TEXTOR to measure electron density n{sub e} and temperature T{sub e} with high spatial ({delta}r=2 mm) and temporal resolution ({delta}t=20 {mu}s). The measured plasma structure is compared to modeling results from the fluid plasma and kinetic neutral transport code EMC3-EIRENE. A sequence of five new observations is discussed: (1) Imaging of electron density variations in the plasma edge shows that a fast rotating RMP field imposes an edge plasma structure, which rotates with the external RMP rotation frequency of vertical stroke {nu}{sub RMP} vertical stroke =1 kHz. (2) Measurements of the electron density and temperature provide strong experimental evidence that in the far edge a rotating 3D scrape-off layer (SOL) exists with helical exhaust channels to the plasma wall components. (3) Radially inward, the plasma structure at the next rational flux surface is found to depend on the relative rotation between external RMP field and intrinsic plasma rotation. For low relative rotation the plasma structure is dominated by a particle and energy loss

  14. Scattering of radio frequency waves by turbulence in fusion plasmas

    Science.gov (United States)

    Ram, Abhay K.

    2016-10-01

    In tokamak fusion plasmas, coherent fluctuations in the form of blobs or filaments and incoherent fluctuations due to turbulence are routinely observed in the scrape-off layer. Radio frequency (RF) electromagnetic waves, excited by antenna structures placed near the wall of a tokamak, have to propagate through the scrape-off layer before reaching the core of the plasma. While the effect of fluctuations on RF waves has not been quantified experimentally, there are telltale signs, arising from differences between results from simulations and from experiments, that fluctuations can modify the spectrum of RF waves. Any effect on RF waves in the scrape-off layer can have important experimental consequences. For example, electron cyclotron waves are expected to stabilize the deleterious neoclassical tearing mode (NTM) in ITER. Spectral and polarization changes due to scattering will modify the spatial location and profile of the current driven by the RF waves, thereby affecting the control of NTMs. Pioneering theoretical studies and complementary computer simulations have been pursued to elucidate the impact of fluctuations on RF waves. From the full complement of Maxwell's equations for cold, magnetized plasmas, it is shown that the Poynting flux in the wake of filaments develops spatial structure due to diffraction and shadowing. The uniformity of power flow into the plasma is affected by side-scattering, modifications to the wave spectrum, and coupling to plasma waves other than the incident RF wave. The Snell's law and the Fresnel equations have been reformulated within the context of magnetized plasmas. They are distinctly different from their counterparts in scalar dielectric media, and reveal new and important physical insight into the scattering of RF waves. The Snell's law and Fresnel equations are the basis for the Kirchhoff approximation necessary to determine properties of the scattered waves. Furthermore, this theory is also relevant for studying back

  15. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  16. Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

    International Nuclear Information System (INIS)

    Nishio, S.; Ueda, S.; Kurihara, R.; Kuroda, T.; Miura, H.; Sako, K.; Takase, H.; Seki, Y.; Adachi, J.; Yamazaki, S.; Hashimoto, T.; Mori, S.; Shinya, K.; Murakami, Y.; Senda, I.; Okano, K.; Asaoka, Y.; Yoshida, T.

    2000-01-01

    If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features of the prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900 deg. C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal

  17. Energy confinement of tokamak plasma with consideration of bootstrap current effect

    International Nuclear Information System (INIS)

    Yuan Ying; Gao Qingdi

    1992-01-01

    Based on the η i -mode induced anomalous transport model of Lee et al., the energy confinement of tokamak plasmas with auxiliary heating is investigated with consideration of bootstrap current effect. The results indicate that energy confinement time increases with plasma current and tokamak major radius, and decreases with heating power, toroidal field and minor radius. This is in reasonable agreement with the Kaye-Goldston empirical scaling law. Bootstrap current always leads to an improvement of energy confinement and the contraction of inversion radius. When γ, the ratio between bootstrap current and total plasma current, is small, the part of energy confinement time contributed from bootstrap current will be about γ/2

  18. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1997-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  19. Correlation measurements for fusion plasma diagnostics

    International Nuclear Information System (INIS)

    Pazsit, I.

    1995-01-01

    A list of a few methods for plasma diagnostics via fluctuations (noise) analysis of random (both temporally and spatially) system parameters is reviewed. Analogy is drawn with certain noise analysis methods, used in the diagnostics of fission reactors. These methods have been applied also to fusion measurements to some extent. However, the treatment of fusion plasma fluctuations is dominated by an approach that allows for temporal randomness, but assumes periodicity in space. This approach suits well a large class of phenomena such as magnetic fluctuations (MHD effects), but is much less suited to treat localised effects such as turbulence and density fluctuations. This paper discusses the potentials of the former approach, i.e. ordinary noise analysis methods of non-periodic variables in fusion plasma diagnostics. A new recommendation is to use the crossed beam correlation analysis of soft X-ray signals for determining the local short-range correlations in the plasma and to perform a systematic exploration of the plasma spatial correlation structure with that and other methods. 16 refs, 7 figs

  20. Design study of electrical power supply system for tokamak fusion power reactor

    International Nuclear Information System (INIS)

    1977-01-01

    Design study of the electrical power supply system for a 2000MWt Tokamak-type fusion reactor has been carried out. The purposes are to reveal and study problems in the system, leading to a plan of the research and development. Performed were study of the electrical power supply system and design of superconducting inductive energy storages and power switches. In study of the system, specification and capability of various power supplies for the fusion power reactor and design of the total system with its components were investigated. For the superconducting inductive energy storages, material choice, design calculation, and structural design were conducted, giving the size, weight and performance. For thyristor switches, circuit design in the parallel / series connection of element valves and cooling design were studied, providing the size and weight. (auth.)