Spherical tokamak development in Brazil

International Nuclear Information System (INIS)

Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group

2003-01-01

This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Spherical tokamak development in Brazil

Energy Technology Data Exchange (ETDEWEB)

Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

2003-07-01

The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Spherical tokamak development in Brazil

Energy Technology Data Exchange (ETDEWEB)

Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

2003-12-01

This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Spherical tokamak development in Brazil

International Nuclear Information System (INIS)

Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes

2003-01-01

The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

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

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

Research using small tokamaks

International Nuclear Information System (INIS)

1991-01-01

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

Physics and operation oriented activities in preparation of the JT-60SA tokamak exploitation

Science.gov (United States)

Giruzzi, G.; Yoshida, M.; Artaud, J. F.; Asztalos, Ö.; Barbato, E.; Bettini, P.; Bierwage, A.; Boboc, A.; Bolzonella, T.; Clement-Lorenzo, S.; Coda, S.; Cruz, N.; Day, Chr.; De Tommasi, G.; Dibon, M.; Douai, D.; Dunai, D.; Enoeda, M.; Farina, D.; Figini, L.; Fukumoto, M.; Galazka, K.; Galdon, J.; Garcia, J.; Garcia-Muñoz, M.; Garzotti, L.; Gil, C.; Gleason-Gonzalez, C.; Goodman, T.; Granucci, G.; Hayashi, N.; Hoshino, K.; Ide, S.; Imazawa, R.; Innocente, P.; Isayama, A.; Itami, K.; Joffrin, E.; Kamada, Y.; Kamiya, K.; Kawano, Y.; Kawashima, H.; Kobayashi, T.; Kojima, A.; Kubo, H.; Lang, P.; Lauber, Ph.; de la Luna, E.; Maget, P.; Marchiori, G.; Mastrostefano, S.; Matsunaga, G.; Mattei, M.; McDonald, D. C.; Mele, A.; Miyata, Y.; Moriyama, S.; Moro, A.; Nakano, T.; Neu, R.; Nowak, S.; Orsitto, F. P.; Pautasso, G.; Pégourié, B.; Pigatto, L.; Pironti, A.; Platania, P.; Pokol, G. I.; Ricci, D.; Romanelli, M.; Saarelma, S.; Sakurai, S.; Sartori, F.; Sasao, H.; Scannapiego, M.; Shimizu, K.; Shinohara, K.; Shiraishi, J.; Soare, S.; Sozzi, C.; Stępniewski, W.; Suzuki, T.; Suzuki, Y.; Szepesi, T.; Takechi, M.; Tanaka, K.; Terranova, D.; Toma, M.; Urano, H.; Vega, J.; Villone, F.; Vitale, V.; Wakatsuki, T.; Wischmeier, M.; Zagórski, R.

2017-08-01

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised.

The ETE spherical Tokamak project. IAEA report

Energy Technology Data Exchange (ETDEWEB)

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

2002-07-01

This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

Research using small tokamaks

International Nuclear Information System (INIS)

1993-01-01

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

Overview of JT-60U progress towards steady-state advanced tokamak

International Nuclear Information System (INIS)

Ide, S.

2005-01-01

Recent experimental results on steady state advanced tokamak (AT) research on JT-60U are presented with emphasis on longer time scale in comparison with characteristics time scales in plasmas. Towards this, modification on control in operation, heating and diagnostics systems have been done. As the results, ∼ 60 s I p flat top and an ∼ 30 s H-mode are obtained. The long pulse modification has opened a door into a new domain for JT-60U. The high normalized beta (β N ) of 2.3 is maintained for 22.3 s and 2.5 for 16.5 s in a high β p H-mode plasma. A standard ELMy H-mode plasma is also extended and change in wall recycling in such a longer time scale has been unveiled. Development and investigation of plasmas relevant to AT operation has been continued in former 15 s discharges as well in which higherNB power (≤ 10 s) is available. Higher β N ∼ 3 is maintained for 6.2 s in high β p H-mode plasmas. High bootstrap current fraction (f BS ) of ∼ 75% is sustained for 7.4 s in an RS plasma. On NTM suppression by localized ECCD, ECRF injection preceding the mode saturation is found to be more effective to suppress the mode with less power compared to the injection after the mode saturated. The domain of the NTM suppression experiments is extended to the high β N regime, and effectiveness of m/n=3/2 mode suppression by ECCD is demonstrated at β N ∼ 2.5-3. Genuine center-solenoid less tokamak plasma start up is demonstrated. In a current hole region, it is shown that no scheme drives a current in any direction. Detailed measurement in both spatial and energy spaces of energetic ions showed dynamic change in the energetic ion profile at collective instabilities. Impact of toroidal plasma rotation on ELM behaviors is clarified in grassy ELM and QH domains. (author)

Library system for a one dimensional tokamak transport code: (LIBJT60), 1

International Nuclear Information System (INIS)

Hirayama, Toshio

1982-12-01

A library system is developed to control and manage huge programs in terms of FORTRAN source. It is applied to widely used one dimensional tokamak transport codes (LIBJT60), which have been developed in the Division of Large Tokamak Development. The structure of data and program in the transport code turn out to be flexible enough to respond to various demands and this gigantic code frame work can be decomposed into groups of a compact code with a specific function. Some editing support tools for programming and debugging are also developed to save programming work. By applying this library system, users can obtain a code whose functions can be efficiently developed. (author)

Supravodivý tokamak dobyl Asii

Czech Academy of Sciences Publication Activity Database

Řípa, Milan

2006-01-01

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

Mechanical properties of JT-60 tokamak machine in power tests

International Nuclear Information System (INIS)

Takatsu, Hideyuki; Ohkubo, Minoru; Yamamoto, Masahiro; Ohta, Mitsuru

1986-01-01

JT-60 power tests were carried out from Dec. 10, 1984 to Feb. 20, 1985 to demonstrate, in advance of actual plasma operation, satisfactory performance of tokamak machine, power suppliers and control system in combination. The tests began with low power test of individual coil systems and progressed to full power tests. The coil current was raised step by step, monitoring the mechanical, thermal, electrical and vacuum data. Power tests were concluded with successful results. All of the coil systems were raised up to full power operation in combination and system performance was verified including the structural integrity of tokamak machine. Measured strain and deflection showed good agreements with those predicted in the design, which was an evidence that electromagnetic forces were supported as expected in the design. A few limitations to machine operation was made clear quantitatively. And it was found that existing detectors were insufficient to monitor machine integrity and two kinds of detector were proposed to be installed. (author)

Tokamaks (Second Edition)

Energy Technology Data Exchange (ETDEWEB)

Stott, Peter [JET, UK (United Kingdom)

1998-10-01

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

Tokamaks (Second Edition)

International Nuclear Information System (INIS)

Stott, Peter

1998-01-01

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

ARIES tokamak reactor study

International Nuclear Information System (INIS)

Steiner, D.; Embrechts, M.

1990-07-01

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

Recent results and near-term expectations in Tokamak fusion research in the U.S., Europe, and Japan

International Nuclear Information System (INIS)

Meade, D.

1993-01-01

The development of fusion is often thought about in terms of three different activities: scientific feasibility, engineering feasibility, and economic feasibility. This paper discusses the scientific feasibility of fusion. Reactor temperatures, reactor densities and confinement, particle control, plasma power handling, and self-heating are some of the issues examined. Collaboration and results from research at the Tokamak Fusion Test Reactor (TFTR) at Princeton, the JT-60U in Japan, and JET, the Joint European Torus Tokamak in Oxford are presented

Status of tokamak research

International Nuclear Information System (INIS)

Rawls, J.M.

1979-10-01

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

Advanced fusion technologies developed for JT-60 superconducting tokamak

International Nuclear Information System (INIS)

Sakasai, Akira; Ishida, S.; Matsukawa, M.

2003-01-01

The modification of JT-60U is planned as a full superconducting tokamak (JT-60SC). The objectives of the JT-60SC program are to establish scientific and technological bases for the steady-state operation of high performance plasmas and utilization of reduced-activation materials in economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to DEMO reactor have been developed in the superconducting magnet technology and plasma facing components for the design of JT-60SC. To achieve a high current density in a superconducting strand, Nb 3 Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFC) of JT-60SC. The R and D to demonstrate applicability of Nb 3 Al conductor to the TFC by a react-and-wind technique have been carried out using a full-size Nb 3 Al conductor. A full-size NbTi conductor with low AC loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the CFC target was successfully demonstrated on the electron beam irradiation stand. (author)

Research tokamak system with multi-mode discharges using inverter power supply

International Nuclear Information System (INIS)

Kojima, Hiroki; Kobayashi, Masahiro; Takagi, Makoto; Takamura, Shuichi; Tashiro, Kenji

1999-01-01

In Current Sustaining Tokamak in Nagoya university (CSTN)-IV research tokamak system using a compact 40kHz pulse width modulation (PWM) inverter power supply, which is controlled through LabVIEW program, we construct a new tokamak discharge system with multi-mode including a stable alternating current discharge and a high-repetition high-duty one. These discharge modes can be operated continuously for as long as 60sec. The continuous discharge with long duration is able to simulate the important physical and chemical processes of long time discharges in fusion devices, in which the heat load to the wall and the particle balance in the plasma-wall system are crucial topics in order to realize a long pulse fusion reactor, like ITER. Employing ergodic divertor (ED) is one of tools to control the particle balance and the heat load to the wall. In addition, we installed another inverter power supply to generate a rotating magnetic perturbation for dynamic ergodic divertor (DED) with the appropriate measurement system so that we may carry out experiments on heat and particle control with DED at long time operation. (author)

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.)

Full power in the European tokamak

International Nuclear Information System (INIS)

Lallia, P.P.; Hugon, M.

1987-01-01

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

The life test of a DC circuit breaker of tokamak device JT-60 for a nuclear fusion research

International Nuclear Information System (INIS)

Shimada, Ryuichi; Tani, Keiji; Kishimoto, Hiroshi; Tamura, Sanae; Yanabu, Satoru.

1979-01-01

In the Tokamak devices for nuclear fusion research, the construction of the current transformer circuits having plasma as the secondary circuit and the change of the primary circuit current are necessary for generating current in the plasma. This is considered to be fairly difficult in practice if conventional methods using capacitor discharge and iron core coils are employed. Considering such circumstances, it was decided for JT-60 to use an air-core current transformer coil and to employ the method of storing energy in the form of current in the coil inductance instead of a capacitor. For this reason, a DC circuit breaker is required to interrupt coil current. The authors improved an AV vacuum breaker, which had been developed as the vacuum breaker of longitudinal magnetic field type applying a magnetic field in parallel with an arc, to get the one for DC circuit for the purpose of applying it to JT-60. In this paper, the operational characteristic of the DC breaker is described, the construction and function of the life test circuit is explained, and the test results are reported. Finally, interruptions of 10,000 times at 20 kA were carried out. It is successful that the restrike of arc occurring during tens of milli-seconds after interruptions was improved to 0.05% or less for 10,000 times operations. Further, it was found that the generation of arc restrike can be reduced practically to zero with two breakers in series. (Wakatsuki, Y.)

An overview on plasma disruption mitigation and avoidance in tokamak

International Nuclear Information System (INIS)

He Kaihui; Pan Chuanhong; Feng Kaiming

2002-01-01

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

Neutronics design of the next tokamak. (Swimming pool type)

International Nuclear Information System (INIS)

Seki, Y.; Iida, H.; Kitamura, K.; Minato, A.; Sako, K.; Mori, S.; Nishida, H.

1983-01-01

A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

Design and Structural Analysis for the Vacuum Vessel of Superconducting Tokamak JT-60SC

International Nuclear Information System (INIS)

Kudo, Y.; Sakurai, S.; Masaki, K.; Urata, K.; Sasajima, T.; Matsukawa, M.; Sakasai, A.; Ishida, S.

2003-01-01

A modification of the JT-60 is planned to be a superconducting tokamak (JT-60SC) in order to establish steady-state operation of high beta plasma for 100 s, and to ensure the applicability of ferritic steel as a reduced activation material for reactor relevant break-even class plasmas. This paper describes the detailed design of the vacuum vessel, which has a unique structure for cost effective manufacturing, as well as structural analysis results for a feasibility study

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11-12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of

The basics of spherical tokamaks and progress in European research

International Nuclear Information System (INIS)

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

2003-01-01

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

Results of Joint Experiments and other IAEA activities on research using small tokamaks

Czech Academy of Sciences Publication Activity Database

Brotánková, Jana; Dejarnac, Renaud; Dufková, Edita; Ďuran, Ivan; Hron, Martin; Sentkerestiová, Jana; Stöckel, Jan; Weinzettl, Vladimír; Zajac, Jaromír

2009-01-01

Roč. 49, č. 10 (2009), s. 104026-104026 ISSN 0029-5515. [IAEA Fusion Energy Conference/22nd./. Geneva, 13.10.2008-18.10.2008] R&D Projects: GA AV ČR KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * probe diagnostics * sheared flows * edge plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://iopscience.iop.org/0029-5515/49/10/104026

Validation of neutral point on JT-60U, Alcator C-Mod and ASDEX-Upgrade tokamaks

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Yoshino, Ryuji; Pautasso, Gabriella; Gruber, Otto; Jardin, Stephen

2002-01-01

Validation studies of a neutrally balanced vertical plasma position, so-called ''neutral point'', have been carried out by computational simulations and experiments under trilateral Japan-US-EU collaborations. It was clarified that the neutral point, where VDEs (Vertical Displacement Events) are hardly occurred, does exit in the Alcator C-Mod and ASDEX-Upgrade tokamaks as well as the JT-60U, consistent with the simulations. Meanwhile, precise details of the VDE behavior exhibit their own characters according to the individual of the tokamaks such as an up-down asymmetry of plasma shape. Sensitivity of the neutral point to the plasma shape and current profile was also addressed in detail. (author)

Tokamak devices: towards controlled fusion

International Nuclear Information System (INIS)

Trocheris, M.

1975-01-01

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

Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

International Nuclear Information System (INIS)

Kawashima, H.; Shimizu, K.; Takizuka, T.

2007-01-01

To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

Recent Activities on the Experimental Research Programme Using Small Tokamaks

International Nuclear Information System (INIS)

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

2006-01-01

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

Present status of design, research and development of nuclear fusion reactors and problems

International Nuclear Information System (INIS)

1983-04-01

Seven years have elapsed since the publication of ''Progress of nuclear fusion research and perspective toward the development of power reactors'' by the Atomic Energy Society of Japan in August, 1976. During this period, the research and development of nuclear fusion have changed from plasma physics to reactor technology, being conscious of the realization of fusion reactors. There are the R project in the Institute of Plasma Physics, Nagoya University, and the design and construction of JT-60 in Japan Atomic Energy Research Institute, to put it concretely. Now the research and development taking the economical efficiency into account are adopted. However, the type of fusion reactors is not reduced to tokamak type, accordingly the research and development to meet the diverse possibilities are forwarded. The progress of tokamak reactor research, core plasma design, nuclear design and shielding design, thermal structure design, the design of superconducting magnets, disassembling and repair, safety, economical efficiency, the conceptual design of other types than tokamak and others are reported. (Kako, I.)

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

International Nuclear Information System (INIS)

Lopez C, R.

1986-07-01

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

Varennes Tokamak

International Nuclear Information System (INIS)

Cumyn, P.B.

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Research using small tokamaks

International Nuclear Information System (INIS)

1991-05-01

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

36 CFR 1256.60 - Information relating to financial institutions.

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2010-07-01

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Review of ICRF antenna development and heating experiments up to advanced experiment I, 1989 on the JT-60 tokamak

International Nuclear Information System (INIS)

Fujii, Tsuneyuki

1992-03-01

Two main subjects of ion cyclotron range of frequencies (ICRF) heating on JT-60 are described in this paper from development phase of the JT-60 ICRF heating system up to advanced experiment I, 1989. One is antenna design and development for the high power JT-60 ICRF heating system (6 MW for 10 s at a frequency range of 108 - 132 MHz). The other is the experimental investigation of characteristics of second harmonic ICRF heating in a large tokamak. (J.P.N.)

Design concepts and performance tests of the 60 GHz electron cyclotron heating (ECH) system for the JFT-2M tokamak

International Nuclear Information System (INIS)

Hoshino, Katsumichi; Yamamoto, Takumi; Kawashima, Hisato; Shibata, Takatoshi; Shibuya, Toshihiro

1985-11-01

60 GHz overmoded microwave launch system for the JFT-2M tokamak is described. The basic design concepts, specifications of each microwave component and the results of the performance tests are reported. The transmission of the microwave power is done in the circular TE 01 mode which has a low loss along the overmoded circular transmission components of 33 m in length. The microwave power of 80 - 90 kW, pulse width 100 ms in the circular TE 11 mode is finally launched into the JFT-2M tokamak plasma. (author)

Very fast feedback control of coil-current in JT-60 tokamak

International Nuclear Information System (INIS)

Aoyagi, T.; Terakado, T.; Takahashi, M.; Nobusaka, H.; Yagyu, J.; Matsuzaki, Y.

1992-01-01

A direct digital control (DDC) system is adopted for controlling thyristor converters of power supplies in the JT-60 tokamak built in 1984. Microcomputers of the DDC were 5 MHz i8086 microprocessor and programs were written by assembler language and the processing time was under 1ms. They were, however, too old in hardware and too complicated in software. New DDC system has been made in the JT-60 Upgrade (JT-60U) to control the power supplies more quickly under 0.25 and 0.5 ms of the processing time and also to write the programs used by high-level language. The new system consists of a host computer and five microcomputers with microprocessor on VME bus system. The host computer AS3260 performs on-line processing such as setting the DDC under the discharge conditions and so on. Functions of the microcomputers with a 32-bit, 20 MHz microprocessor MC68030, whose OS are VxWorks and programs are written by C language, are real-time processing such as taking in instructions from a ZENKEI computer and in feedback control of currents and voltages of coils every 0.25 and 0.5 ms. The system is now operating very smoothly. (author)

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

International Nuclear Information System (INIS)

2005-01-01

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

Research using small tokamaks. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1997-09-01

The technical reports in these proceedings were presented at the IAEA Technical Committee Meeting on research Using Small Tokamaks, held in Ahmedabad, India, 6-7 December 1995. The purpose of this annual meeting is to provide a forum for the exchange of information on various small and medium sized plasma experiments, not only for tokamaks. The potential benefits of these research programmes are to: test theories, such as effects of the plasma rotation; check empirical scalings, such as density limits; develop fusion technology hardware; develop plasma diagnostics; such as tomography; and to train scientists, engineers, technicians, and students, particularly in developing IAEA Member States

Mechanical and thermal characteristics of JT-60 tokamak machine demonstrated in its power tests

International Nuclear Information System (INIS)

Takatsu, Hideyuki; Yamamoto, Masahiro; Ohkubo, Minoru

1985-09-01

JT-60 power tests were carried out from Dec. 10, 1984 to Feb. 20, 1985 to demonstrate, in advance of actual plasma operation, satisfactory performance of tokamak machine, power suppliers and control system in combination. The tests began with low power test of individual coil systems and progressed to full power tests. Power tests were successfully concluded with the following conclusions. (1) All of the coil systems were raised up to full power operation in combination and system performance was verified including thermal and structural integrity of tokamak machine. (2) Measured strain and deflection showed good agreements with those predicted in the design, which was an evidence that electromagnetic loads were supported adequately as expected in the design. (3) Vibration of lateral port was found to be large up to 50 m/s 2 and caused excessive vibration of gate-valves. (4) A few limitations to machine operation were made clear quantatively. (5) It was found that the existing detectors were insufficient to monitor the machine integrity and a few kinds of detectors were necessary to be installed. (author)

Requirements for tokamak remote operation: Application to JT-60SA

International Nuclear Information System (INIS)

Innocente, Paolo; Barbato, Paolo; Farthing, Jonathan; Giruzzi, Gerardo; Ide, Shunsuke; Imbeaux, Frédéric; Joffrin, Emmanuel; Kamada, Yutaka; Kühner, Georg; Naito, Osamu; Urano, Hajime; Yoshida, Maiko

2015-01-01

Highlights: • We analyzed the data management system (DMS) appropriate for international collaboration. • We define the principal requirements for all components of the DMS. • We evaluated application of DMS requirements to the JT-60SA experiment. • We evaluated the role network bandwidth and time delay between EU and Japan. - Abstract: Remote operation and data analysis are becoming key requirements of any fusion devices. In this framework a well-conceived data management system integrated with a suite of analysis and support tools are essential components for an efficient remote exploitation of any fusion device. The following components must be considered: data archiving data model architecture; remote data and computers access; pulse schedule, data analysis software and support tools; remote control room specifications and security issues. The definition of a device-generic data model plays also important role in improving the ability to share solution and reducing learning time. As for the remote control room, the implementation of an Operation Request Gateway has been identified as an answer to security issues meanwhile remotely proving all the required features to effectively operate a device. Previous requirements have been analyzed for the new JT-60SA tokamak device. Remote exploitation is paramount in the JT-60SA case which is expected to be jointly operated between Japan and Europe. Due to the geographical distance of the two parties an optimal remote operation and remote data-analysis is considered as a key requirement in the development of JT-60SA. It this case the effects of network speed and delay have been also evaluated and tests have confirmed that the performance can vary significantly depending on the technology used.

Requirements for tokamak remote operation: Application to JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Innocente, Paolo, E-mail: paolo.innocente@igi.cnr.it [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Barbato, Paolo [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Farthing, Jonathan [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Giruzzi, Gerardo [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Ide, Shunsuke [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Imbeaux, Frédéric; Joffrin, Emmanuel [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Kamada, Yutaka [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Kühner, Georg [Max-Planck-Institute for Plasma Physics, EURATOM Association, Wendelsteinstr. 1, 17491 Greifswald (Germany); Naito, Osamu; Urano, Hajime; Yoshida, Maiko [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

2015-10-15

Highlights: • We analyzed the data management system (DMS) appropriate for international collaboration. • We define the principal requirements for all components of the DMS. • We evaluated application of DMS requirements to the JT-60SA experiment. • We evaluated the role network bandwidth and time delay between EU and Japan. - Abstract: Remote operation and data analysis are becoming key requirements of any fusion devices. In this framework a well-conceived data management system integrated with a suite of analysis and support tools are essential components for an efficient remote exploitation of any fusion device. The following components must be considered: data archiving data model architecture; remote data and computers access; pulse schedule, data analysis software and support tools; remote control room specifications and security issues. The definition of a device-generic data model plays also important role in improving the ability to share solution and reducing learning time. As for the remote control room, the implementation of an Operation Request Gateway has been identified as an answer to security issues meanwhile remotely proving all the required features to effectively operate a device. Previous requirements have been analyzed for the new JT-60SA tokamak device. Remote exploitation is paramount in the JT-60SA case which is expected to be jointly operated between Japan and Europe. Due to the geographical distance of the two parties an optimal remote operation and remote data-analysis is considered as a key requirement in the development of JT-60SA. It this case the effects of network speed and delay have been also evaluated and tests have confirmed that the performance can vary significantly depending on the technology used.

Study of grounding system of large tokamak device JT-60

International Nuclear Information System (INIS)

Arakawa, Kiyotsugu; Shimada, Ryuichi; Kishimoto, Hiroshi; Yabuno, Kohei; Ishigaki, Yukio.

1982-01-01

In the critical plasma testing facility JT-60 constructed by the Japan Atomic Energy Research Institute, high voltage, large current is required in an instant. Accordingly, for the protection of human bodies and the equipment, and for realizing the stable operation of the complex, precise control and measurement system, a large scale facility of grounding system is required. In case of the JT-60 experimental facility, the equipments with different functions in separate buildings are connected, therefore, it is an important point to avoid high potential difference between buildings. In the grounding system for the JT-60, a reticulate grounding electrode is laid for each building, and these electrodes are connected with a low impedance metallic duct called grounding trunk line. The power supply cables for various magnetic field coils, control lines and measurement lines are laid in the duct. It is a large problem to grasp quantitatively the effect of a grounding trunk line by analysis. The authors analyzed the phenomenon that large current flows into a grounding system by lightning strike or grounding. The fundamental construction of the grounding system for the JT-60, the condition for the analysis and the result of simulation are reported. (Kako, I.)

Neutronic analysis of fusion tokamak devices by PHITS

International Nuclear Information System (INIS)

Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

2011-01-01

A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

[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

Tokamak experiments

International Nuclear Information System (INIS)

Robinson, D.C.

1987-01-01

With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

Calibration of power systems and measurements of discharge currents generated for different coils in the EGYPTOR tokamak

Czech Academy of Sciences Publication Activity Database

Hegazy, H.; Žáček, František

2006-01-01

Roč. 25, 1-2 (2006), s. 73-86 ISSN 0164-0313 Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * EGYPTOR tokamak * Rogowski coil Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.381, year: 2006

Tokamaks. 2. ed.

International Nuclear Information System (INIS)

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

1997-01-01

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

Controlled thermonuclear fusion and the latest progress on China's HT-7 superconducting tokamak

International Nuclear Information System (INIS)

Li Jiangang; Yang Yu

2003-01-01

After 50 years of research on controlled thermonuclear fusion, a new stage will be reached in 2003, when a site for the International Thermonuclear Experimental Reactor project will be chosen to start the construction. Scientists hope that this project could herald a new era in which the energy problem will be solved completely. The great progress made on the HT-7 superconducting tokamak in China has provided positive and powerful support for fusion research. The HT-7 is one of the only two superconducting tokamaks in the world that can carry out minute-scale high temperature plasma research, and has achieved a duration of 63.95s for the hot plasma discharge. This is a major step towards real steady-state operation of the tokamak configuration. We present an overview of the latest progress on the tokamak experiments in the Institute of Plasma Physics, Chinese Academy of Sciences

Plasma regimes and research goals of JT-60SA towards ITER and DEMO

International Nuclear Information System (INIS)

Kamada, Y.; Ide, S.; Fujita, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; Barabaschi, P.; Lackner, K.; Ishida, S.; Bolzonella, T.

2011-01-01

The JT-60SA device has been designed as a highly shaped large superconducting tokamak with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) in order to satisfy the central research needs for ITER and DEMO. In the ITER- and DEMO-relevant plasma parameter regimes and with DEMO-equivalent plasma shapes, JT-60SA quantifies the operation limits, plasma responses and operational margins in terms of MHD stability, plasma transport and confinement, high-energy particle behaviour, pedestal structures, scrape-off layer and divertor characteristics. By integrating advanced studies in these research fields, the project proceeds 'simultaneous and steady-state sustainment of the key performances required for DEMO' with integrated control scenario development applicable to the highly self-regulating burning high-β high bootstrap current fraction plasmas.

[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

Helical-type device and laser fusion. Rivals for tokamak-type device at n-fusion development in Japan

International Nuclear Information System (INIS)

Anon.

1994-01-01

Under the current policy on the research and development of nuclear fusion in Japan, as enunciated by the Atomic Energy Commission of Japan, the type of a prototype fusion reactor will be chosen after 2020 from tokamak, helical or some other type including the inertial confinement fusion using lasers. A prototype fusion reactor is the next step following the tokamak type International Thermonuclear Experimental Reactor (ITER). With the prototype reactor, the feasibility as a power plant will be examined. At present the main research and development of nuclear fusion in Japan are on tokamak type, which have been promoted by Japan Atomic Energy Research Institute (JAERI). As for the other types of nuclear fusion, researches have been carried out on the helical type in Kyoto University and National Institute for Fusion Science (NIFS), the mirror type in Tsukuba University, the tokamak type using superconductive coils in Kyushu University, and the laser fusion in Osaka University. The features and the present state of research and development of the Large Helical Device and the laser fusion which is one step away from the break-even condition are reported. (K.I.)

Neutral atom analyzers for diagnosing hot plasmas: A review of research at the ioffe physicotechnical institute

International Nuclear Information System (INIS)

Kislyakov, A. I.; Petrov, M. P.

2009-01-01

Research on neutral particle diagnostics of thermonuclear plasmas that has been carried out in recent years at the Ioffe Physicotechnical Institute of the Russian Academy of Sciences (St. Petersburg, Russia) is reviewed. Work on the creation and improvement of neutral atom analyzers was done in two directions: for potential applications (in particular, on the International Thermonuclear Experimental Reactor, which is now under construction at Cadarache in France) and for investigation of the ion plasma component in various devices (in particular, in the largest tokamaks, such as JET, TFTR, and JT-60). Neutral atom analyzers are the main tool for studying the behavior of hydrogen ions and isotopes in magnetic confinement systems. They make it possible to determine energy spectra, to perform the isotope analysis of atom fluxes from the plasma, to measure the absolute intensity of the fluxes, and to record how these parameters vary with time. A comparative description of the analyzers developed in recent years at the Ioffe Institute is given. These are ACORD-12/24 analyzers for recording 0.2-100-keV hydrogen and deuterium atoms with a tunable range of simultaneously measured energies, CNPA compact analyzers for a fixed energy gain in the ranges 80-1000 eV and 0.8-100 keV, an ISEP analyzer for simultaneously recording the atoms of all the three hydrogen isotopes (H, D, and T) in the energy range 5-700 keV, and GEMMA analyzers for recording atom fluxes of hydrogen and helium isotopes in the range 0.1-4 MeV. The scintillating detectors of the ISEP and GEMMA analyzers have a lowered sensitivity to neutrons and thus can operate without additional shielding in neutron fields of up to 10 9 n/(cm 2 s). These two types of analyzers, intended to operate under deuterium-tritium plasma conditions, are prototypes of atom analyzers created at the Ioffe Institute for use in the International Thermonuclear Experimental Reactor. With these analyzers, a number of new results have been

COMPASS Tokamak in Czech Republic now up and running

Czech Academy of Sciences Publication Activity Database

Mlynář, Jan

2009-01-01

Roč. 3, č. 1 (2009), s. 10-10 ISSN 1818-5355 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * EURATOM Subject RIV: BL - Plasma and Gas Discharge Physics http://www.efda.org/news_and_events/downloads/efda_newsletter/nl_2009_05.pdf

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

TBR-1 (Brazilian Tokamak) - Recent Results

International Nuclear Information System (INIS)

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

1987-08-01

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

FIR-laser scattering for JT-60

International Nuclear Information System (INIS)

Itagaki, Tokiyoshi; Matoba, Tohru; Funahashi, Akimasa; Suzuki, Yasuo

1977-09-01

An ion Thomson scattering method with far infrared (FIR) laser has been studied for measuring the ion temperature in large tokamak JT-60 to be completed in 1981. Ion Thomson scattering has the advantage of measuring spatial variation of the ion temperature. The ion Thomson scattering in medium tokamak (PLT) and future large tokamak (JET) requires a FIR laser of several megawatts. Research and development of FIR high power pulse lasers with power up to 0.6 MW have proceeded in ion Thomson scattering for future high-temperature tokamaks. The FIR laser power will reach to the desired several megawatts in a few years, so JAERI plans to measure the ion temperature in JT-60 by ion Thomson scattering. A noise source of the ion Thomson scattering with 496 μm-CH 3 F laser is synchrotron radiation of which the power is similar to NEP of the Schottky-barrier diode. However, the synchrotron radiation power is one order smaller than that when a FIR laser is 385 μm-D 2 O laser. The FIR laser power corresponding to a signal to noise ratio of 1 is about 4 MW for CH 3 F laser, and 0.4 MW for D 2 O laser if NEP of the heterodyne mixer is one order less. A FIR laser scattering system for JT-60 should be realized with improvement of FIR laser power, NEP of heterodyne mixer and reduction of synchrotron radiation. (auth.)

The recent research progress on the J-TEXT tokamak

International Nuclear Information System (INIS)

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

2013-01-01

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

Southwestern Institute of Physics annual report (2000)

International Nuclear Information System (INIS)

2001-01-01

The research results and engineering progress of SWIP (Southwestern Institute of Physics) during the year of 2000 was summarized in this annual report. The contents divided into five parts: 1. tokamak experimental diagnoses and tokamak engineering; 2. fusion reactor and fusion reactor materials; 3. plasma theory and calculation; 4. technique development and application; 5. appendix 31 theses and presented in this report

Short-term power sources for tokamaks and other physical experiments

Czech Academy of Sciences Publication Activity Database

Zajac, Jaromír; Žáček, František; Brettschneider, Zbyněk; Lejsek, V.

2007-01-01

Roč. 82, č. 4 (2007), s. 369-379 ISSN 0920-3796 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * Impulse power sources * Energy accumulation Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 1.058, year: 2007 http://www.sciencedirect.com/science/journal/09203796

Tokamak GOLEM se vydává do světa

Czech Academy of Sciences Publication Activity Database

Svoboda, V.; Jex, I.; Žára, J.; Stöckel, Jan; Mlynář, Jan

-, 01 (2012), s. 18-19 ISSN 1213-5348 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * remote participation * training Subject RIV: BL - Plasma and Gas Discharge Physics http://jaderka.fjfi.cvut.cz/ sites /default/files/attachment/ptgolem_0.pdf

Heat load material studies: Simulated tokamak disruptions

International Nuclear Information System (INIS)

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

1991-01-01

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

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.)

Fusion energy research, the tokamak of CRPP-EPFL, electrotechnical equipment

International Nuclear Information System (INIS)

1993-01-01

The topics of this information meeting were: fusion energy research at CRPP, the TCV tokamak, an alternating current generator which does not stress the grid, AC/DC multi-megawatt converters, stabilisation of the plasma, a fast and modular power AC/DC converter. figs., tabs., refs

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

International Nuclear Information System (INIS)

Isayama, Akihiko; Tanaka, Hitoshi

2009-01-01

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

Cluster storage for COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

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

2012-01-01

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

Fusion research program in Korea

International Nuclear Information System (INIS)

Hwang, Y.S.

1996-01-01

Fusion research in Korea is still premature, but it is a fast growing program. Groups in several universities and research institutes were working either in small experiments or in theoretical areas. Recently, couple of institutes who have small fusion-related experiments, proposed medium-size tokamak programs to jump into fusion research at the level of international recognition. Last year, Korean government finally approved to construct 'Superconducting Tokamak' as a national fusion program, and industries such as Korea Electric Power Corp. (KEPCO) and Samsung joined to support this program. Korea Basic Science Institute (KBSI) has organized national project teams including universities, research institutes and companies. National project teams are performing design works since this March. (author)

Summary report on tokamak confinement experiments

International Nuclear Information System (INIS)

1982-03-01

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

Multi-mode remote participation on the GOLEM tokamak

Czech Academy of Sciences Publication Activity Database

Svoboda, V.; Huang, B.; Mlynář, Jan; Pokol, G.I.; Stöckel, Jan; Vondrášek, G.

2011-01-01

Roč. 86, 6-8 (2011), s. 1310-1314 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT) /26th./. Porto, 27.09.2010-01.10.2010] Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * remote participation * education Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.490, year: 2011 http://www.sciencedirect.com/science/article/pii/S0920379611002390

Alcator C-Mod Tokamak

Data.gov (United States)

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

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

International Nuclear Information System (INIS)

Navratil, G.A.

1983-08-01

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

Annual report of Naka Fusion Research Establishment from April 1, 2002 to March 31, 2003

International Nuclear Information System (INIS)

Tsuji, Hiroshi; Hamamatsu, Kiyotaka; Matsumoto, Hiroshi; Yoshida, Hidetoshi

2003-11-01

This annual report provides an overview of research and development (R and D) activities at Naka Fusion Research Establishment, including those performed in collaboration with other research establishments of JAERI, research institutes, and universities, during the period from 1 April, 2002 to 31 March, 2003. The activities in the Naka Fusion Research Establishment are highlighted by high performance plasma researches in JT-60 and JFT-2M, research and development of fusion reactor technologies towards ITER and fusion power demonstration plants, and activities in support of ITER design and construction. JT-60 program has continued to produce fruitful knowledge and understanding necessary to achieve reactor relevant performances of tokamak fusion devices. JFT-2M has made contributions in more basic areas of tokamak plasma research and development in pursuit of high performance plasma. The objectives of JT-60 research have been more shifted to physics R and Ds in support of the International Thermonuclear Experimental Reactor (ITER) and establishment of physics basis for a steady state tokamak fusion reactor like SSTR as a fusion power demonstration plant. In JFT-2M, the advanced material tokamak experiment program has been carried out to test the low activation ferritic steel for development of the structural material for a fusion reactor. In the area of theories and analyses, significant progress has been made in understanding of the ITB, energy confinement scaling in ITB plasmas, MHD equilibrium in the current hole region, asymmetric feature of divertor plasmas and the divertor detachment. In addition, through the project of numerical experiment on tokamak, the mechanism of the ion temperature gradient mode was clarified by particle simulations. The physics of divertor plasma was also studied by particle simulations. R and Ds of fusion reactor technologies have been carried out both to further improve technologies necessary for ITER construction, and to accumulate

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)

What is past is prologue: future directions in tokamak power reactor design research

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

Conceptual tokamak power reactor designs over the last five years have provided us with many fundamental insights regarding tokamaks as fusion reactors. This first generation of studies has helped lay the groundwork upon which to build improvements in reactor design and begin a process of optimization. After reviewing the first generation of studies and the primary conclusions they produced, we discuss four current designs that are representative of present trends in this area of research. In particular, we discuss the trends towards reduced reactor size and higher neutron wall loadings. Moving in this direction requires new approaches to many subsystem designs. We describe new approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets. We close with a discussion of the future role of conceptual reactor design research and the need for close interaction with ongoing experiments in fusion technology

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

User's manual of JT-60 experimental data analysis system

International Nuclear Information System (INIS)

Hirayama, Takashi; Morishima, Soichi; Yoshioka, Yuji

2010-02-01

In the Japan Atomic Energy Agency Naka Fusion Institute, a lot of experiments have been conducted by using the large tokamak device JT-60 aiming to realize fusion power plant. In order to optimize the JT-60 experiment and to investigate complex characteristics of plasma, JT-60 experimental data analysis system was developed and used for collecting, referring and analyzing the JT-60 experimental data. Main components of the system are a data analysis server and a database server for the analyses and accumulation of the experimental data respectively. Other peripheral devices of the system are magnetic disk units, NAS (Network Attached Storage) device, and a backup tape drive. This is a user's manual of the JT-60 experimental data analysis system. (author)

Reinstallation of the COMPASS-D tokamak in IPP ASCR

Czech Academy of Sciences Publication Activity Database

Pánek, Radomír; Hronová-Bilyková, Olena; Fuchs, Vladimír; Hron, Martin; Chráska, Pavel; Pavlo, Pavol; Stöckel, Jan; Urban, Jakub; Weinzettl, Vladimír; Zajac, Jaromír; Žáček, František

2006-01-01

Roč. 56, suppl. B (2006), s. 125-137 ISSN 0011-4626. [Symposium on Plasma Physics and Technology/22nd./. Prague, 26.6.2006-29.6.2006] R&D Projects: GA AV ČR(CZ) KJB100430602 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamaks * plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.568, year: 2006

Disruption generated secondary runaway electrons in present day tokamaks

International Nuclear Information System (INIS)

Pankratov, I.M.; Jaspers, R.

2000-01-01

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

Characteristics of large scale ionic source for JT-60

Energy Technology Data Exchange (ETDEWEB)

Fujiwara, Yukio; Honda, Atsushi; Inoue, Takashi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1997-02-01

The Neutral Beam Injection (NBI) apparatus is expected for important role sharing apparatus to realize the plasma electric current drive and the plasma control in not only temperature upgrading of the plasma but also Tokamak nuclear fusion reactor for the next generation such as JT-60, ITER and so forth. Japan Atomic Energy Research Institute has developed the ionic source with high energy and large electric current for about 10 years. Some arrangement tests of the large negative ion source for JT-60 No. 1 were executed from June to October, 1995. As a series of arrangement tests, 400 KeV and 13.5 A of deuterium negative ion beam was successfully accelerated for 0.12 sec. under 0.22 Pa of low gas pressure. And, it was elucidated that electron electric current could be controlled efficiently even in deuterium negative ion beam. Here is described on the testing results in details. (G.K.)

Advanced probes for edge plasma diagnostics on the CASTOR tokamak

Czech Academy of Sciences Publication Activity Database

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

2006-01-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

The Tokamak IST-TOK

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Progress in JT-60 innovative technologies

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-09-01

This review report provides the synthetic archives of innovative technologies in 20-year facility developments for the large tokamak experimental device JT-60, first founded as the major magnetic fusion device in the Second Basic Program for Fusion Research and Development of Japan. Manufacture of JT-60 was started in 1978, and the first plasma was achieved on April 8, 1985. In 1989-1991, the vacuum vessel and poloidal field coils were entirely reconfigured to improve the plasma performance. The major original mission of the JT-60 project, a breakeven condition for a D-T equivalent plasma, was finally attained in 1996. After this, JT-60, as a leading device for magnetic fusion research in the world, continues to challenge many experimental issues, which has been achieved by collaboration of innovative facility developments and experimental improvements. In addition, at this time to start the ITER construction phase in 2005, this review is expected to contribute the construction and operation activities for the next generation tokamak by providing the basic ideas in facility developments. We classified a tremendous number of development items into the selected sections for this review. Since the authors have been in charge of each development activity of their own, the contents are full of essential stories, points, and keywords in spite of its compact handbook size. We believe this review could provide highly sophisticated, informative ideas matured in JT-60 technological developments. (author)

Focus on nuclear fusion research

Czech Academy of Sciences Publication Activity Database

Křenek, Petr; Mlynář, Jan

2011-01-01

Roč. 61, - (2011), s. 62-63 ISSN 0375-8842 Institutional research plan: CEZ:AV0Z20430508 Keywords : ITER * COMPASS * fusion energy * tokamak * EURATOM Subject RIV: BL - Plasma and Gas Discharge Physics http://www.ipp.cas.cz/Tokamak/clanky/energetika_COMPASS.pdf

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

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

Overview on the progress of tokamak experimental research in China

International Nuclear Information System (INIS)

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

2001-01-01

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

DIII-D RESEARCH OPERATIONS ANNUAL REPORT October 1, 2001 through September 30, 2002

International Nuclear Information System (INIS)

EVANS, T.E.

2003-01-01

OAK-B135 The mission of the DIII-D research program is: ''To establish the scientific basis for the optimization of the tokamak approach to fusion energy production. The program is focused on developing the ultimate potential of the tokamak by building a better fundamental understanding of the physics of plasma confinement, stability, current drive and heating in high performance discharges while utilizing new scientific discoveries and improvements in their knowledge of these basic areas to create more efficient control systems, improved plasma diagnostics and to identify new types of enhanced operating regimes with improved stability properties. In recent years, this development path has culminated in the advanced tokamak (AT) approach. An approach that has shown substantial promise for improving both the fusion yield and the energy density of a burning plasma device. While the challenges of increasing AT plasma performance levels with greater stability for longer durations are significant, the DIII-D program has an established plan that brings together both the critical resources and the expertise needed to meet these challenges. The DIII-D research staff is comprised of about 300 individuals representing 60 institutions with many years of integrated research experience in tokamak physics, engineering and technology. The DIII-D tokamak is one of the most productive, flexible and best diagnosed magnetic fusion research devices in the world. It has significantly more flexibility than most tokamaks and continues to pioneer the development of sophisticated new plasma feedback control tools that enable the explorations of new frontiers in fusion science and engineering

DIII-D RESEARCH OPERATIONS ANNUAL REPORT TO THE U.S. DEPARTMENT OF ENERGY

Energy Technology Data Exchange (ETDEWEB)

EVANS,TE

2003-12-01

OAK-B135 The mission of the DIII-D research program is: ''To establish the scientific basis for the optimization of the tokamak approach to fusion energy production. The program is focused on developing the ultimate potential of the tokamak by building a better fundamental understanding of the physics of plasma confinement, stability, current drive and heating in high performance discharges while utilizing new scientific discoveries and improvements in their knowledge of these basic areas to create more efficient control systems, improved plasma diagnostics and to identify new types of enhanced operating regimes with improved stability properties. In recent years, this development path has culminated in the advanced tokamak (AT) approach. An approach that has shown substantial promise for improving both the fusion yield and the energy density of a burning plasma device. While the challenges of increasing AT plasma performance levels with greater stability for longer durations are significant, the DIII-D program has an established plan that brings together both the critical resources and the expertise needed to meet these challenges. The DIII-D research staff is comprised of about 300 individuals representing 60 institutions with many years of integrated research experience in tokamak physics, engineering and technology. The DIII-D tokamak is one of the most productive, flexible and best diagnosed magnetic fusion research devices in the world. It has significantly more flexibility than most tokamaks and continues to pioneer the development of sophisticated new plasma feedback control tools that enable the explorations of new frontiers in fusion science and engineering.

Operation and management manual of JT-60 experimental data analysis system

International Nuclear Information System (INIS)

Hirayama, Takashi; Morishima, Soichi

2014-03-01

In the Japan Atomic Energy Agency Naka Fusion Institute, a lot of experiments have been conducted by using the large tokamak device JT-60 aiming to realize fusion power plant. In order to optimize the JT-60 experiment and to investigate complex characteristics of plasma, JT-60 experimental data analysis system was developed and used for collecting, referring and analyzing the JT-60 experimental data. Main components of the system are a data analysis server and a database server for the analyses and accumulation of the experimental data respectively. Other peripheral devices of the system are magnetic disk units, NAS (Network Attached Storage) device, and a backup tape drive. This is an operation and management manual the JT-60 experimental data analysis system. (author)

Survey of Tokamak experiments

International Nuclear Information System (INIS)

Bickerton, R.J.

1977-01-01

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

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

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-03-01

This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

Measurement of Safety Factor Using Hall Probes on CASTOR Tokamak

Czech Academy of Sciences Publication Activity Database

Kovařík, Karel; Ďuran, Ivan; Bolshakova, I.; Holyaka, R.; Erashok, V.

2006-01-01

Roč. 56, suppl.B (2006), s. 104-110 ISSN 0011-4626. [Symposium on Plasma Physics and Technology/22nd./. Praha, 26.6.2006-29.6.2006] R&D Projects: GA AV ČR(CZ) KJB100430504 Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * safety factor * hall probe Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.568, year: 2006

Development and performance of high speed processing system of magnetohydrodynamic equilibria for discharge analyses on the J T-60 tokamak

International Nuclear Information System (INIS)

Hasegawa, Yukihiro; Nakamura, Yukiharu; Shirai, Hiroshi; Hamamatsu, Kiyotaka; Harada, Yoshio; Kikuchi, Mitsuru; Nakata, Yoshihiro

1999-01-01

In order to provide a set of magnetohydrodynamic (MHD) equilibrium database which is indispensable for both the studies on improvement of energy confinement and stabilization of MHD activities in tokamaks, a high speed data-processing system synchronizing with J T-60 discharge sequence was newly developed by utilizing the latest model of hugh speed workstation and by optimizing the parallel processing technique to perform fast calculation of MHD equilibria. This high speed system was found to have a sufficient ability to complete the whole equilibrium calculations during each inter-shot period. Cooperating with the mass data storage subsystem preserving the latest equilibrium database automatically, the animated discharge monitoring subsystem provides valuable information for the J T-60 operator to determine control parameters of the succeeding discharge. This report describes the system performance realized in the J T-60 experiment. (author)

Japan Atomic Energy Research Institute in the 21st century

International Nuclear Information System (INIS)

Sato, Y.

2001-01-01

Major nuclear research institutes in Japan are the Japan Atomic Energy Research Institute (JAERI), Nuclear Cycle Development Institute (JNC), National Research Institute of Radiological Science (NIRS), and the Institute of Physical and Chemical Research (RIKEN). In the 50s and 60s JAERI concentrated on the introduction of nuclear technology from overseas. Energy security issues led to the development of a strong nuclear power programme in the next two decades resulting in Japan having 50 light water cooled nuclear power plants in operation. Japan also worked on other reactor concepts. The current emphasis of JAERI is on advanced reactors and nuclear fusion. Its budget of 270 million US$ supports five research establishments. JAERI has strong collaboration with industry and university system on nuclear and other advanced research topics (neutron science, photon science). In many areas Japan has strong international links. JAERI has also been transferring know-how on radioisotope and radiation applications to the developing countries particularly through IAEA-RCA mechanisms. (author)

Recent results on steady state and confinement improvement research on JT-60U

International Nuclear Information System (INIS)

Ide, Shunsuke

2000-01-01

On the JT-60U tokamak, fusion plasma research for realization of a steady state tokamak reactor has been pursued. Towards that goal, confinement improved plasmas such as H-mode, high β p , reversed magnetic shear (RS) and latter two combined with H-mode edge pedestal have been developed and investigated intensively. A key issue to achieve non-inductive current drive relevant to a steady state fusion reactor is to increase the fraction of the bootstrap current and match the spatial profile to the optimum. In 1999, as the result of the optimization, the equivalent deuterium-tritium (D-T) fusion gain (Q DT eq ) of 0.5 was sustained for 0.8 s, which is roughly equal to the energy confinement time, in a RS plasma. In order to achieve a RS plasma in steady state two approach have been explored. One is to use external current driver such as lower hybrid current drive (LHCD), and by optimizing LHCD a quasi-steady RS discharge was obtained. The other approach is to utilize bootstrap current as much as possible, and with highly increased fraction of the bootstrap current, a confinement enhancement factor of 3.6 was maintained for 2.7 s in a RS plasma with H-mode edge. A heating and current drive system in the electron cyclotron range of frequency for localized heating and current drive has been installed on JT-60U, and in initial experiments a clear increase of the central electron temperature in a RS high density central region was confirmed only with injected power of 0.75 MW. (author)

KDAS: General-Purpose Data Acquisition System Developed for KAIST-Tokamak

International Nuclear Information System (INIS)

Seo, Seong-Heon; Choe, Wonho; Chang, Hong-Young; Jeong, Seung-Ho

2000-01-01

The Korea Advanced Institute of Science and Technology (KAIST)-Tokamak Data Acquisition System (KDAS) was originally developed for KAIST-Tokamak (R/a = 0.53 m/0.14 m). It operates on a distributed system based on personal computers and has a driver-based hierarchical structure. Since KDAS can be dynamically composed of any number of available computers, and the hardware-dependent codes can be thoroughly separated into external drivers, it exhibits excellent system performance flexibility and extensibility and can optimize various user needs. It collectively controls the VXI, CAMAC, GPIB, and RS232 instrument hybrids. With these useful and convenient features, it can be applied to any computerized experiment, especially to fusion-related research. The system design and features are discussed in detail

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

Summary discussion: An integrated advanced tokamak reactor

International Nuclear Information System (INIS)

Sauthoff, N.R.

1994-01-01

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

Conceptual radiation shielding design of superconducting tokamak fusion device by PHITS

International Nuclear Information System (INIS)

Sukegawa, Atsuhiko M.; Kawasaki, Hiromitsu; Okuno, Koichi

2010-01-01

A complete 3D neutron and photon transport analysis by Monte Carlo transport code system PHITS (Particle and Heavy Ion Transport code System) have been performed for superconducting tokamak fusion device such as JT-60 Super Advanced (JT-60SA). It is possible to make use of PHITS in the port streaming analysis around the devices for the tokamak fusion device, the duct streaming analysis in the building where the device is installed, and the sky shine analysis for the site boundary. The neutron transport analysis by PHITS makes it clear that the shielding performance of the superconducting tokamak fusion device with the cryostat is improved by the graphical results. From the standpoint of the port streaming and the duct streaming, it is necessary to calculate by 3D Monte Carlo code such as PHITS for the neutronics analysis of superconducting tokamak fusion device. (author)

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

Science.gov (United States)

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

2011-09-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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)

Scrape-off layer flows in the Tore Supra tokamak

Czech Academy of Sciences Publication Activity Database

Gunn, J. P.; Boucher, C.; Dionne, M.; Ďuran, Ivan; Fuchs, Vladimír; Loarer, T.; Pánek, Radomír; Saint Laurent, F.; Stöckel, Jan; Adámek, Jiří; Bucalossi, J.; Dejarnac, Renaud; Devynck, P.; Hertout, P.; Hron, Martin; Nanobashvili, I.; Rimini, F.G.; Sarkissian, A.

2006-01-01

Roč. 812, - (2006), s. 27-34 ISSN 0094-243X. [AIP Conference Proceedings. Opole-Turawa, 06.09.2006-09.09.2006] R&D Projects: GA ČR GP202/03/P062 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * scrape-off layer * plasma flow * radial transport * Mach probe Subject RIV: BL - Plasma and Gas Discharge Physics http://proceedings.aip.org/dbt/dbt.jsp?KEY=APCPCS&Volume=812&Issue=1

Tokamak edge electron diffusion and distribution function in the lower hybrid antenna electric field

Czech Academy of Sciences Publication Activity Database

Fuchs, Vladimír; Gunn, J. P.; Goniche, M.; Petržílka, Václav

2003-01-01

Roč. 43, č. 5 (2003), s. 341-351 ISSN 0029-5515 R&D Projects: GA ČR GA202/00/1217 Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, grill electric field Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.390, year: 2003

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.)

Conceptual design of JT-60SA cryostat

International Nuclear Information System (INIS)

Shibama, Y.K.; Sakurai, S.; Masaki, K.; Sukekawa, A.M.; Kaminaga, A.; Sakasai, A.; Matsukawa, M.

2008-01-01

This paper describes the conceptual design of cryostat for the JT-60SA, which is a research device for the commercial production of electricity from the controlled fusion reaction in the future. JT-60SA is designed to be a fully superconducting device and cryostat is one of the main components to allow the normal operation. Cryostat covers up the tokamak device, which is 15 m of total height and 7 m of radius, and supports the total weight of 25 MN. Cryostat components consist of vessel body, gravity support and auxiliary systems, such as 80 K thermal shield and vacuum exhaust. The functions required of cryostat are these three, thermal insulation for superconducting magnets, gravity support for the tokamak device, and bio-shielding. The design conditions for each cryostat component are outlined and the features of auxiliary systems such as capacity of vacuum exhaust related to 80 K thermal shield design are summarized

Establishment of an Institute for Fusion Studies. Technical progress report, November 1, 1994--October 31, 1995

International Nuclear Information System (INIS)

1995-07-01

The Institute for Fusion Studies is a national center for theoretical fusion plasma physics research. Its purposes are to (1) conduct research on theoretical questions concerning the achievement of controlled fusion energy by means of magnetic confinement--including both fundamental problems of long-range significance, as well as shorter-term issues; (2) serve as a national and international center for information exchange by hosting exchange visits, conferences, and workshops; and (3) train students and postdoctoral research personnel for the fusion energy program and plasma physics research areas. During FY 1995, a number of significant scientific advances were achieved at the IFS, both in long-range fundamental problems as well as in near-term strategic issues, consistent with the Institute's mandate. Examples of these achievements include, for example, tokamak edge physics, analytical and computational studies of ion-temperature-gradient-driven turbulent transport, alpha-particle-excited toroidal Alfven eigenmode nonlinear behavior, sophisticated simulations for the Numerical Tokamak Project, and a variety of non-tokamak and non-fusion basic plasma physics applications. Many of these projects were done in collaboration with scientists from other institutions. Research discoveries are briefly described in this report

Assembly work and transport of JT-60SA cryostat base

International Nuclear Information System (INIS)

Okano, Fuminori; Masaki, Kei; Yagyu, Jun-ichi; Shibama, Yusuke; Sakasai, Akira; Miyo, Yasuhiko; Kaminaga, Atsushi; Nishiyama, Tomokazu; Suzuki, Sadaaki; Nakamura, Shigetoshi; Shibanuma, Kiyoshi

2013-11-01

Japan Atomic Energy Agency started to construct a fully superconducting tokamak experiment device, JT-60SA, to support the ITER since January, 2013 at the Fusion Research and Development Directorate in Naka, Japan. The JT-60SA will be constructed with enhancing the previous JT-60 infrastructures, in the JT-60 torus hall, where the ex-JT-60 machine was disassembled. The JT-60SA Cryostat Base, for base of the entire tokamak structure, were assembly as the first step of this construction. The Cryostat Base (CB, 250tons) is consists of 7 main components made of stainless steel, in 12 m diameter and 3 m height. The CB was built in the Spain and transported to the Naka site, via Hitachi port. After pre-assembly work including preliminary measurements and sole plate adjustments of its height/flatness, the JT-60SA CB was carefully set on the sole plate. JT-60SA CB was assembled with high accuracy by using a laser tracker. The CB was adjusted in the height and flatness against the assembly reference position and determined by the absolute coordinates. This report introduces the concrete result of assembly work and transport of JT-60SA CB. (author)

Electronic system of TBR tokamak device

International Nuclear Information System (INIS)

Silva, R.P. da.

1980-01-01

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

Irradiation Facilities of the Takasaki Advanced Radiation Research Institute

Directory of Open Access Journals (Sweden)

Satoshi Kurashima

2017-03-01

Full Text Available The ion beam facility at the Takasaki Advanced Radiation Research Institute, the National Institutes for Quantum and Radiological Science and Technology, consists of a cyclotron and three electrostatic accelerators, and they are dedicated to studies of materials science and bio-technology. The paper reviews this unique accelerator complex in detail from the viewpoint of its configuration, accelerator specification, typical accelerator, or irradiation technologies and ion beam applications. The institute has also irradiation facilities for electron beams and 60Co gamma-rays and has been leading research and development of radiation chemistry for industrial applications in Japan with the facilities since its establishment. The configuration and utilization of those facilities are outlined as well.

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

Tokamak

International Nuclear Information System (INIS)

Wesson, John.

1996-01-01

This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

Evaluation of the plasma parameters in COMPASS tokamak divertor area

Czech Academy of Sciences Publication Activity Database

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

2012-01-01

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

Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

Energy Technology Data Exchange (ETDEWEB)

Matsukawa, M. E-mail: matsukaw@naka.jaeri.go.jp; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T

2003-09-01

The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control.

Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

International Nuclear Information System (INIS)

Matsukawa, M.; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T.

2003-01-01

The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control

Thermonuclear Power Engineering: 60 Years of Research. What Comes Next?

Science.gov (United States)

Strelkov, V. S.

2017-12-01

This paper summarizes results of more than half a century of research of high-temperature plasmas heated to a temperature of more than 100 million degrees (104 eV) and magnetically insulated from the walls. The energy of light-element fusion can be used for electric power generation or as a source of fissionable fuel production (development of a fusion neutron source—FNS). The main results of studies of tokamak plasmas which were obtained in the Soviet Union with the greatest degree of thermal plasma isolation among all other types of devices are presented. As a result, research programs of other countries were redirected to tokamaks. Later, on the basis of the analysis of numerous experiments, the international fusion community gradually came to an opinion that it is possible to build a tokamak (ITER) with Q > 1 (where Q is the ratio of the fusion power to the external power injected into the plasma). The ITER program objective is to achieve Q = 1-10 for a discharge time of up to 1000 s. The implementation of this goal does not solve the problem of a steadystate operation. The solution to this problem is a reliable first wall and current generation. This is a task of the next fusion power plant construction stage, called DEMO. Comparison of DEMO and FNS parameters shows that, at this development stage, the operating parameters and conditions of these devices are identical.

First experiments with SST-1 tokamak

International Nuclear Information System (INIS)

Saxena, Y.C.

2005-01-01

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

Low cost alternative of high speed visible light camera for tokamak experiments

Czech Academy of Sciences Publication Activity Database

Odstrčil, T.; Odstrčil, Michal; Grover, O.; Svoboda, V.; Ďuran, Ivan; Mlynář, Jan

2012-01-01

Roč. 83, č. 10 (2012), 10E505-10E505 ISSN 0034-6748. [Topical Conference High-Temperature Plasma Diagnostics/19./. Monterey, 06.05.2012-10.05.2012] Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma * tokamak * diagnostic * high speed camera * GOLEM Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.602, year: 2012 http://dx.doi.org/10.1063/1.4731003

Preliminary design study of a steady state tokamak device

International Nuclear Information System (INIS)

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

1992-09-01

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

Diagnostics for the Rijnhuizen Tokamak Project

NARCIS (Netherlands)

Donne, A. J. H.

1994-01-01

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

Recent fusion research in the National Institute for Fusion Science

International Nuclear Information System (INIS)

Komori, Akio; Sakakibara, Satoru; Sagara, Akio; Horiuchi, Ritoku; Yamada, Hiroshi; Takeiri, Yasuhiko

2011-01-01

The National Institute for Fusion Science (NIFS), which was established in 1989, promotes academic approaches toward the exploration of fusion science for steady-state helical reactor and realizes the establishment of a comprehensive understanding of toroidal plasmas as an inter-university research organization and a key center of worldwide fusion research. The Large Helical Device (LHD) Project, the Numerical Simulation Science Project, and the Fusion Engineering Project are organized for early realization of net current free fusion reactor, and their recent activities are described in this paper. The LHD has been producing high-performance plasmas comparable to those of large tokamaks, and several new findings with regard to plasma physics have been obtained. The numerical simulation science project contributes understanding and systemization of the physical mechanisms of plasma confinement in fusion plasmas and explores complexity science of a plasma for realization of the numerical test reactor. In the fusion engineering project, the design of the helical fusion reactor has progressed based on the development of superconducting coils, the blanket, fusion materials and tritium handling. (author)

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

Integrated plasma control for high performance tokamaks

International Nuclear Information System (INIS)

Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.

2005-01-01

Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)

Tokamak startup with electron cyclotron heating

International Nuclear Information System (INIS)

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

1980-04-01

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

Tokamak startup with electron cyclotron heating

Energy Technology Data Exchange (ETDEWEB)

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

1980-04-01

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

20 years of research on the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Greenwald, M.; Baek, S.; Barnard, H.; Beck, W.; Bonoli, P.; Brunner, D.; Burke, W.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Gao, C.; Golfinopoulos, T.; Granetz, R.; Hartwig, Z.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.

2014-01-01

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental

The scientific program of the Tokamak de Varennes

International Nuclear Information System (INIS)

Daughney, C.C.

1989-01-01

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

Gamma ray imager on the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.; Van Zeeland, M. A.; Watkins, M. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Cooper, C. M. [Oak Ridge Associated Universities, Oak Ridge, Tennessee 37830 (United States); Hollmann, E. M. [University of California-San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Riso, V. [State University of New York-Buffalo, 12 Capen Hall, Buffalo, New York 14260-1660 (United States)

2016-04-15

A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electrons in the energy range of 1–60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.

Spatially resolved characterization of electrostatic fluctuations in the scrape-off layer of the CASTOR tokamak

Czech Academy of Sciences Publication Activity Database

Devynck, P.; Bonhomme, G.; Martines, E.; Stöckel, Jan; Van Oost, G.; Voitsekhovitch, I.; Adámek, Jiří; Azeroual, A.; Doveil, F.; Ďuran, Ivan; Gravier, E.; Gunn, J.; Hron, Martin

2005-01-01

Roč. 47, č. 2 (2005), s. 269-280 ISSN 0741-3335 R&D Projects: GA ČR GA202/03/0786 Grant - others:GA - INTAS 2001 2056 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.902, year: 2005

Magnetic confinement experiment. I: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1995-08-01

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

Bibliography of fusion product physics in tokamaks

International Nuclear Information System (INIS)

Hively, L.M.; Sigmar, D.J.

1989-09-01

Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

Accomplishment of JT-60U disassembly work dealing with radioactive components

International Nuclear Information System (INIS)

Ikeda, Yoshitaka

2015-01-01

The upgrade of the JT-60U to the superconducting tokamak 'JT-60SA' has been carried out to contribute the early realization of fusion energy by addressing key physics issues relevant for ITER and DEMO. Disassembly of the JT-60U tokamak was required so as to newly install the JT-60SA torus at the same position in the torus hall. The JT-60U tokamak was featured by the complicated and welded structure against the strong electromagnetic force, and by the radioactivation due to deuterium-deuterium (D-D) reactions of 1.5x10"2"0 (n) in total. Since this work was the first experience of disassembling a large radioactivated fusion device in Japan, careful preparations of disassembly activities, including treatment of the radioactivated materials and safety work, have been made. About 13,000 components with a total weight of more than 5,400 tonnes were removed from the torus hall and stored safely in storage facilities. All disassembly components were stored with recording the data such as dose rate, weight and kind of material, so as to apply the clearance level regulation in future. It was confirmed that the main radioactive material of the disassembly components was the stainless steel and that its dose rate was almost background level (∼0.1 μSv/h) at ∼10 m far from the vacuum vessel. It seems that the disassembly components with background dose level are in the clearance level. The assembly of JT-60SA tokamak has started in January 2013 after this disassembly of the JT-60U tokamak. (author)

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1993-04-01

This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

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

High Beta Tokamak research

International Nuclear Information System (INIS)

Navratil, G.A.; Mauel, M.E.; Ivers, T.H.; Sankar, M.K.V.; Eisner, E.; Gates, D.; Garofalo, A.; Kombargi, R.; Maurer, D.; Nadle, D.; Xiao, Q.

1993-01-01

During the past 6 months, experiments have been conducted with the HBT-EP tokamak in order to (1) test and evaluate diagnostic systems, (2) establish basic machine operation, (3) document MHD behavior as a function of global discharge parameters, (4) investigate conditions leading to passive stabilization of MHD instabilities, and (5) quantify the external saddle coil current required for DC mode locking. In addition, the development and installation of new hardware systems has occurred. A prototype saddle coil was installed and tested. A five-position (n,m) = (1,2) external helical saddle coil was attached for mode-locking experiments. And, fabrication of the 32-channel UV tomography and the multipass Thomson scattering diagnostics have begun in preparation for installation later this year

Virtual reality applications in remote handling development for tokamaks in India

International Nuclear Information System (INIS)

Dutta, Pramit; Rastogi, Naveen; Gotewal, Krishan Kumar

2017-01-01

Highlights: • Evaluation of Virtual Reality (VR) in design and operation phases of Remote Handling (RH) equipment for tokamak. • VR based centralized facility, to cater RH development and operation, is setup at Institute for Plasma Research, India. • The VR facility system architecture and components are discussed. • Introduction to various VR applications developed for design and development of tokamak RH equipment. - Abstract: A tokamak is a plasma confinement device that can be used to achieve magnetically confined nuclear fusion within a reactor. Owing to the harsh environment, Remote Handling (RH) systems are used for inspection and maintenance of the tokamak in-vessel components. As the number of in-vessel components requiring RH maintenance is large, physical prototyping of all strategies becomes a major challenge. The operation of RH systems poses further challenge as all equipment have to be controlled remotely within very strict accuracy limits with minimum reliance on the available camera feedback. In both design and operation phases of RH equipment, application of Virtual Reality (VR) becomes imperative. The scope of this paper is to introduce some applications of VR in the design and operation cycle of RH, which are not available commercially. The paper discusses the requirement of VR as a tool for RH equipment design and operation. The details of a comprehensive VR facility that has been established to support the RH development for Indian tokamaks are also presented. Further, various cases studies are provided to highlight the utilization of this VR facility within phases of RH development and operation.

Virtual reality applications in remote handling development for tokamaks in India

Energy Technology Data Exchange (ETDEWEB)

Dutta, Pramit, E-mail: pramitd@ipr.res.in; Rastogi, Naveen; Gotewal, Krishan Kumar

2017-05-15

Highlights: • Evaluation of Virtual Reality (VR) in design and operation phases of Remote Handling (RH) equipment for tokamak. • VR based centralized facility, to cater RH development and operation, is setup at Institute for Plasma Research, India. • The VR facility system architecture and components are discussed. • Introduction to various VR applications developed for design and development of tokamak RH equipment. - Abstract: A tokamak is a plasma confinement device that can be used to achieve magnetically confined nuclear fusion within a reactor. Owing to the harsh environment, Remote Handling (RH) systems are used for inspection and maintenance of the tokamak in-vessel components. As the number of in-vessel components requiring RH maintenance is large, physical prototyping of all strategies becomes a major challenge. The operation of RH systems poses further challenge as all equipment have to be controlled remotely within very strict accuracy limits with minimum reliance on the available camera feedback. In both design and operation phases of RH equipment, application of Virtual Reality (VR) becomes imperative. The scope of this paper is to introduce some applications of VR in the design and operation cycle of RH, which are not available commercially. The paper discusses the requirement of VR as a tool for RH equipment design and operation. The details of a comprehensive VR facility that has been established to support the RH development for Indian tokamaks are also presented. Further, various cases studies are provided to highlight the utilization of this VR facility within phases of RH development and operation.

Southwestern Institute of Physics annual report 2001

International Nuclear Information System (INIS)

2002-01-01

In the year 2001, significant progresses in the engineering construction of the HL-2A tokamak were made at the Southwestern Institute of Physics (SWIP). At the same time, the research projects from Nuclear Energy Development Foundation, the National Defense Basic Research Foundation and the National Science Foundation of China were completely fulfilled. In addition 283 papers and reports were contributed, among them, 67 are included in the Annual Report

On Use of Semiconductor Detector Arrays on COMPASS Tokamak

Czech Academy of Sciences Publication Activity Database

Weinzettl, Vladimír; Imríšek, Martin; Havlíček, Josef; Mlynář, Jan; Naydenkova, Diana; Háček, Pavel; Hron, Martin; Janky, Filip; Sarychev, D.; Berta, M.; Bencze, A.; Szabolics, T.

-, č. 71 (2012), s. 844-850 ISSN 2010-376X. [ICPP 2012 : International Conference on Plasma Physics. Venice, 14.11.2012-16.11.2012] R&D Projects: GA ČR GA202/09/1467; GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : bolometry * plasma diagnostics * soft X-rays * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics https://www.waset.org/journals/waset/v71/v71-143.pdf

Continuous tokamaks

International Nuclear Information System (INIS)

Peng, Y.K.M.

1978-04-01

A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

Energy confinement and transport of H-mode plasmas in tokamak

International Nuclear Information System (INIS)

Urano, Hajime

2005-02-01

A characteristic feature of the high-confinement (H-mode) regime is the formation of a transport barrier near the plasma edge, where steepening of the density and temperature gradients is observed. The H-mode is expected to be a standard operation mode in a next-step fusion experimental reactor, called ITER-the International Thermonuclear Experimental Reactor. However, energy confinement in the H-mode has been observed to degrade with increasing density. This is a critical constraint for the operation domain in the ITER. Investigation of the main cause of confinement degradation is an urgent issue in the ITER Physics Research and Development Activity. A key element for solving this problem is investigation of the energy confinement and transport properties of H-mode plasmas. However, the influence of the plasma boundary characterized by the transport barrier in H-modes on the energy transport of the plasma core has not been examined sufficiently in tokamak research. The aim of this study is therefore to investigate the energy confinement properties of H-modes in a variety of density, plasma shape, seed impurity concentration, and conductive heat flux in the plasma core using the experimental results obtained in the JT-60U tokamak of Japan Atomic Energy Research Institute. Comparison of the H-mode confinement properties with those of other tokamaks using an international multi-machine database for extrapolation to the next step device was also one of the main subjects in this study. Density dependence of the energy confinement properties has been examined systematically by separating the thermal stored energy into the H-mode pedestal component determined by MHD stability called the Edge Localized Modes (ELMs) and the core component governed by gyro-Bohm-like transport. It has been found that the pedestal pressure imposed by the destabilization of ELM activities led to a reduction in the pedestal temperature with increasing density. The core temperature for each

Numerical study for determining PF coil system parameters in MHD equilibrium of KT-2 tokamak plasma

International Nuclear Information System (INIS)

Ryu, J.; Hong, S.H.; Lee, K.W.; Hong, B.G.; In, S.R.; Kim, S.K.

1995-01-01

The KT-2 is a large-aspect-ratio medium-sized divertor tokamak in the conceptual design phase and planned to be operational in 1998 at the Korea Atomic Energy Research Institute (KAERI). Plasma equilibrium in tokamak can be acquired by controlling the current of poloidal field (PF) coils in appropriate geometries and positions. In this study, the authors have performed numerical calculations to achieve the various equilibrium conditions fitting given plasma shapes and satisfying PF current limitations. Usually an ideal magnetohydrodynamic (MHD) equation is used to obtain the equilibrium solution of tokamak plasma, and it is practical to take advantage of a numerical method in solving the MHD equation because it has nonlinear source terms. Two equilibrium codes have been applied to find a double-null configuration of free-boundary tokamak plasma in KT-2: one is of the authors' own developing and the other is a free-boundary tokamak equilibrium code (FBT) that has been used mainly for the verification of developed code's results. PF coil system parameters including their positions and currents are determined for the optimization of input power required when the specifications of KT-2 tokamak are met. Then, several sets of equilibrium conditions during the tokamak operation are found to observe the changes of poloidal field currents with the passing of operation time step, and the basic stability problems related with the magnetic field structure is also considered

Computer and engineering calculations of Brazilian Tokamak-II

International Nuclear Information System (INIS)

Wang, S.; Chen, Y.; Sa, W.P. de; Nascimento, I.C.; Tuszel, A.G.; Galvao, R.M.O.; Machida, M.

1990-01-01

Analytical and computer calculations carried out by researches of Physics Institute - University of Sao Paulo (IFUSP), for defining the engineering project and constructing the TBR-II tokamak are presented. The hydrodynamics behavioue and determined parameters for magnetic confinement of the plasma were analysed. The computer code was developed using magnetohydrodynamics (MHD) equations which involve plasma interactions, magnetic field and electrical current circulating in more than 20 coils distributed around toroidal vase of the plasma. The electromagnetic, thermal and mechanical couplings are also presented. The TBR-II will be feed by two turbo-generators with 15 MW each one. (M.C.K.) [pt

Magnetic measurements using array of integrated Hall sensors on the CASTOR tokamak

Czech Academy of Sciences Publication Activity Database

Ďuran, Ivan; Hronová-Bilyková, Olena; Stöckel, Jan; Sentkerestiová, J.; Havlíček, Josef

2008-01-01

Roč. 79, č. 10 (2008), 10F123-10F123 ISSN 0034-6748. [Topical Conference on High-Temperature Plasma Diagnostics/17th./. Albuquerque, 11.05.2008-15.05.2008] R&D Projects: GA MPO 2A-1TP1/101 Institutional research plan: CEZ:AV0Z20430508 Keywords : Galvanomagnetic Sensor * Fusion Reactor * Magnetic Diagnostics * CASTOR tokamak Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.738, year: 2008

MAST: a Mega Amp Spherical Tokamak

International Nuclear Information System (INIS)

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

1995-01-01

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

Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak; Projeto e analise dos circuitos de producao de campo magnetico toroidal e de formacao do plasma do Tokamak ETE (Experimento Tokamak Esferico)

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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.

Forgotten research institute makes money from ideas

International Nuclear Information System (INIS)

Sobinkovic, B.

2008-01-01

Robots that stack magnets weighing several tons in the world's biggest nuclear laboratory with a millimetre precision. Small machines that can destroy bombs, detect bombs in trains, planes or cars. A leading position in an expert group that, with NATO funds, tests how robotic systems can be used in the fight against terrorism. This summary indicates that ideas are an integral part of the work done at the ZTS Vyskumno-vyvojovy ustav (ZTS VVU) research institute in Kosice. This is nothing special for a research institute. But this is a joint stock company. And so it needed one additional vision: producing goods that sell from the research. ZTS VVU has delivered robotic system for accurate positioning of cryo-magnets for the CERN. Cryo-magnet is 16 m long and weights 34 tonnes. For the CERN five robotic systems were delivered. The value of the contract with the CERN was about 60 millions slovak crowns (≅ 2 million EUR). Transport containers, manipulators for decontamination and manipulators with radioactive wastes were manufactured for the Bohunicke spracovatelske centrum (Bohunice Radioactive Waste Processing Center). (authors)

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Plasma diagnostics on large tokamaks

International Nuclear Information System (INIS)

Orlinskij, D.V.; Magyar, G.

1988-01-01

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

Installation and pre-commissioning of the cryogenic system of JT-60SA tokamak

Science.gov (United States)

Hoa, C.; Michel, F.; Roussel, P.; Fejoz, P.; Girard, S.; Goncalves, R.; Lamaison, V.; Natsume, K.; Kizu, K.; Koide, Y.; Yoshida, K.; Cardella, A.; Portone, A.; Verrecchia, M.; Wanner, M.; Beauvisage, J.; Bertholat, F.; Gaillard, G.; Heloin, V.; Langevin, B.; Legrand, J.; Maire, S.; Perrier, J. M.; Pudys, V.

2017-02-01

The cryogenic system for the superconducting tokamak JT-60SA is currently being commissioned in Naka, Japan and shall be ready for operation in summer 2016. This contribution is part of the Broader Approach agreement between Japan and Europe. With an equivalent refrigeration capacity of about 9.5 kW at 4.5 K the cryogenic system will supply cryo-pump panels at 3.7 K, superconducting magnets and their structures at 4.4 K, high temperature superconducting current leads at 50 K and thermal shields between 80 K and 100 K. The system has been specifically designed to handle large pulse loads at 4.4 K during plasma operation. The mechanical and electrical assembly of the cryogenic system has been achieved within six months by October 2015. The main contractor Air Liquide Advanced Technology (AL-aT) have supplied eight parallel working screw compressors with a common oil removal and dryer system, a Refrigeration Cold Box and an Auxiliary Cold box with cold rotating machines. F4E has provided six GHe storage vessels and QST has provided the complete infrastructure and the facilities for the utilities. The paper gives an overview of the main design features, the infrastructure and the status of installation and pre-commissioning.

The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1989-10-01

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

Fast IR diodes thermometer for tokamak

International Nuclear Information System (INIS)

Chen Xiangbo

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Ion cyclotron system design for KSTAR tokamak

International Nuclear Information System (INIS)

Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H.

1998-05-01

The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs

Ion cyclotron system design for KSTAR tokamak

Energy Technology Data Exchange (ETDEWEB)

Hong, B. G.; Hwang, C. K.; Jeong, S. H.; Yoony, J. S.; Bae, Y. D.; Kwak, J. G.; Ju, M. H

1998-05-01

The KSTAR (Korean Superconducting Tokamak Advanced Research) tokamak (R=1.8 m, a=0.5 m, k=2, b=3.5T, I=2MA, t=300 s) is being constructed to do long-pulse, high-b, advanced-operating-mode fusion physics experiments. The ion cyclotron (IC) system (in conjunction with an 8-MW neutral beam and a 1.5-MW lower hybrid system) will provide heating and current drive capability for the machine. The IC system will deliver 6 MW of RF power to the plasma in the 25 to 60 MHz frequency range, using a single four-strap antenna mounted in a midplane port. It will be used for ion heating, fast-wave current drive (FWCD), and mode-conversion current drive (MCCD). The phasing between current straps in the antenna will be adjustable quickly during operation to provide the capability of changing the current-drive efficiency. This report describes the design of the IC system hardware: the electrical characteristics of the antenna and the matching system, the requirements on the power sources, and electrical analyses of the launcher. (author). 7 refs., 2 tabs., 40 figs.

Radiation processing project at the Institute of Nuclear Energy Research

International Nuclear Information System (INIS)

Tsai, C.M.; Fu, Y.K.; Yang, Y.H.; Chen, Y.T.; Wei, Y.H.; Lee, K.P.; Wang, Y.K.

1981-01-01

The utilization of scientific approach to preserve and sterilize the agricultural products has long been studied since 1954 and was adopted by several countries gradually since 1958. Starting from July 1977 this Institute began to study the preservation of potatoes and onions with reference to sprout inhibition which is discussed and its economical aspect is evaluated. The design concept of a megacurie 60 Co irradiator at this Institute is illustrated. The progress of construction work for the irradiator and the safety device in particular are reported. Current research project on the preservation of agricultural products in this Institute is presented. (author)

The ARIES-II and ARIES-IV second-stability tokamak reactors

International Nuclear Information System (INIS)

Najmabadi, F.; Conn, R.W.; Hasan, M.Z.; Mau, T.-K.; Sharafat, S.; Baxi, C.B.; Leuer, J.A.; McQuillan, B.W.; Puhn, F.A.; Schultz, K.R.; Wong, C.P.C.; Brooks, J.; Ehst, D.A.; Hassanein, A.; Hua, T.; Hull, A.; Mattis, R.; Picologlou, B.; Sze, D.-K.; Dolan, T.J.; Herring, J.S.; Bathke, C.G.; Krakowski, R.A.; Werley, K.A.; Bromberg, L.; Schultz, J.; Davis, F.; Holmes, J.A.; Lousteau, D.C.; Strickler, D.J.; Jardin, S.C.; Kessel, C.; Snead, L.; Steiner, D.; Valenti, M.; El-Guebaly, L.A.; Emmert, G.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.; Sviatoslavsky, I.N.; Cheng, E.T.

1992-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. The ARIES-III study focuses on the potential of tokamaks to operate with D- 3 He fuel system as an alternative to deuterium and tritium. The ARIES-II and ARIES-IV designs have the same fusion plasma but different fusion-power-core designs. The ARIES-II reactor uses liquid lithium as the coolant and tritium breeder and vanadium alloy as the structural material in order to study the potential of low-activation metallic blankets. The ARIES-IV reactor uses helium as the coolant, a solid tritium-breeding material, and silicon carbide composite as the structural material in order to achieve the safety and environmental characteristic of fusion. In this paper the authors describe the trade-off leading to the optimum regime of operation for the ARIES-II and ARIES-IV second-stability reactors and review the engineering design of the fusion power cores

Lenin nuclear reactor research institute in the tenth five-year plan

International Nuclear Information System (INIS)

Tsykanov, V.A.; Kulov, E.V.

1980-01-01

Main tasks and research results of Lenin Nuclear Reactor Reseach Institute in the 10-th Five-Year Plan are considered. Main research achievements are noted in nuclear power, radiation material testing, accumulation of transuranium elements and investigation of their physicochemical properties at VK-50, BOR-60, SM-2, RBT-6 and MIR reactor plants and in material testing laboratories

Massachusetts Institute of Technology, Plasma Fusion Center, technical research programs

International Nuclear Information System (INIS)

1982-02-01

Research programs have produced significant results on four fronts: (1) the basic physics of high-temperature fusion plasmas (plasma theory, RF heating, development of advanced diagnostics and small-scale experiments on the Versator tokamak and Constance mirror devices); (2) major confinement results on the Alcator A and C tokamaks, including pioneering investigations of the equilibrium, stability, transport and radiation properties of fusion plasmas at high densities, temperatures and magnetic fields; (3) development of a new and innovative design for axisymmetric tandem mirrors with inboard thermal barriers, with initial operation of the TARA tandem mirror experimental facility scheduled for 1983; and (4) a broadly based program of fusion technology and engineering development that addresses problems in several critical subsystem areas

Acceleration mechanism of vertical displacement event and its amelioration in tokamak disruptions

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Yoshino, Ryuji; Pomphrey, N.; Jardin, S.C.

1996-01-01

Vertical displacement events (VDEs), which are frequently observed in disruptive discharges of elongated tokamaks, are investigated using the Tokamak Simulation Code. We show that disruption events such as a sudden plasma pressure drop (β p collapse) and the subsequent plasma current quench (I p quench) can accelerate VDEs due to the adverse destabilizing effect of the resistive shell, which has previously been thought to stabilize VDEs. In a tokamak with a surrounding shell which is asymmetric with respect to the geometric midplane, the I p quench also causes an additional VDE acceleration due to the vertical imbalance of the attractive force. While the shell-geometry characterizes the VDE dynamics, the growth rate of VDEs depends strongly on the magnitude of the β p collapse, the speed of the I p quench and the n-index of the plasma equilibrium just before the disruption. An amelioration of I p quench-induced VDEs was experimentally established in the JT-60U tokamak by optimizing the vertical location of the plasma just prior to the disruption. The JT-60U vacuum vessel is shown to be suitable for preventing the β p collapse-induced VDE. (author)

Inhalation Toxicology Research Institute. Annual report, October 1, 1992--September 30, 1993

Energy Technology Data Exchange (ETDEWEB)

Nikula, K.J.; Belinsky, S.A.; Bradley, P.L. [eds.

1993-11-01

This annual report for the Inhalation Toxicology Research Institute for 1992-1993 consists of 60 individual reports prepared separately by investigators describing progress in their own projects. Most papers are 2-5 pages long.

Integrated Tokamak modeling: When physics informs engineering and research planning

Science.gov (United States)

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. It discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.

Application of diamond window for infrared laser diagnostics in a tokamak device

International Nuclear Information System (INIS)

Kawano, Yasunori; Chiba, Shinichi; Inoue, Akira

2004-01-01

Chemical vapor deposited diamond disks have been successfully applied as the vacuum windows for infrared CO 2 laser interferometry and polarimetry used in electron density measurement in the JT-60U tokamak. In comparison with the conventional zinc-selenide windows, the Faraday rotation component of diamond windows was negligible. This results in an improvement of the Faraday rotation measurement of tokamak plasma by polarimetry

Analysis of line integrated electron density using plasma position data on Korea Superconducting Tokamak Advanced Research

International Nuclear Information System (INIS)

Nam, Y. U.; Chung, J.

2010-01-01

A 280 GHz single-channel horizontal millimeter-wave interferometer system has been installed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR) device. This system has a triangular beam path that does not pass through the plasma axis due to geometrical constraints in the superconducting tokamak. The term line density on KSTAR has a different meaning from the line density of other tokamaks. To estimate the peak density and the mean density from the measured line density, information on the position of the plasma is needed. The information has been calculated from tangentially viewed visible images using the toroidal symmetry of the plasma. Interface definition language routines have been developed for this purpose. The calculated plasma position data correspond well to calculation results from magnetic analysis. With the position data and an estimated plasma profile, the peak density and the mean density have been obtained from the line density. From these results, changes of plasma density themselves can be separated from effects of the plasma movements, so they can give valuable information on the plasma status.

Tokamak Systems Code

International Nuclear Information System (INIS)

Reid, R.L.; Barrett, R.J.; Brown, T.G.

1985-03-01

The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

Electron cyclotron-electron Bernstein wave emission diagnostics for the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Zajac, Jaromír; Preinhaelter, Josef; Urban, Jakub; Žáček, František; Šesták, David; Nanobashvili, S.

2010-01-01

Roč. 81, č. 10 (2010), 10D911-10D911 ISSN 0034-6748. [TOPICAL CONFERENCE ON HIGH-TEMPERATURE PLASMA DIAGNOSTICS/18th./. Wildwood, New Jersey, 16.05.2010-20.05.2010] R&D Projects: GA ČR GA202/08/0419 Institutional research plan: CEZ:AV0Z20430508 Keywords : antenna radiation patterns * antennas in plasma * plasma diagnostics * Tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.598, year: 2010 http://link.aip.org/link/?RSI/81/10D911

BRIC-60: Biological Research in Canisters (BRIC)-60

Science.gov (United States)

Richards, Stephanie E. (Compiler); Levine, Howard G.; Romero, Vergel

2016-01-01

The Biological Research in Canisters (BRIC) is an anodized-aluminum cylinder used to provide passive stowage for investigations evaluating the effects of space flight on small organisms. Specimens flown in the BRIC 60 mm petri dish (BRIC-60) hardware include Lycoperscion esculentum (tomato), Arabidopsis thaliana (thale cress), Glycine max (soybean) seedlings, Physarum polycephalum (slime mold) cells, Pothetria dispar (gypsy moth) eggs and Ceratodon purpureus (moss).

Nuclear Fusion Fuel Cycle Research Perspectives

International Nuclear Information System (INIS)

Chung, Hongsuk; Koo, Daeseo; Park, Jongcheol; Kim, Yeanjin; Yun, Sei-Hun

2015-01-01

As a part of the International Thermonuclear Experimental Reactor (ITER) Project, we at the Korea Atomic Energy Research Institute (KAERI) and our National Fusion Research Institute (NFRI) colleagues are investigating nuclear fusion fuel cycle hardware including a nuclear fusion fuel Storage and Delivery System (SDS). To have a better knowledge of the nuclear fusion fuel cycle, we present our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). To have better knowledge of the nuclear fusion fuel cycle, we presented our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). Our efforts to enhance the tritium confinement will be continued for the development of cleaner nuclear fusion power plants

Neutral beam in ALVAND IIC tokamak

International Nuclear Information System (INIS)

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

1992-01-01

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

Annual report of the Fusion Research Center for the period of April 1, 1982 to March 31, 1983

International Nuclear Information System (INIS)

1983-11-01

Research and development activities of the Fusion Research Center (Department of Thermonuclear Fusion Research and Department of Large Tokamak Development) from April 1982 to March 1983 are described. The JFT-2 tokamak was shutdown after 10 years operation. Operation test of a new device JFT-2M was near completion. In the joint JAERI-USDOE experiment on Doublet-III a record value of beta, 4.6 %, was achieved. Major efforts in theory and computation was on high beta tokamak stability, second stability regions being found for low m internal modes. The JT-60 program progressed as scheduled, installation of the tokamak machine being initiated in February 1983. A 100 kV test was completed of prototype unit for JT-60 NBI. In the development of a high power klystron for JT-60 LH heating, a test fabricated tube generated 1 MW, 10 s RF pulses. Development of TiC coatings for JT-60 first wall was successfully concluded. In the superconducting magnet technology, the Japanese coil for IEA Large Coil Task was installed in a test facility at ORNL after successful performance test at Naka site. A 10 T experiment of a Nb 3 Sn coil with 60 cm inner bore was made. Construction of the Tritium Process Laboratory was started in February 1983. Design studies of the Fusion Experimental Reactor and INTOR were continued. (author)

Core transport properties in JT-60U and JET identity plasmas

NARCIS (Netherlands)

Litaudon, X.; Sakamoto, Y.; de Vries, P. C.; Salmi, A.; Tala, T.; Angioni, C.; Benkadda, S.; Beurskens, M. N. A.; Bourdelle, C.; Brix, M.; Crombe, K.; Fujita, T.; Futatani, S.; Garbet, X.; Giroud, C.; Hawkes, N. C.; Hayashi, N.; Hoang, G. T.; Hogeweij, G. M. D.; Matsunaga, G.; Nakano, T.; Oyama, N.; Parail, V.; Shinohara, K.; Suzuki, T.; Takechi, M.; Takenaga, H.; Takizuka, T.; Urano, H.; Voitsekhovitch, I.; Yoshida, M.

2011-01-01

The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma

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

Conceptual design of the steady state tokamak reactor (SSTR)

International Nuclear Information System (INIS)

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

1992-01-01

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

PPPL tokamak program

International Nuclear Information System (INIS)

Furth, H.P.

1984-10-01

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

Pneumatic pellet injector for JT-60

International Nuclear Information System (INIS)

Onozuka, Masanori; Hiratsuka, Hajime; Kawasaki, Kouzo.

1990-01-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas. (author)

Pneumatic pellet injector for JT-60

Energy Technology Data Exchange (ETDEWEB)

Onozuka, Masanori (Mitsubishi Heavy Industries Ltd., Tokyo (Japan)); Hiratsuka, Hajime; Kawasaki, Kouzo

1990-11-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas. (author).

A synchronization system to digitalize TJ-1 Tokamak data

International Nuclear Information System (INIS)

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

1983-01-01

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

Southwestern Institute of Physics: Annual Report 1998

International Nuclear Information System (INIS)

1999-10-01

The main achievements of controlled nuclear fusion research are presented for Southwestern Institute of Physics in 1998 year. With the establishment and operation of two auxiliary heating systems (NBI, ICRH), the HL-1M Tokamak is equipped with main auxiliary heating and current driving systems such as NBI, ECRH, ICRH and LHCD etc. . In addition, a variety of advanced fueling system, i.e. , multi-shot pellet and supersonic molecular beam injection, the first wall processing technologies of boronization, siliconization and lithiumization as well as more than 20 diagnostic facilities with partial space-time resolution capability have been established on the device. The construction of a larger Tokamak with divertors, the HL-2A, and its complementary systems are being carried out

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

Energy Technology Data Exchange (ETDEWEB)

Wootton, A.J.

1994-07-01

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

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

International Nuclear Information System (INIS)

Wootton, A.J.

1994-07-01

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

Harish-Chandra Research Institute, Allahabad

Indian Academy of Sciences (India)

The Harish-Chandra Research Institute (known as the Mehta Research Institute of Math- ematics and Mathematical Physics until October 2000) came into existence in 1975, with a donation of some land and Rs. 40 lakhs from the B S Mehta Trust in Calcutta. With the aim of converting it into a top-class research Institute in ...

Interlock system for the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Hron, Martin; Sova, J.; Šíba, J.; Kovář, J.; Adámek, Jiří; Pánek, Radomír; Havlíček, Josef; Písačka, Jan; Mlynář, Jan; Stöckel, Jan

2010-01-01

Roč. 85, 3-4 (2010), s. 505-508 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] R&D Projects: GA MŠk 7G09042; GA ČR GD202/08/H057 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak operation * Interlock * Personnel safety Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.143, year: 2010 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6V3C-5003BXW-1&_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=ef5794d05cc6530a905d1de43aa0ac6a&searchtype=a

DIII-D tokamak long range plan. Revision 3

International Nuclear Information System (INIS)

1992-08-01

The DIII-D Tokamak Long Range Plan for controlled thermonuclear magnetic fusion research will be carried out with broad national and international participation. The plan covers: (1) operation of the DIII-D tokamak to conduct research experiments to address needs of the US Magnetic Fusion Program; (2) facility modifications to allow these new experiments to be conducted; and (3) collaborations with other laboratories to integrate DIII-D research into the national and international fusion programs. The period covered by this plan is 1 November 19983 through 31 October 1998

Tokamak engineering mechanics

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Weiyue; Du, Shijun

2014-01-01

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

The Research Progress of the J-TEXT Tokamak

Science.gov (United States)

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

2010-11-01

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

Gas Fuelling System for SST-1Tokamak

Science.gov (United States)

Dhanani, Kalpesh; Raval, D. C.; Khan, Ziauddin; Semwal, Pratibha; George, Siju; Paravastu, Yuvakiran; Thankey, Prashant; Khan, M. S.; Pradhan, Subrata

2017-04-01

SST-1 Tokamak, the first Indian Steady-state Superconducting experimental device is at present under operation in the Institute for Plasma Research. For plasma break down & initiation, piezoelectric valve based gas feed system is implemented as a primary requirement due to its precise control, easy handling, low construction and maintenance cost and its flexibility in the selection of the working gas. Hydrogen gas feeding with piezoelectric valve is used in the SST-1 plasma experiments. The piezoelectric valves used in SST-1 are remotely driven by a PXI based platform and are calibrated before each SST-1 plasma operation with precise control. This paper will present the technical development and the results of the gas fuelling system of SST-1.

Engineering Design of KSTAR tokamak main structure

International Nuclear Information System (INIS)

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

2001-01-01

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

New directions in tokamak reactors

International Nuclear Information System (INIS)

Baker, C.C.

1985-01-01

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

Advanced tokamak physics in DIII-D

Energy Technology Data Exchange (ETDEWEB)

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

2000-12-01

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

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

Tokamak concept innovations

International Nuclear Information System (INIS)

1986-04-01

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

Power and particle exhaust in tokamaks

International Nuclear Information System (INIS)

Stambaugh, R.D.

1998-01-01

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

Kinetic theory of plasma adiabatic major radius compression in tokamaks

International Nuclear Information System (INIS)

Gorelenkova, M.V.; Gorelenkov, N.N.; Azizov, E.A.; Romannikov, A.N.; Herrmann, H.W.

1998-01-01

In order to understand the individual charged particle behavior as well as plasma macroparameters (temperature, density, etc.) during the adiabatic major radius compression (R-compression) in a tokamak, a kinetic approach is used. The perpendicular electric field from the Ohm close-quote s law at zero resistivity is made use of in order to describe particle motion during the R-compression. Expressions for both passing and trapped particle energy and pitch angle change are derived for a plasma with high aspect ratio and circular magnetic surfaces. The particle behavior near the passing trapped boundary during the compression is studied to simulate the compression-induced collisional losses of alpha particles. Qualitative agreement is obtained with the alphas loss measurements in deuterium-tritium (D-T) experiments in the Tokamak Fusion Test Reactor (TFTR) [World Survey of Activities in Controlled Fusion Research [Nucl. Fusion special supplement (1991)] (International Atomic Energy Agency, Vienna, 1991)]. The plasma macroparameters evolution at the R-compression is calculated by solving the gyroaveraged drift kinetic equation. copyright 1998 American Institute of Physics

TPX tokamak construction management

International Nuclear Information System (INIS)

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

1995-01-01

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

Multi-machine studies of the role of turbulence and electric fields in the establishment of improved confinement in tokamak plasma

Czech Academy of Sciences Publication Activity Database

Van Oost, G.; Bulanin, V.V.; Donné, A.J.H.; Gusakov, E.Z.; Kraemer-Flecken, A.; Krupnik, L.I.; Melnikov, A.; Nanobashvili, S.; Peleman, P.; Razumova, K.A.; Stöckel, Jan; Vershkov, V.; Adámek, Jiří; Altukov, A.B.; Andreev, V.F.; Askinazi, L.G.; Bondarenko, I.S.; Brotánková, Jana; Dnestrovskij, A.Yu.; Ďuran, Ivan; Eliseev, L.G.; Esipov, L.A.; Grashin, S.A.; Gurchenko, A.D.; Hogeweij, G.M.D.; Hron, Martin; Ionita, C.; Jachmich, S.; Khrebtov, S.M.; Kouprienko, D.V.; Lysenko, S.E.; Martines, E.; Perfilov, S.V.; Petrov, A.V.; Popov, A.Yu.; Reiser, D.; Schrittwieser, R.; Soldatov, S.; Spolaore, M.; Stepanov, A.Yu.; Telesca, G.; Urazbaev, A.O.; Verdoolaege, G.; Žáček, František; Zimmermann, O.

2007-01-01

Roč. 49, 5A (2007), A29-A44 ISSN 0741-3335. [International Congress on Plasma Physics/13th./. Kiev, 22.05.2006-26.05.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * turbulence * electric fields Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 3.070, year: 2007

JT-60SA power supply system

International Nuclear Information System (INIS)

Coletti, A.; Baulaigue, O.; Cara, P.; Coletti, R.; Ferro, A.; Gaio, E.; Matsukawa, M.; Novello, L.; Santinelli, M.; Shimada, K.; Starace, F.; Terakado, T.; Yamauchi, K.

2011-01-01

The paper describes the main features of the Superconducting Magnets Power Supply to generate the toroidal and poloidal magnetic fields in JT-60SA tokamak, with special regard to coil current regulation mode and magnets protection.

A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

Czech Academy of Sciences Publication Activity Database

Urban, Jakub; Decker, J.; Peysson, Y.; Preinhaelter, Josef; Shevchenko, V.; Taylor, G.; Vahala, L.; Vahala, G.

2011-01-01

Roč. 51, č. 8 (2011), 083050-083050 ISSN 0029-5515 R&D Projects: GA ČR GA202/08/0419; GA MŠk 7G10072 Institutional research plan: CEZ:AV0Z20430508 Keywords : spherical tokamak * electron Bernstein wave (EBW) * heating * current drive * electron cyclotron wave Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.090, year: 2011 http://iopscience.iop.org/0029-5515/51/8/083050/pdf/0029-5515_51_8_083050.pdf

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

International Nuclear Information System (INIS)

Burrell, K.H.

2003-01-01

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

Issues for the electric utilities posed by DT tokamak fusion powerplants

International Nuclear Information System (INIS)

Roth, J.R.

1990-01-01

The DT tokamak is the mainline approach to magnetic fusion energy in all industrialized countries with a major commitment to fusion research. It achieved this status largely through historical accident and not as the result of considered choice among alternatives. After twenty-five years of intensive tokamak research, it is appropriate to ask whether the path down which the tokamak concept is leading the fusion community is the way to an acceptable powerplant for the electric utilities, or an aberration which should be replaced with an approach more promising in the long term. Issues surrounding the DT tokamak can be grouped in three broad areas: physics; safety/environmental; and engineering/economic. In addition to these problems, detailed engineering design studies of DT tokamak fusion powerplants over a twenty year period have revealed a number of additional problems. Most of thee are related to the presence of tritium and energetic neutron fluxes, which tend to make the cost of electricity of DT tokamaks higher than that of fossil or fission powerplants. These safety and economic issues of the DT tokamak powerplant also appear to be intractable, and have not been made to go away by twenty years of progressively more detailed and extensive engineering design studies

Dynamic simulations of the cryogenic system of a tokamak

International Nuclear Information System (INIS)

Cirillo, R.; Hoa, C.; Michel, F.; Rousset, B.; Poncet, J.M.

2015-01-01

In a tokamak plasma confinement is achieved through high magnetic fields generated by superconductive coils that need to be cooled down to 4.4 K with a forced flow of supercritical Helium. Tokamak's coil system works cyclically and so it is subject to pulsed heat loads which have to be handled by the refrigerator. This latter has to be sized on the average power value and not according to the peak to limit investment and operation costs and hence the heat load needs to be smoothed. CEA Grenoble is in charge of providing the cryogenic system for the Japanese tokamak JT60-SA, currently under construction in Naka (Japan). Hence, in order to model and study the smoothing strategies, an experimental set up: HELIOS (Helium Loop for high load smoothing) has been built. This is a scaled down model (1:20) of the helium distribution system whose main components are a saturated helium bath and a supercritical helium loop. This large installation can reproduce conditions of pressure, temperature and transport times, similar to those expected in the cooling circuits of the central solenoid superconducting magnets of JT-60SA. The peak loads representative of the tokamak operation have been reproduced and smoothed before they arrive in the refrigerator, by means of a saturated helium bath (thermal reservoir). A dynamic modelling of the cryogenic system is presented, with results on the pulsed load scenarios. All the simulations have been performed with EcosimPro software developed and the cryogenic library: CRYOLIB. This document is made up of an abstract and the slides of the presentation

The collaborative tokamak control room

International Nuclear Information System (INIS)

Schissel, D.P.

2006-01-01

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

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

International Nuclear Information System (INIS)

Wan Yuanxi

2002-01-01

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

Computer Science Research Institute 2005 annual report of activities.

Energy Technology Data Exchange (ETDEWEB)

Watts, Bernadette M.; Collis, Samuel Scott; Ceballos, Deanna Rose; Womble, David Eugene

2008-04-01

This report summarizes the activities of the Computer Science Research Institute (CSRI) at Sandia National Laboratories during the period January 1, 2005 to December 31, 2005. During this period, the CSRI hosted 182 visitors representing 83 universities, companies and laboratories. Of these, 60 were summer students or faculty. The CSRI partially sponsored 2 workshops and also organized and was the primary host for 3 workshops. These 3 CSRI sponsored workshops had 105 participants, 78 from universities, companies and laboratories, and 27 from Sandia. Finally, the CSRI sponsored 12 long-term collaborative research projects and 3 Sabbaticals.

Improvement of tokamak performance by injection of electrons

International Nuclear Information System (INIS)

Ono, Masayuki.

1992-12-01

Concepts for improving tokamak performance by utilizing injection of hot electrons are discussed. Motivation of this paper is to introduce the research work being performed in this area and to refer the interested readers to the literature for more detail. The electron injection based concepts presented here have been developed in the CDX, CCT, and CDX-U tokamak facilities. The following three promising application areas of electron injection are described here: 1. Non-inductive current drive, 2. Plasma preionization for tokamak start-up assist, and 3. Charging-up of tokamak flux surfaces for improved plasma confinement. The main motivation for the dc-helicity injection current drive is in its efficiency that, in theory, is independent of plasma density. This property makes it attractive for driving currents in high density reactor plasmas

Max-Planck-Institut fuer Plasmaphysik. Annual report 1993

International Nuclear Information System (INIS)

1993-01-01

In 1993 the first particle injector of the ASDEX Upgrade divertor tokamak was put into operation up to 6 MW of heating power. The main diagnostics were put into operation as also was a newly developed pellet centrifuge. At the Wendelstein 7-AS stellarator experiment the cooperation with Russian and German research institutes an electron cyclotron resonance heating was successful. The works of the Berlin Division of IPP and the coordination of research efforts with Kernforschungszentrum Karlsruhe are reported. (DG)

Removal of the lower hybrid (LH) frequency time scale in test electron simulations of LH-induced tokamak edge electron flow

Czech Academy of Sciences Publication Activity Database

Fuchs, Vladimír; Petržílka, Václav; Gunn, J. P.; Goniche, M.

2002-01-01

Roč. 52, supplement D (2002), s. 45-50 ISSN 0011-4626. [Symposium on Plasma Physics and Technology/20th./. Prague, 10.06.2002-13.06.2002] Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, edge electrons, lower hybrid antenna Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.311, year: 2002

Digital control of plasma position in Damavand tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2002-03-01

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

Researches at organizations of minatom of Russia

International Nuclear Information System (INIS)

Beljaev, I.A.

1994-01-01

The brief information about the major scientific research institute of the Ministry for Atomic Energy of the Russian Federation such as the 'Kurchatov Institute' of Atomic Energy, the Institute of Theoretical and Experimental Physics, the Institute of High Energy Physics, the Institute of Innovation and Thermonuclear Research, etc., and their outstanding research achievements have been discussed. The activity of applied research organizations has also been written. Informations about the installations such as TOKAMAK, TOPAZ, the unique 3000 GeV proton accelerator in Protvino are presented too. The scientific problems of nuclear weapons and disarmament, aspects of conversation of the enterprises of nuclear weapons complex have been spoken about. The main fields of international scientific cooperation are also discussed. (author)

DIII-D Advanced Tokamak Research Overview

International Nuclear Information System (INIS)

V.S. Chan; C.M. Greenfield; L.L. Lao; T.C. Luce; C.C. Petty; G.M. Staebler

1999-01-01

This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously β N H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues

Development of Atomic Beam Probe for tokamaks

Czech Academy of Sciences Publication Activity Database

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

2013-01-01

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

Decay of enhanced density and damping of plasma flows after the electrode biasing terminaton on the CASTOR tokamak

Czech Academy of Sciences Publication Activity Database

Hron, Martin; Ďuran, Ivan; Stöckel, Jan; Hidalgo, C.

2004-01-01

Roč. 54, suppl. C (2004), C22-C27 ISSN 0011-4626. [Symposium on Plasma Physics and Technology /21st/. Praha, 14.06.2004-17.06.2004] R&D Projects: GA ČR GA202/03/0786 Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, edge plasma, polarization Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.292, year: 2004

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

International Nuclear Information System (INIS)

2000-10-01

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

Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of); Chung, Kyoo Sun [Hanyang Univ., Seoul (Korea, Republic of); Hong, Sang Heui [Seoul National Univ., Seoul (Korea, Republic of); Kang, Heui Dong [Kyungpook National Univ., Taegu (Korea, Republic of); Lee, Jae Koo [Pohang Inst. of Science and Technology, Kyungnam (Korea, Republic of)

1994-11-01

A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the `advanced tokamak` physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs.

Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

International Nuclear Information System (INIS)

Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub; Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui; Chung, Kyoo Sun; Hong, Sang Heui; Kang, Heui Dong; Lee, Jae Koo

1994-11-01

A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the 'advanced tokamak' physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs

[German research institute/Max-Planck Institute for psychiatry].

Science.gov (United States)

Ploog, D

1999-12-01

The Deutsche Forschungsanstalt für Psychiatrie (DFA, German Institute for Psychiatric Research) in Munich was founded in 1917 bel Emil Kraepelin. For a long time it was the only institution in Germany entirely devoted to psychiatric research. Because of its strictly science-oriented and multidisciplinary approach it also became a model for institutions elsewhere. Kraepelin's ideas have certainly had a strong influence on psychiatry in the twentieth century. The fascinating and instructive history of the DFA reflects the central issues and determinants of psychiatric research. First, talented individuals are needed to conduct such research, and there was no lack in this regard. Second, the various topics chosen are dependent on the available methods and resources. And finally, the issues addressed and the ethical standards of the researchers are heavily dependent on the zeitgeist, as is evident in the three epochs of research at the DFA, from 1917 to 1933, from 1933 to 1945, and from the postwar period to the present. With the introduction of molecular biology and neuroimaging techniques into psychiatric research a change in paradigm took place and a new phase of the current epoch began.

The physics of tokamak start-up

International Nuclear Information System (INIS)

Mueller, D.

2013-01-01

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

Spectra in the 60 /angstrom/ to 345 /angstrom/ wavelength region of elements injected into the PLT tokamak

International Nuclear Information System (INIS)

Wouters, A.; Schwob, J.L.; Suckewer, S.; Seely, J.F.; Feldman, U.; Dave, J.H.

1988-03-01

High resolution spectra of the elements Fe, Ni, Zn, Ge, Se, and Mo injected into the PLT tokamak were recorded by the 2-meter Schwob-Fraenkel soft X-ray multichannel spectrometer (SOXMOS). Spectra were recorded every 50 ms during the time before and after injection. The spectral lines of the injected element were very strong in the spectrum recorded immedately after injection, and the transition in the injected element were easily distinguished from the transitions in te intrinsic elements (C, O, Ti, Cr, Fe, and Ni). An accurate wavelength scale was established using well-known reference transitions in the intrinsic elements. The spectra recorded just prior to injection were substracted from the spectra recorded after injection, and the resulting spectrum was composed almost entirely of transitions from the injected element. A large number of Δn + 0 transitions between the ground and the first excited configurations in the Li I through K I isoelectronic sequences of the injected elements were identified in the wavelength region 60 /angstrom/ to 345 /angstrom/. 33 refs., 5 figs., 1 tab

Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

Science.gov (United States)

Vdovin, V.

2014-02-01

The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

Energy Technology Data Exchange (ETDEWEB)

Vdovin, V. [NRC Kurchatov Institute Tokamak Physics Institute, Moscow (Russian Federation)

2014-02-12

The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

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

ICRF heating experiments in JFT-2 tokamak

International Nuclear Information System (INIS)

Matsumoto, Hiroshi

1986-01-01

This is an experimental study of ICRF heating on JFT-2 Tokamak in Japan Atomic Energy Research Institute. In this study, we first clarified physical and engineering problems of ICRF heating of tokamak plasma. Next, we optimized the design of the ICRF heating system, and the plasma parameters for the heating. Finally, we could demonstrate a high efficiency of this additional heating method by launching RF power which is two or three times as large as an ohmic input power to a plasma. And we achieved following things. (1) We optimized a design of an antenna, and we improved a durability of the system for high voltage. With the result that we achieved the maximum power density on an antenna. (2) We demonstrated that electron heating regime and ion heating regime can be easily accessed by controlling plasma parameters. Also we found the optimum heating conditions in each heating regime. (3) We experimentally clarified the production mechanism of impurities during ICRF heating. We could reduce the influx of metal impurity ions to a plasma by employing low z materials for limiters and antenna shields. Consequently, we improved a heating efficiency of electrons. Next, we studied a power balance of plasma during ICRF heating, and we could compare heating characteristics of ICRF with other additional heatings on JFT-2. (author)

Institutional Researchers' Use of Qualitative Research Methods for Institutional Accountability at Two Year Colleges in Texas

Science.gov (United States)

Sethna, Bishar M.

2011-01-01

This study examined institutional researchers' use of qualitative methods to document institutional accountability and effectiveness at two-year colleges in Texas. Participants were Institutional Research and Effectiveness personnel. Data were collected through a survey consisting of closed and open ended questions which was administered…

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

International Nuclear Information System (INIS)

Jackson, M.C.

1989-01-01

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

Annual report of the Fusion Research Center for the period of April 1, 1983 to March 31, 1984

International Nuclear Information System (INIS)

1985-03-01

Research and development activities of the Fusion Research Center (Department of Thermonuclear Fusion Research and Department of Large Tokamak Development) from April 1983 to March 1984 are described. Installation and commissioning of the new tokamak JFT-2M had been completed. The 2nd ICRF heating experiment and LH current drive experiment were started. In the field of plasma theory, the scaling law of the critical beta in a tokamak was obtained and the ICRF heating was analyzed in detail. The first phase of the cooperation of Doublet III will be finished in Sept. 1984. The JT-60 program progressed as scheduled. Installation of the tokamak machine, initiated in Feb. 1983, will be finished in Sept. 1984. The tests of power supply and control system on site and the fabrication of the neutral beam injectors in factory proceeded successfully. Performance tests of prototype injector unit for JT-60 NBI progressed as scheduled. A new advanced source plasma generator was developed to provide a high proton ratio exceeding 90%. Klystrons for JT-60 LH heating achieved the output power of 1 MW for 10 sec. Performance tests of titanium evaporators for JT-60 were completed. The Japanese coil for IEA Large Coil Task was installed in a test facility at ORNL and the partial cool-down was carried out. Construction of the Tritium Process Laboratory was completed. Design studies of the Fusion Experimental Reactor (FER) and INTOR proceeded. (author)

Tokamak TCABR: Acquisition system, data analysis, and remote participation using MDSplus

International Nuclear Information System (INIS)

Sá, W.P. de

2012-01-01

Highlights: ► The implementation of MDSplus in TCABR tokamak. ► Interfaces between the system already installed and the MDSplus. - Abstract: Each plasma physics laboratory has a proprietary scheme to control and data acquisition system. Usually, it is different from one laboratory to another. It means that each laboratory has its own way to control the experiment and retrieving data from the database. Fusion research relies to a great extent on international collaboration and this private system makes it difficult to follow the work remotely. The TCABR data analysis and acquisition system has been upgraded to support a joint research programme using remote participation technologies. The choice of MDSplus (Model Driven System plus) is proved by the fact that it is widely utilized, and the scientists from different institutions may use the same system in different experiments in different tokamaks without the need to know how each system treats its acquisition system and data analysis. Another important point is the fact that the MDSplus has a library system that allows communication between different types of language (JAVA, Fortran, C, C++, Python) and programs such as MATLAB, IDL, OCTAVE. In the case of tokamak TCABR interfaces (object of this paper) between the system already in use and MDSplus were developed, instead of using the MDSplus at all stages, from the control, and data acquisition to the data analysis. This was done in the way to preserve a complex system already in operation and otherwise it would take a long time to migrate. This implementation also allows add new components using the MDSplus fully at all stages.

Tokamak TCABR: Acquisition system, data analysis, and remote participation using MDSplus

Energy Technology Data Exchange (ETDEWEB)

Sa, W.P. de, E-mail: pires@if.usp.br [Instituto de Fisica, Universidade de Sao Paulo, Rua do Matao, Travessa R, 187, CEP 05508-090 Cidade Universitaria, Sao Paulo (Brazil)

2012-12-15

Highlights: Black-Right-Pointing-Pointer The implementation of MDSplus in TCABR tokamak. Black-Right-Pointing-Pointer Interfaces between the system already installed and the MDSplus. - Abstract: Each plasma physics laboratory has a proprietary scheme to control and data acquisition system. Usually, it is different from one laboratory to another. It means that each laboratory has its own way to control the experiment and retrieving data from the database. Fusion research relies to a great extent on international collaboration and this private system makes it difficult to follow the work remotely. The TCABR data analysis and acquisition system has been upgraded to support a joint research programme using remote participation technologies. The choice of MDSplus (Model Driven System plus) is proved by the fact that it is widely utilized, and the scientists from different institutions may use the same system in different experiments in different tokamaks without the need to know how each system treats its acquisition system and data analysis. Another important point is the fact that the MDSplus has a library system that allows communication between different types of language (JAVA, Fortran, C, C++, Python) and programs such as MATLAB, IDL, OCTAVE. In the case of tokamak TCABR interfaces (object of this paper) between the system already in use and MDSplus were developed, instead of using the MDSplus at all stages, from the control, and data acquisition to the data analysis. This was done in the way to preserve a complex system already in operation and otherwise it would take a long time to migrate. This implementation also allows add new components using the MDSplus fully at all stages.

Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

International Nuclear Information System (INIS)

Menard, J.E.; Bromberg, L.; Brown, T.; Burgess, Thomas W.; Dix, D.; Gerrity, T.; Goldston, R.J.; Hawryluk, R.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G.H.; Neumeyer, C.L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.G.; Zarnstorff, M.C.

2011-01-01

A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

New design of microwave interferometer for tokamak COMPASS

Czech Academy of Sciences Publication Activity Database

Varavin, Mykyta; Zajac, Jaromír; Žáček, František; Nanobashvili, S.I.; Ermak, G.; Varavin, A.; Vasilev, A.; Stumbra, M.; Vetoshko, A.; Fateev, A.; Shevchenko, V.

2014-01-01

Roč. 73, č. 10 (2014), s. 935-942 ISSN 0040-2508 Institutional support: RVO:61389021 Keywords : Microwave interferometry * Plasma diagnostics * Tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://www.dl.begellhouse.com/journals/0632a9d54950b268,0f0577fd35766f78,6e6cc5d616079cba.html

A systems analysis of the ARIES tokamak reactors

International Nuclear Information System (INIS)

Bathke, C.G.

1992-01-01

The multi-institutional ARIES study has completed a series of cost-of-electricity optimized conceptual designs of commercial tokamak fusion reactors that vary the assumed advances in technology and physics. A comparison of these designs indicates the cost benefit of various design options. A parametric systems analysis suggests a possible means to obtain a marginally competitive fusion reactor

Development of 3D ferromagnetic model of tokamak core withstrong toroidal asymmetry

Czech Academy of Sciences Publication Activity Database

Markovič, Tomáš; Gryaznevich, M.; Ďuran, Ivan; Svoboda, V.; Pánek, Radomír

96-97, October (2015), s. 302-305 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA ČR GAP205/11/2341; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : tokamak * ferromagnetic core * model of ferromagnet * integral method * tokamak GOLEM Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015 http://www.sciencedirect.com/science/article/pii/S0920379615002100

Personnel protection during the operation of Thomson scattering laser system on COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Böhm, Petr; Hron, Martin; Kovar, J.; Sova, J.; Zvolanek, M.; Aftanas, Milan; Bílková, Petra; Pánek, Radomír; Walsh, M.J.

2011-01-01

Roč. 86, 6-8 (2011), s. 699-702 ISSN 0920-3796. [Symposium on Fusion Technology, SOFT-26/26th./. Porto, 27.09.2010-01.10.2010] R&D Projects: GA ČR GA202/09/1467; GA ČR GD202/08/H057; GA MŠk 7G09042 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * Thomson scattering * Laser safety * Personnel protection * PLC Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.490, year: 2011 http://www.sciencedirect.com/science/article/pii/S0920379611002432

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

International Nuclear Information System (INIS)

1988-05-01

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

Research and development of the JAERI large tokamak (JT-60), (4)

International Nuclear Information System (INIS)

Takashima, Tetsuo; Shimizu, Masatsugu; Ohta, Mitsuru; Minaguchi, Tadayoshi; Maeda, Hideto.

1978-01-01

A pair of fast-acting movable rail limiters are to be installed in the vacuum chamber of JT-60 to suppress skin current in the plasma column. They should travel across the vacuum chamber over a stroke of about 1 m in 0.1 sec in the build-up phase of the plasma current. Each movable limiter system consists of a drive mechanism, a vacuum seal, a bearing usable at high temperatures in a vacuum, a molybdenum rail limiter head and its auxiliary members. Various engineering problems are involved in constructing such a system because the design specifications outlined above exceed the present technology. A full-scale movable limiter, therefore, was designed, constructed and then put to mechanical, electrical and vacuum-technological tests. The model features a hydraulic drive mechanism with servovalves to control the oil flow. A special vacuum seal allowing a movement at high speeds was developed. It consists of welded bellows jointed together and connected to a pantograph at the joints. It allows uniform expansion of each bellows at high speeds. Molybdenum disulphide with 20% Ta is chosen as the most suitable bearing material after conducting tests on various bearing materials. The overall test of the model showed that its specifications were met with satisfactory reliability and reproducibility. Furthermore, the endurance test demonstrated that it functioned reliably over 50,000 times of operation. (author)

Runaway acceleration during magnetic reconnection in tokamaks

International Nuclear Information System (INIS)

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

2002-01-01

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

Institutional Support : Kenya Institute for Public Policy Research and ...

International Development Research Centre (IDRC) Digital Library (Canada)

In 2006 the Government of Kenya passed an Act of Parliament making the Kenya Institute for Public Policy Research and Analysis (KIPPRA) the government's lead socioeconomic research institute. The Act exerts enormous demands on KIPPRA at a time when it is trying to recover from the senior staff turnover suffered in ...

System of institutional radioactive waste management in the Nuclear Research Institute Rez plc

International Nuclear Information System (INIS)

Podlaha, J.; Burian, P.

2005-01-01

The Nuclear Research Institute Rez plc (NRI) is a leading institution in the area of nuclear Research and Development in the Czech Republic. The NRI has had a dominant position in the nuclear programme since it was established in 1955 as a state-owned research organization and it has developed to its current status. In December 1992 the NRI has been transformed into a joint-stock company. The NRI's activity encompasses nuclear physics, chemistry, nuclear power, experiments at the research reactor and many other topics. Main issues addressed in the NRI in the past decades were concentrated on research, development and services provided to the nuclear power plants operating WWER reactors, development of chemical technologies for fuel cycle and irradiation services to research and development in the industrial sector, agriculture, food processing and medicine. At present the research activities are mainly targeted to assist the State Office for Nuclear Safety -the nuclear safety regulating body, power plant operator and nuclear facilities contractors. Significant attention is also paid to the use of nuclear technology outside the nuclear power sector, providing a wide range of services to industry , medicine and the preparation of radiopharmaceuticals. NRI operates two research nuclear reactors and another facilities such as a hot cell facility , research laboratories, technology for radioactive waste (RAW) management, 60 Co irradiators, an electron accelerator, etc. In this paper the Centre of RAW management, system of RAW management, facilities for RAW management as well as decontamination and decommissioning activities of the NRI are presented. The NRI provides complex services in the area of RAW management and has gained many experience and full qualification not only in this area but also in the area of decontamination and decommissioning and spent fuel management. The NRI guarantees safe RAW and spent fuel management. (authors)

Hybrid model for simulation of plasma jet injection in tokamak

Science.gov (United States)

Galkin, Sergei A.; Bogatu, I. N.

2016-10-01

Hybrid kinetic model of plasma treats the ions as kinetic particles and the electrons as charge neutralizing massless fluid. The model is essentially applicable when most of the energy is concentrated in the ions rather than in the electrons, i.e. it is well suited for the high-density hyper-velocity C60 plasma jet. The hybrid model separates the slower ion time scale from the faster electron time scale, which becomes disregardable. That is why hybrid codes consistently outperform the traditional PIC codes in computational efficiency, still resolving kinetic ions effects. We discuss 2D hybrid model and code with exact energy conservation numerical algorithm and present some results of its application to simulation of C60 plasma jet penetration through tokamak-like magnetic barrier. We also examine the 3D model/code extension and its possible applications to tokamak and ionospheric plasmas. The work is supported in part by US DOE DE-SC0015776 Grant.

Using institutional theory in enterprise systems research

DEFF Research Database (Denmark)

Svejvig, Per

2013-01-01

This paper sets out to examine the use of institutional theory as a conceptually rich lens to study social issues of enterprise systems (ES) research. More precisely, the purpose is to categorize current ES research using institutional theory to develop a conceptual model that advances ES research...... model that advocates multi-level and multi-theory approaches and applies newer institutional aspects such as institutional logics. The findings show that institutional theory in ES research is in its infancy and adopts mainly traditional institutional aspects like isomorphism, with the organization....... Key institutional features are presented such as isomorphism, rationalized myths, and bridging macro and micro structures, and institutional logics and their implications for ES research are discussed. Through a literature review of 181 articles, of which 18 papers are selected, we build a conceptual...

Status of the tokamak program

Science.gov (United States)

Sheffield, J.

1981-08-01

For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

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.)

Recent results of JT-60U ICRF antenna operation

International Nuclear Information System (INIS)

Fujii, T.; Saigusa, M.; Kimura, H.

1994-01-01

Ion cyclotron range of frequencies (ICRF) heating is one of attractive plasma heating methods for reactor grade tokamaks, because it is quite effective in the wide ranges of plasma density and temperature. An antenna which should inject high power into plasma has been developed intensively because the heating efficiency and the coupling properties depend on its design. The antenna was operated at a small antenna-plasma gap in the JT-60 in out of phase mode, which showed the high heating efficiency to obtain high loading resistance, and similarly to other tokamaks. However, in order to reduce heat load to the antenna from plasma, a wide gap is required in reactor grade tokamaks such as ITER, in which the gap is designed to be 0.15 m in CDA. Two new antennas were fabricated for the JT-60U, which were designed to obtain high loading resistance at a wide gap for (π,0) phasing. The JT-60U ICRF heating system is explained. Also the JT-60U antenna is described. Antenna conditioning has been conducted well in the initial operation period. The phasing mode was set at (π,0) phasing, in which high heating efficiency is expected. The procedure is explained. The coupling and radiation loss properties during ICRF heating are reported. The JT-60U ICRF antennas were conditioned quickly with about 70 shots. The maximum coupled power was 6.4 MW for (π,0) phasing, and the power density was 6.1 MW/m 2 . (K.I.)

Timing and triggering of the Thomson scattering diagnostics on the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Mikulín, Ondřej; Hron, Martin; Böhm, Petr; Naylor, G.; Bílková, Petra; Janky, Filip; Salášek, J.; Pánek, Radomír

2014-01-01

Roč. 89, č. 5 (2014), s. 693-697 ISSN 0920-3796. [The 9th Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/9./. Hefei, 06.05.2013-10.05.2013] R&D Projects: GA MŠk 7G10072; GA MŠk(CZ) LM2011021; GA ČR GAP205/11/2470 Institutional support: RVO:61389021 Keywords : Tokamak * Timing and triggering * FPGA * Real-time control * Diagnostics control * Thomson scattering Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.152, year: 2014 http://www.sciencedirect.com/science/article/pii/S0920379614002105#

Physical design of JT-60 Super Upgrade

International Nuclear Information System (INIS)

Nagashima, K.; Kikuchi, M.; Kurita, G.; Ozeki, T.; Aoyagi, T.; Ushigusa, K.; Neyatani, Y.; Kubo, T.; Mori, K.; Nakagawa, S.; Kuriyama, M.; Nagami, M.

1997-01-01

The JT-60 Super Upgrade (JT-60SU) is an upgraded tokamak device of JT-60U for developing the steady-state reactor and advanced tokamak operation in the International Thermonuclear Experimental Reactor. The device is planned to utilize the JT-60 facilities fully and to minimize the needed modification. The major radius is 4.8 m and the maximum plasma current is 10 MA. Neutral beam injection with 750 keV beam energy is the primary heating method. The machine is capable of steady-state operation with high density up to 8.8 x 10 19 m -3 at 5 MA plasma current. The high operating density, over the Greenwald et al. limit, is critically important in order to achieve high bootstrap current fraction. Ballooning mode and low n ideal magnetohydrodynamic (MHD) mode including the bootstrap current were analyzed for steady-state operation. The current profile must be optimized to obtain a normalized beta up to 3. The plasma configuration with high triangularity was adopted in order to get good MHD stability and high energy confinement. A compact divertor was designed in order to get the large plasma space. (orig.)

The Knowledge Management Research of Agricultural Scientific Research Institution

Institute of Scientific and Technical Information of China (English)

2010-01-01

Based on the perception of knowledge management from experts specializing in different fields,and experts at home and abroad,the knowledge management of agricultural scientific research institution can build new platform,offer new approach for realization of explicit or tacit knowledge,and promote resilience and innovative ability of scientific research institution.The thesis has introduced functions of knowledge management research of agricultural science.First,it can transform the tacit knowledge into explicit knowledge.Second,it can make all the scientific personnel share knowledge.Third,it is beneficial to the development of prototype system of knowledge management.Fourth,it mainly researches the realization of knowledge management system.Fifth,it can manage the external knowledge via competitive intelligence.Sixth,it can foster talents of knowledge management for agricultural scientific research institution.Seventh,it offers the decision-making service for leaders to manage scientific program.The thesis also discusses the content of knowledge management of agricultural scientific research institution as follows:production and innovation of knowledge;attainment and organizing of knowledge;dissemination and share of knowledge;management of human resources and the construction and management of infrastructure.We have put forward corresponding countermeasures to further reinforce the knowledge management research of agricultural scientific research institution.

Effects of pressure profile and plasma shaping on the n=1 internal kink mode in JT-60/JT-60U pellet fuelled plasmas

International Nuclear Information System (INIS)

Ozeki, Takahisa; Azumi, Masafumi

1990-10-01

The stability of the n=1 internal kink mode in a tokamak is numerically analyzed for plasmas with a centrally peaked pressure profile. These studies are carried out with the strongly peaked pressure inside the q=1 surface, which is based on the experimentally observed plasmas by means of injections of hydrogen-ice pellets in JT-60 tokamak. The effects of peaked pressure and shaping, i.e., elongation and triangularity, are also studied for JT-60U tokamak. The plasma with the strongly peaked pressure profile has higher critical value of poloidal beta defined within the q=1 surface than that with a parabolic pressure profile. Though the beta limit reduces with the increase of the elongation, the plasma with the peaked pressure profile has larger improvement due to the triangularity than that with the parabolic pressure profile. To access the second stability of the n=1 internal kink mode, the plasma with a flat pressure profile and the large minor radius of the q=1 surface is effective. (author)

Tokamak engineering mechanics

CERN Document Server

Song, Yuntao; Du, Shijun

2013-01-01

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

Validation of equilibrium tools on the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Urban, Jakub; Appel, L.C.; Artaud, J.; Faugeras, B.; Havlíček, Josef; Komm, Michael; Lupelli, I.; Peterka, Matěj

96-97, October (2015), s. 998-1001 ISSN 0920-3796. [Symposium on Fusion Technology 2014(SOFT-28)/28./. San Sebastián, 29.09.2014-03.10.2014] R&D Projects: GA ČR GP13-38121P Institutional support: RVO:61389021 Keywords : Tokamak * Equilibrium * COMPASS Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.301, year: 2015

The tokamak as a neutron source

International Nuclear Information System (INIS)

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

1989-11-01

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

Module description of TOKAMAK equilibrium code MEUDAS

International Nuclear Information System (INIS)

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

2002-01-01

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

Midplane Faraday rotation: A densitometer for large tokamaks

International Nuclear Information System (INIS)

Jobes, F.C.; Mansfield, D.K.

1992-01-01

The density in a large tokamak such as International Thermonuclear Experimental Reactor (ITER), or any of the proposed future US machines, can be determined by measuring the Faraday rotation of a 10.6 μm laser directed tangent to the toroidal field. If there is a horizontal array of such beams, then n e (R) can be readily obtained with a simple Abel inversion about the center line of the tokamak. For a large machine, operated at a full field of 30 T m and a density of 2x10 20 /m 3 , the rotation angle would be quite large-about 60 degree for two passes. A layout in which a single laser beam is fanned out in the horizontal midplane of the tokamak, with a set of retroreflectors on the far side of the vacuum vessel, would provide good spatial resolution, depending only upon the number of reflectors. With this proposed layout, only one window would be needed. Because the rotation angle is never more than 1 ''fringe,'' the data is always good, and it is also a continuous measurement in time. Faraday rotation is dependent only upon the plasma itself, and thus is not sensitive to vibration of the optical components. Simulations of the expected results show that ITER, or any large tokamak, existing or proposed, would be well served even at low densities by a midplane Faraday rotation densitometer of ∼64 channels

Do Research Participants Trust Researchers or Their Institution?

Science.gov (United States)

Guillemin, Marilys; Barnard, Emma; Allen, Anton; Stewart, Paul; Walker, Hannah; Rosenthal, Doreen; Gillam, Lynn

2018-07-01

Relationships of trust between research participants and researchers are often considered paramount to successful research; however, we know little about participants' perspectives. We examined whom research participants trusted when taking part in research. Using a qualitative approach, we interviewed 36 research participants, including eight Indigenous participants. Thematic analysis was used to analyze the data. This article focuses on findings related to non-Indigenous participants. In contrast to Indigenous participants, non-Indigenous participants placed their trust in research institutions because of their systems of research ethics, their reputation and prestige. Researchers working in non-Indigenous contexts need to be cognizant that the trust that participants place in them is closely connected with the trust that participants have in the institution.

Experimental results from the TUMAN 3 tokamak

International Nuclear Information System (INIS)

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

1995-01-01

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

Low Vision Research at the Schepens Eye Research Institute

National Research Council Canada - National Science Library

D'Amore, Patricia

2003-01-01

This research proposal, Low Vision at the Schepens Eye Research Institute, is a collaborative effort on the part of four Investigators at the Institute whose goal is to advance the studies on low vision...

Tokamak ARC damage

International Nuclear Information System (INIS)

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

1985-01-01

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

Tokamak ARC damage

Energy Technology Data Exchange (ETDEWEB)

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

1985-01-01

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

Numerical and experimental analysis of eddy currents induced in tokamak machines

International Nuclear Information System (INIS)

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

1977-01-01

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

Twenty Years of Research on the Alcator C-Mod Tokamak

Science.gov (United States)

Greenwald, Martin

2013-10-01

Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

Relativistic runaway electrons in tokamak plasmas

International Nuclear Information System (INIS)

Jaspers, R.E.

1995-01-01

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

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

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

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

Filamentary probe on the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

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

2017-01-01

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

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)

Beta-limit of a large tokamak with a circular cross-section

International Nuclear Information System (INIS)

Tsunematsu, Toshihide; Takeda, Tatsuoki; Kurita, Gen-ichi; Azumi, Masafumi; Matsuura, Toshihiko; Gruber, R.; Troyon, F.

1982-01-01

The dependence of stabilizing effect of a conducting shell on a poloidal beta value (βsub(p)) is investigated as to instabilities with low toroidal mode numbers (n = 1 and 2) for a tokamak with a circular cross-section such as JT-60. The n = 1 mode is completely stabilized by the conducting shell which is located at a practically possible position and the critical position of the shell becomes closer to the plasma surface with increasing βsub(p). The stabilizing effect on the n = 2 mode is remarkable for higher βsub(p) when the shell is placed sufficiently close to the plasma surface but the shell far from the plasma surface has hardly an effect on the stability property of a higher βsub(p) plasma. It is concluded that critical β of about 2% is attainable even in a standard circular tokamak such as JT-60 and higher β value is also expected by taking advantage of the closely located conducting shell. (author)

Proceedings of the workshop of three large tokamak cooperation on energy confinement scaling under intensive auxiliary heating, May 18 ∼ 20, 1992, Naka

International Nuclear Information System (INIS)

1992-09-01

The workshop of three large tokamak cooperation W22 on 'Energy confinement scaling under intensive auxiliary heating' was held 18-20 May, 1992, at Naka Fusion Research Establishment. This proceedings compiles 14 synopses of contributions (5 from JET, 4 from JT-60, 3 from TFTR, and 1 each from DIII-D JFT-2M) and the summary of the workshop. Topic sections are ; (i) L-mode confinement and scaling, (ii) Confinement at high β P regimes, Supershots, High poloidal beta enhanced confinement mode etc., (iii) Confinement at various H-mode regimes and scaling (including the VH-mode), (iv) Characteristic time scales for present tokamak regimes, and (v) Theoretical comparison with experimental data. (author)

Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

Science.gov (United States)

Pathan, F. S.; Khan, Z.; Semwal, P.; Raval, D. C.; Joshi, K. S.; Thankey, P. L.; Dhanani, K. R.

2008-05-01

Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN2 panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN2 panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN2 panels during sniffer test and pressure drop test using helium gas.

Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

Pathan, F S; Khan, Z; Semwal, P; Raval, D C; Joshi, K S; Thankey, P L; Dhanani, K R [Institute for Plasma Research, Bhat, Gandhinagar - 382 428, Gujarat (India)], E-mail: firose@ipr.res.in

2008-05-01

Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN{sub 2} panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN{sub 2} panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN{sub 2} panels during sniffer test and pressure drop test using helium gas.

Quantitative study of sniffer leak rate and pressure drop leak rate of liquid nitrogen panels of SST-1 tokamak

International Nuclear Information System (INIS)

Pathan, F S; Khan, Z; Semwal, P; Raval, D C; Joshi, K S; Thankey, P L; Dhanani, K R

2008-01-01

Steady State Super-conducting (SST-1) Tokamak is in commissioning stage at Institute for Plasma Research. Vacuum chamber of SST-1 Tokamak consists of 1) Vacuum vessel, an ultra high vacuum (UHV) chamber, 2) Cryostat, a high vacuum (HV) chamber. Cryostat encloses the liquid helium cooled super-conducting magnets (TF and PF), which require the thermal radiation protection against room temperature. Liquid nitrogen (LN2) cooled panels are used to provide thermal shield around super-conducting magnets. During operation, LN 2 panels will be under pressurized condition and its surrounding (cryostat) will be at high vacuum. Hence, LN 2 panels must have very low leak rate. This paper describes an experiment to study the behaviour of the leaks in LN 2 panels during sniffer test and pressure drop test using helium gas

Institutional Support : Ethiopian Development Research Institute ...

International Development Research Centre (IDRC) Digital Library (Canada)

The Ethiopian Development Research Institute (EDRI) was established in 1999 and became operational in 2003 as a semi-autonomous organization accountable to ... International Water Resources Association, in close collaboration with IDRC, is holding a webinar titled “Climate change and adaptive water management: ...

What is past is prologue: future directions in Tokamak Power Reactor Design Research

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

After reviewing the first generation of studies and the primary conclusions they produced, four current designs are discussed that are representative of present trends in this area of research. In particular, the trends towards reduced reactor size and higher neutron wall loadings are discussed. Moving in this direction requires new approaches to many subsystem designs. New approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets are described. A discussion is given of the future role of conceptual reactor design research and the need for close interactions with ongoing experiments in fusion technology

Research misconduct definitions adopted by U.S. research institutions.

Science.gov (United States)

Resnik, David B; Neal, Talicia; Raymond, Austin; Kissling, Grace E

2015-01-01

In 2000, the U.S. federal government adopted a uniform definition of research misconduct as fabrication, falsification, or plagiarism (FFP), which became effective in 2001. Institutions must apply this definition of misconduct to federally-funded research to receive funding. While institutions are free to adopt definitions of misconduct that go beyond the federal standard, it is not known how many do. We analyzed misconduct policies from 183 U.S. research institutions and coded them according to thirteen different types of behavior mentioned in the misconduct definition. We also obtained data on the institution's total research funding and public vs. private status, and the year it adopted the definition. We found that more than half (59%) of the institutions in our sample had misconduct policies that went beyond the federal standard. Other than FFP, the most common behaviors included in definitions were "other serious deviations" (45.4%), "significant or material violations of regulations" (23.0%), "misuse of confidential information" (15.8%), "misconduct related to misconduct" (14.8%), "unethical authorship other than plagiarism" (14.2%), "other deception involving data manipulation" (13.1%), and "misappropriation of property/theft" (10.4%). Significantly more definitions adopted in 2001 or later went beyond the federal standard than those adopted before 2001 (73.2% vs. 26.8%), and significantly more definitions adopted by institutions in the lower quartile of total research funding went beyond the federal standard than those adopted by institutions in the upper quartiles. Public vs. private status was not significantly associated with going beyond the federal standard.

Radiation protection code of practice in academic and research institutes

International Nuclear Information System (INIS)

Abdalla, A. A. M.

2010-05-01

The main aim of this study was to establish a code of practice on radiation protection for safe control of radiation sources used in academic and research institutes, another aim of this study was to assess the current situation of radiation protection in some of the academic and research institutes.To achieve the aims of this study, a draft of a code of practice has been developed which is based on international and local relevant recommendation. The developed code includes the following main issues: regulatory responsibilities, radiation protection program and design of radiation installations. The second aim had been accomplished by conducting inspection visits to five (A, B, C, D and E) academic and to four (F, G, H and I ) research institutes. Eight of such institutes are located in Khartoum State and the ninth one is in Madani city (Aljazeera State). The inspection activities have been carried out using a standard inspection check list developed by the regulatory authority of the Sudan. The inspection missions to the above mentioned institutes involved also evaluation of radiation levels around the premises and storage areas of radiation sources. The dose rate measurement around radiation sources locations were found to be quite low. This mainly is due to the fact that the activities of most radionuclides that are used in these institutes are quite low ( in the range of micro curies). Also ,most the x-ray machines that were found in use for scientific academic and research purposes work at low k Vp of maximum 60 k Vp. None of the radiation workers in the inspected institutes has a personal radiation monitoring device, therefor staff dose levels have not been assessed. However it was noted that in most of the academic/ research studies radiation workers are only exposed to very low levels of radiation and for a very short time that dose not exceed 1 minute, therefore the expected occupational exposure of the staff is very low. Radiation measurement in public

Large Aspect Ratio Tokamak Study

International Nuclear Information System (INIS)

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

1980-06-01

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

Recrystallized graphite utilization as the first wall material in Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.; Novokhatsky, A.N.; Petrov, Y.V.; Sakharov, N.V.; Terukov, E.I.; Trapeznikova, I.N.; Denisov, E.A.; Kurdumov, A.A.; Kompaniec, T.N.; Lebedev, V.M.; Litunovstkii, N.V.; Mazul, I.

2007-01-01

Full text of publication follows: Globus-M spherical tokamak, built at A.F. Ioffe Physico-Technical Institute in 1999 is the first Russian spherical tokamak and has the broad area of research in controlled fusion [1]. Besides small aspect ratio (A=1.5) the distinguishing feature of the tokamak is the powerful energy supply system and auxiliary heating, which give opportunity to reach high specific power deposition up to few W/cm 3 . The utmost plasma current density and B/R ratio among spherical tokamaks allow operation in the range of high plasma densities ∼ 10 20 m -3 . This feature results in big power density loads to the first wall due to small plasma-wall spacing. The area of the first wall amour was gradually increased during few years since 2003, and nowadays reaches almost 90% of the inner vessel surface faced to plasma. Plasma facing protecting tiles are manufactured from recrystallized graphite doped by different elements (Ti, Si, B). Additionally the plasma facing surface was protected by films deposited during boronization. The tendency of short time and long time scale plasma parameters variation are discussed including the plasma performance improvement with increase of protected area. Technology of tiles preparation before installation into the tokamak vessel is briefly described, as well as technology of plasma facing armor preparation before the plasma experiments. Few protecting tiles doped by different elements which were exposed to plasma fluxes of dissimilar power densities for a long time were extracted from the vacuum vessel. The analysis of tiles material (RGT-91) to hold (accumulate) deuterium was made. The distribution of absorbed deuterium concentration along poloidal coordinate was measured. The elementary composition of the films deposited on the tiles was studied by Rutherford back scattering technique and by nuclear resonance reaction method. Other modern methods of surface and structural analysis of material exposed to prolonged

Recrystallized graphite utilization as the first wall material in Globus-M spherical tokamak

Energy Technology Data Exchange (ETDEWEB)

Gusev, V.; Novokhatsky, A.N.; Petrov, Y.V.; Sakharov, N.V.; Terukov, E.I.; Trapeznikova, I.N. [A.F. IOFFE Physico-technical Institute, Russian Academy of Sciences, St Petersburg (Russian Federation); Denisov, E.A.; Kurdumov, A.A.; Kompaniec, T.N. [St. Petersburg State Univ., Research Institute of Physics (Russian Federation); Lebedev, V.M. [B.P. Konstantinov Nuclear Physics Institute, Russian Academy of Science, Gatchina (Russian Federation); Litunovstkii, N.V. [D.V. Efremov Institute of Electrophysical Apparatus, St.Petersburg (Russian Federation); Mazul, I. [Development of Plasma Facing Materials and Components Laboratory, EFREMOV INSTITUTE, St Petersbourg (Russian Federation)

2007-07-01

Full text of publication follows: Globus-M spherical tokamak, built at A.F. Ioffe Physico-Technical Institute in 1999 is the first Russian spherical tokamak and has the broad area of research in controlled fusion [1]. Besides small aspect ratio (A=1.5) the distinguishing feature of the tokamak is the powerful energy supply system and auxiliary heating, which give opportunity to reach high specific power deposition up to few W/cm{sup 3}. The utmost plasma current density and B/R ratio among spherical tokamaks allow operation in the range of high plasma densities {approx} 10{sup 20} m{sup -3}. This feature results in big power density loads to the first wall due to small plasma-wall spacing. The area of the first wall amour was gradually increased during few years since 2003, and nowadays reaches almost 90% of the inner vessel surface faced to plasma. Plasma facing protecting tiles are manufactured from recrystallized graphite doped by different elements (Ti, Si, B). Additionally the plasma facing surface was protected by films deposited during boronization. The tendency of short time and long time scale plasma parameters variation are discussed including the plasma performance improvement with increase of protected area. Technology of tiles preparation before installation into the tokamak vessel is briefly described, as well as technology of plasma facing armor preparation before the plasma experiments. Few protecting tiles doped by different elements which were exposed to plasma fluxes of dissimilar power densities for a long time were extracted from the vacuum vessel. The analysis of tiles material (RGT-91) to hold (accumulate) deuterium was made. The distribution of absorbed deuterium concentration along poloidal coordinate was measured. The elementary composition of the films deposited on the tiles was studied by Rutherford back scattering technique and by nuclear resonance reaction method. Other modern methods of surface and structural analysis of material

Research at the Paul Scherrer Institut

International Nuclear Information System (INIS)

Walter, H.K.

1996-01-01

The Paul Scherrer Institut (PSI) is a multidisciplinary research institute for natural sciences and technology. In national and international collaboration with universities, other research institutes and industry, PSI is active in elementary particle physics, life sciences, solid-state physics, material sciences, nuclear and non-nuclear energy research, and energy-related ecology. PSI's priorities lie in research fields which are relevant to sustainable development, serve educational needs and are beyond the possibilities of a single university department. PSI develops and operates complex research installations open of the world's most powerful cyclotron, allowing to operate high intensity secondary pion and muon beams, a neutron spallation source and various applications in medicine and materials research. A short review on research at PSI is presented, with special concentration on particle physics experiments. (author)

Dynamics and feedback control of plasma equilibrium position in a tokamak

International Nuclear Information System (INIS)

Burenko, O.

1983-01-01

A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. 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. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems

Evaluation of applicability of 2D iron core model for two-limb configuration of GOLEM tokamak

Czech Academy of Sciences Publication Activity Database

Markovič, Tomáš; Gryaznevich, M.; Ďuran, Ivan; Svoboda, V.; Vondrášek, G.

2013-01-01

Roč. 88, 6-8 (2013), s. 835-838 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Tokamak * Ferromagnetic core * Integral method * Tokamak GOLEM Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613002573#

Teaching and Research at Undergraduate Institutions

Science.gov (United States)

Garg, Shila

2006-03-01

My own career path has been non-traditional and I ended up at a primarily undergraduate institution by pure accident. However, teaching at a small college has been extremely rewarding to me, since I get to know and interact with my students, have an opportunity to work with them one-on-one and promote their intellectual growth and sense of social responsibility. One of the growing trends at undergraduate institutions in the past decade has been the crucial role of undergraduate research as part of the teaching process and the training of future scientists. There are several liberal arts institutions that expect research-active Faculty who can mentor undergraduate research activities. Often faculty members at these institutions consider their roles as teacher-scholars with no boundary between these two primary activities. A researcher who is in touch with the developments in his/her own field and contributes to new knowledge in the field is likely to be a more exciting teacher in the classroom and share the excitement of discovery with the students. At undergraduate institutions, there is generally very good support available for faculty development projects in both teaching and research. Often, there is a generous research leave program as well. For those who like advising and mentoring undergraduates and a teaching and learning centered paradigm, I will recommend a career at an undergraduate institution. In my presentation, I will talk about how one can prepare for such a career.

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

Development in Diagnostics Application to Control Advanced Tokamak Plasma

International Nuclear Information System (INIS)

Koide, Y.

2008-01-01

For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)

Theory of incremental turbulent transport in tokamaks

International Nuclear Information System (INIS)

Similon, P.L.

1991-01-01

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

Enhancing current density profile control in tokamak experiments using iterative learning control

NARCIS (Netherlands)

Felici, F.A.A.; Oomen, T.A.E.

2015-01-01

Tokamaks are toroidal devices to create and confine high-temperature plasmas, and are presently at the forefront of nuclear fusion research. Many parameters in a tokamak are feedback controlled, but some quantities that are either difficult to measure or difficult to control are still controlled by

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

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.

Module description of TOKAMAK equilibrium code MEUDAS

Energy Technology Data Exchange (ETDEWEB)

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

2002-01-01

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

Institutional Repositories in Indian Universities and Research Institutes: A Study

Science.gov (United States)

Krishnamurthy, M.; Kemparaju, T. D.

2011-01-01

Purpose: The purpose of this paper is to report on a study of the institutional repositories (IRs) in use in Indian universities and research institutes. Design/methodology/approach: Repositories in various institutions in India were accessed and described in a standardised way. Findings: The 20 repositories studied covered collections of diverse…

Annual report of the Japan Atomic Energy Research Institute, for fiscal 1989

International Nuclear Information System (INIS)

1990-01-01

Japan Atomic Energy Research Institute has promoted the research on nuclear safety, the research and development of high temperature engineering and nuclear fusion which are the leading projects bringing about the breakthrough in atomic energy technology, the research on radiation utilization and the research and development of nuclear-powered ships, following the 'Plan of development and long term utilization of atomic energy' decided in 1987, as the central, general research institute in atomic energy field in Japan. Also the advanced basic research for opening atomic energy frontier and various international cooperation as well as the cooperation in Japan have been promoted. The engineering safety of nuclear facilities and environmental safety, the construction of the Nuclear Fuel Cycle Safety Engineering Research Facility, the design of the High Temperature Engineering Test Reactor and the various tests related to it, the reconstruction of JT-60 for increasing the current, the design of a nuclear fusion reactor, the high utilization of radiation using ion beam, the construction of Sekinehama Port for the nuclear-powered ship 'Mutsu', the power increasing test of the reactor of the Mutsu, the reconstruction of JRR-3 and others are reported. (K.I.)

Dust limit management strategy in tokamaks

Science.gov (United States)

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

2009-06-01

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

Dust limit management strategy in tokamaks

International Nuclear Information System (INIS)

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

2009-01-01

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

The design of the KSTAR tokamak

International Nuclear Information System (INIS)

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

1999-01-01

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

Microwave Tokamak Experiment

International Nuclear Information System (INIS)

Anon.

1988-01-01

The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

Financial Support for Institutional Research, 1969-70.

Science.gov (United States)

Pieper, W. C., Jr.

The Association for Institutional Research conducted a survey of all institutions of higher education in the U.S. and Canada in order to assess the number, size, and financial support of institutional research offices. Data were requested for the 1969-70 academic year. This report is based on the responses of 1,444 institutions that returned the…

Analysis of magnetohydrodynamic modes in tokamaks by x-ray techniques

International Nuclear Information System (INIS)

Sauthoff, N.R.

1977-01-01

A brief review of recent studies of fluctuations in x-ray emission from tokamak plasmas of controlled thermonuclear fusion interest is given. The origin of the x-rays, the nature of the oscillations, and measurement and analysis techniques are discussed, with emphasis on the work performed on the ST and PLT tokamaks. Areas for future research, particularly in the region of reconstruction, are stressed

Advanced Tokamak Stability Theory

Science.gov (United States)

Zheng, Linjin

2015-03-01

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

Annual report of the Fusion Research Center for the period of April 1, 1984 to March 31, 1985

International Nuclear Information System (INIS)

1986-01-01

Research and development activities of the Fusion Research Center (Department of Large Tokamak Development and Department of Thermonuclear Fusion Research) from April 1984 to March 1985 are described. The JT-60 program progressed as scheduled. Commissioning of the JT-60 tokamak was completed by the end of the period under review. In parallel with installation and test of the tokamak machine, installation of basic diagnostic instruments and examination of the procedure for experiment had been made to meet the first phase Joule heating experiment. (The first plasma discharge was recorded on April 8, 1985). Construction of auxiliary heating systems had continued. A medium-sized tokamak, JFT-2M, had been operated for high-power ICRF heating and ECH assisted LH current drive experiments. Installation of a power supply for plasma shaping in JFT-2M was completed. In the field of plasma theory, detailed analysis had been made on nonlinear kink/tearing modes in a plasma with free boundary and also on ICRF heating. Development of a high-voltage, high-current He ion source for JT-60 plasma diagnostics had proceeded successfully, and tests of JT-60 LH and ICRF luanchers as well. Surface erosion of a new ceramics, SiC with BeO addition by proton bombardment was studied. In IEA's Large Coil Task, three coil test was made at ORNL. A 11 T experiment of TMC-1, a large-bore Nb 3 Sn coil was completed. Commissioning tests of tritium handling facilities had proceeded in the Tritium Process Laboratory. Design studies of the Fusion Experimental Reactor (FER) and INTOR had been advanced. (author)

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Conditioning of the vacuum chamber of the Tokamak Novillo; Acondicionamiento de la camara de vacio del Tokamak Novillo

Energy Technology Data Exchange (ETDEWEB)

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

1992-03-15

The obtained experimental results of the implementation of two techniques of present time for the conditioning of the internal wall of the chamber of discharges of the Tokamak Novillo are presented, which has been designed, built and put in operation in the Laboratory of Plasma Physics of the National Institute of Nuclear Research (ININ). These techniques are: the vacuum baking and the low energy pulsed discharges, which were applied after having reached an initial pressure of the order of 10{sup -7} Torr. with a system of turbomolecular pumping previous preparation of surfaces and vacuum seals. The analysis of residual gases was carried out with a mass spectrometer before and after conditioning. The obtained results show that the vacuum baking it was of great effectiveness to reduce the value of the initial pressure in short time, in more of a magnitude order and the low energy discharges reduced the oxygen at worthless levels with regard to the initial values. (Author)

Design study of an AC power supply system in JT-60SA

International Nuclear Information System (INIS)

Shimada, Katsuhiro; Baulaigue, Olivier; Cara, Philippe; Coletti, Alberto; Coletti, Roberto; Matsukawa, Makoto; Terakado, Tsunehisa; Yamauchi, Kunihito

2011-01-01

In the initial research phase of JT-60SA, which is the International Thermonuclear Experimental Reactor (ITER) satellite Tokamak with superconducting toroidal and poloidal magnetic field coils, the plasma heating operation of 30 MW-60 s or 20 MW-100 s is planned for 5.5 MA single null divertor plasmas. To achieve this operation, AC power source of the medium voltage of 18 kV and ∼7 GJ has to be provided in total to the poloidal field coil power supplies and additional heating devices such as neutral beam injection (NBI) and electron cyclotron radio frequency (ECRF). In this paper, the proposed AC power supply system in JT-60SA was estimated from the view point of available power, and harmonic currents based on the standard plasma operation scenario during the initial research phase. This AC power supply system consists of the reused JT-60 power supply facilities including motor generators with flywheel, AC breakers, harmonic filters, etc., to make it cost effective. In addition, the conceptual design of the upgraded AC power supply system for the ultimate heating power of 41 MW-100 s in the extended research phase is also described.

A Proposed Framework of Institutional Research Development Phases

Science.gov (United States)

Bosch, Anita; Taylor, John

2011-01-01

Globally, research has become a key driver for the achievement of status and the procurement of funding for higher education institutions. Although there is mounting pressure on institutions to become research active, many institutions are rooted in a strong tradition of teaching. These institutions find it challenging to develop research capacity…

Homogeneous group, research, institution

Directory of Open Access Journals (Sweden)

Francesca Natascia Vasta

2014-09-01

Full Text Available The work outlines the complex connection among empiric research, therapeutic programs and host institution. It is considered the current research state in Italy. Italian research field is analyzed and critic data are outlined: lack of results regarding both the therapeutic processes and the effectiveness of eating disorders group analytic treatment. The work investigates on an eating disorders homogeneous group, led into an eating disorder outpatient service. First we present the methodological steps the research is based on including the strong connection among theory and clinical tools. Secondly clinical tools are described and the results commented. Finally, our results suggest the necessity of validating some more specifical hypothesis: verifying the relationship between clinical improvement (sense of exclusion and painful emotions reduction and specific group therapeutic processes; verifying the relationship between depressive feelings, relapses and transition trough a more differentiated groupal field.Keywords: Homogeneous group; Eating disorders; Institutional field; Therapeutic outcome

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.

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

Czech Academy of Sciences Publication Activity Database

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

2012-01-01

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

Tokamak control simulator

International Nuclear Information System (INIS)

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

1976-01-01

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

Relation of vertical stability and aspect ratio in tokamaks

International Nuclear Information System (INIS)

Stambaugh, R.D.; Lao, L.L.; Lazarus, E.A.

1992-01-01

It is evaluated how the upper limit to plasma elongation κ, caused by vertical stability, varies with the aspect ratio A=R/a of the tokamak. Equilibria were generated with EFITD and the vertical stability was assessed by GATO. For a 'generic' tokamak with a superconducting wall conformal to the plasma shape and a distance 0.5 a away from the plasma edge and a constant current profile (q 0 =1.0, l i ≅1.0, q 95 =3.2) it is found that the maximum stable κ decreased only slowly from 2.65 at A=2.0 to 2.4 at A=6.0. To first order, a reasonable assumption in trade-off studies of new machine designs is no dependence of κ max on A. (author). Letter-to-the-editor. 13 refs, 3 figs, 1 tab

Development of the pellet injector for JT-60

International Nuclear Information System (INIS)

Kawasaki, Kouzo; Hiratsuka, Hajimo; Takatsu, Hideyuki; Shimizu, Masatsugu; Onozuka, Masanori; Uchikawa, Takashi; Iwamoto, Syuichi; Hashiri, Nobuo

1989-01-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proved that the device provides high speed hydrogen pellets just as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 1.6 km/sec at 50 bar propellant gas. The device is now in use for JT-60 contributing to plasma study. In this paper the outline of features and performance of the device is presented. (author). 4 refs.; 8 figs

Lithium beam diagnostic system on the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Anda, G.; Bencze, A.; Berta, Miklós; Dunai, D.; Háček, Pavel; Krbec, Jaroslav; Réfy, D.; Krizsanóczi, T.; Bató, S.; Ilkei, T.; Kiss, I.G.; Veres, G.; Zoletnik, S.

2016-01-01

Roč. 108, October (2016), s. 1-6 ISSN 0920-3796 R&D Projects: GA MŠk(CZ) LM2011021 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : BES * Plasma diagnostics * COMPASS tokamak Plasma density profile Plasma current fluctuations * Plasma density profile * Plasma current fluctuations Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379616303131

Overview of the TCV tokamak program: scientific progress and facility upgrades.

Czech Academy of Sciences Publication Activity Database

Coda, S.; Ficker, Ondřej; Horáček, Jan; Papřok, Richard

2017-01-01

Roč. 57, October (2017), č. článku 102011. ISSN 0029-5515 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : TCV * tokamak * overview Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016

Work and safety managements for on-site installation, commissioning, tests by EU of quench protection circuits for JT-60SA

International Nuclear Information System (INIS)

Yamauchi, Kunihito; Okano, Jun; Shimada, Katsuhiro; Ohmori, Yoshikazu; Terakado, Tsunehisa; Matsukawa, Makoto; Koide, Yoshihiko; Kobayashi, Kazuhiro; Ikeda, Yoshitaka; Fukumoto, Masahiro; Kushita, Kouhei N.

2016-03-01

The superconducting Satellite Tokamak machine “JT-60SA” under construction in Naka Fusion Institute is an international collaborative project between Japan Atomic Energy Agency (JAEA) as the Implementing Agency (IA) of Japan (JA) and Fusion for Energy (F4E) as the IA of Europe (EU). The contributions for this project are based on the supply of components, and thus European manufacturer shall conduct the installation, commissioning and tests on Naka site under the general supervision by F4E via the designated institute in each EU nation. This means that JAEA had an issue to manage the works by European workers and their safety although there is no direct contract. This report describes the approaches for the work and safety managements, which were agreed with EU after the negotiation, and the completed on-site works for Quench Protection Circuits (QPC) as the first experience for EU in JT-60SA project. (author)

Institutional failures and transaction costs of Bulgarian private research institutes

OpenAIRE

Nozharov, Shteryo

2016-01-01

The paper analyses the reasons for poor performance of private research institutes in Bulgaria. In this regard the Institutional Economics methods are used. A connection between smart growth policy goals and Bulgarian membership in EU is made. The gaps in the institutional environment are identified as well as measures for their elimination are proposed. The main accent of the study is put on the identification of transaction costs, arisen as a result of the failures of the institutional envi...

Tokamak building-design considerations for a large tokamak device

International Nuclear Information System (INIS)

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

1981-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-09-01

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

Perspectives from the Aldo Leopold Wilderness Research Institute: The Wildland Research institute

Science.gov (United States)

J. M. Bowker; H. Ken Cordell; Neelam C. Poudyal

2014-01-01

The Wildland Research Institute (WRi) at the University of Leeds (UK) came into being in October 2009. Its origins go back to a United Kingdom research councilfunded seminar series called Wilderness Britain? which ran between 1998 and 2000 and was coordinated from the University of Leeds. This opened up the wider debate on wilderness and rewilding in the UK and later...

Prediction of density limits in tokamaks: Theory, comparison with experiment, and application to the proposed Fusion Ignition Research Experiment

International Nuclear Information System (INIS)

Stacey, Weston M.

2002-01-01

A framework for the predictive calculation of density limits in future tokamaks is proposed. Theoretical models for different density limit phenomena are summarized, and the requirements for additional models are identified. These theoretical density limit models have been incorporated into a relatively simple, but phenomenologically comprehensive, integrated numerical calculation of the core, edge, and divertor plasmas and of the recycling neutrals, in order to obtain plasma parameters needed for the evaluation of the theoretical models. A comparison of these theoretical predictions with observed density limits in current experiments is summarized. A model for the calculation of edge pedestal parameters, which is needed in order to apply the density limit predictions to future tokamaks, is summarized. An application to predict the proximity to density limits and the edge pedestal parameters of the proposed Fusion Ignition Research Experiment is described

Flow shear stabilization of hybrid electron-ion drift mode in tokamaks

International Nuclear Information System (INIS)

Bai, L.

1999-01-01

In this paper, a model of sheared flow stabilization on hybrid electron-ion drift mode is proposed. At first, in the presence of dissipative trapped electrons, there exists an intrinsic oscillation mode in tokamak plasmas, namely hybrid dissipative trapped electron-ion temperature gradient mode (hereafter, called as hybrid electron-ion drift mode). This conclusion is in agreement with the observations in the simulated tokamak experiment on the CLM. Then, it is found that the coupling between the sheared flows and dissipative trapped electrons is proposed as the stabilization mechanism of both toroidal sheared flow and poloidal sheared flow on the hybrid electron-ion drift mode, that is, similar to the stabilizing effect of poloidal sheared flow on edge plasmas in tokamaks, in the presence of both dissipative trapped electrons and toroidal sheared flow, large toroidal sheared flow is always a strong stabilizing effect on the hybrid electron-ion drift mode in internal transport barrier location, too. This result is consistent with the experimental observations in JT-60U. (author)

Flow shear stabilization of hybrid electron-ion drift mode in tokamaks

International Nuclear Information System (INIS)

Bai, L.

2001-01-01

In this paper, a model of sheared flow stabilization on hybrid electron-ion drift mode is proposed. At first, in the presence of dissipative trapped electrons, there exists an intrinsic oscillation mode in tokamak plasmas, namely hybrid dissipative trapped electron-ion temperature gradient mode (hereafter, called as hybrid electron-ion drift mode). This conclusion is in agreement with the observations in the simulated tokamak experiment on the CLM. Then, it is found that the coupling between the sheared flows and dissipative trapped electrons is proposed as the stabilization mechanism of both toroidal sheared flow and poloidal sheared flow on the hybrid electron-ion drift mode, that is, similar to the stabilizing effect of poloidal sheared flow on edge plasmas in tokamaks, in the presence of both dissipative trapped electrons and toroidal sheared flow, large toroidal sheared flow is always a strong stabilizing effect on the hybrid electron-ion drift mode in internal transport barrier location, too. This result is consistent with the experimental observations in JT-60U. (author)

25 years TNO Road-Vehicles Research Institute

NARCIS (Netherlands)

1995-01-01

Since the founding of the TNO Road-Vehicles Research Institute 25years ago, the institute has managed to develop a leading position in automotive research in several disciplines. A steady growth of the institute during the first 20 years has turned into a strong growth during the last 5 years. A

Advanced commercial tokamak study

International Nuclear Information System (INIS)

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

1985-12-01

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

D-D tokamak reactor assessment

International Nuclear Information System (INIS)

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

1983-01-01

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

Advanced tokamak research in DIII-D

International Nuclear Information System (INIS)

Greenfield, C M; Murakami, M; Ferron, J R

2004-01-01

Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with β N ≤ 3.5 and β T ≤ 3.6% sustained for up to 1 s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by

Stability and heating of a poloidal divertor tokamak

Energy Technology Data Exchange (ETDEWEB)

Biddle, A. P.; Dexter, R. N.; Holly, D. T.; Lipschultz, B.; Osborne, T. H.; Prager, S. C.; Shepard, D.A., Sprott, J.C.; Witherspoon, F. D.

1980-06-01

Five experimental studies - two stability and three heating investigations - have been carried out on Tokapole II, a Tokamak with a four node poloidal divertor. First, discharges have been attained with safety factor q as low as 0.6 over most of the column without degradation of confinement, and correlation of helical instability onset with current profile shape is being studied. Second, the axisymmetric instability has been investigated in detail for various noncircular cross-sectional shapes, and results have been compared with a numerical stability code adapted to the Tokapole machine. Third, application of high power fast wave ion cyclotron resonance heating doubles the ion temperature and permits observation of heating as a function of harmonic number and spatial location of the resonance. Fourth, low power shear Alfven wave propagation is underway to test the applicability of this heating method to tokamaks. Fifth, preionization by electron cyclotron heating has been employed to reduce the startup loop voltage by approx. 60%.

Progress in application of high temperature superconductor in tokamak magnets

Czech Academy of Sciences Publication Activity Database

Gryaznevich, M.; Svoboda, V.; Stöckel, Jan; Sykes, A.; Sykes, N.; Kingham, D.; Hammond, G.; Apte, P.; Todd, T.N.; Ball, S.; Chappell, S.; Melhem, D.; Ďuran, Ivan; Kovařík, Karel; Grover, O.; Markovič, T.; Odstrčil, M.; Odstrčil, T.; Šindlery, A.; Vondrášek, G.; Kocman, J.; Lilley, M.K.; de Grouchy, P.; Kim, H.-T.

2013-01-01

Roč. 88, 9-10 (2013), s. 1593-1596 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] Institutional support: RVO:61389021 Keywords : tokamaks * HTS * magnet s Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001117#

Dust remobilization experiments on the COMPASS tokamak.

Czech Academy of Sciences Publication Activity Database

Weinzettl, Vladimír; Matějíček, Jiří; Ratynskaia, S.; Tolias, P.; De Angeli, M.; Riva, G.; Dimitrova, Miglena; Havlíček, Josef; Adámek, Jiří; Seidl, Jakub; Tomeš, Matěj; Cavalier, Jordan; Imríšek, Martin; Havránek, Aleš; Pánek, Radomír; Peterka, Matěj

2017-01-01

Roč. 124, November (2017), s. 446-449 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA ČR(CZ) GA14-12837S; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : Dust remobilization * Tungsten * Disruption * ELM * Plasma * Tokamak Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617300650

Implementation of rapid imaging system on the COMPASS tokamak.

Czech Academy of Sciences Publication Activity Database

Havránek, Aleš; Weinzettl, Vladimír; Fridrich, David; Cavalier, Jordan; Urban, Jakub; Komm, Michael

2017-01-01

Roč. 123, November (2017), s. 857-860 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : Camera * Data acquisition * Video processing * Tokamak Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S092037961730354X

Joint Global Change Research Institute (JGCRI)

Data.gov (United States)

Federal Laboratory Consortium — The Joint Global Change Research Institute (JGCRI) is dedicated to understanding the problems of global climate change and their potential solutions. The Institute...

Multi-Institution Research Centers: Planning and Management Challenges

Science.gov (United States)

Spooner, Catherine; Lavey, Lisa; Mukuka, Chilandu; Eames-Brown, Rosslyn

2016-01-01

Funding multi-institution centers of research excellence (CREs) has become a common means of supporting collaborative partnerships to address specific research topics. However, there is little guidance for those planning or managing a multi-institution CRE, which faces specific challenges not faced by single-institution research centers. We…

Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

Energy Technology Data Exchange (ETDEWEB)

Lampert, M. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); BME NTI, Budapest (Hungary); Anda, G.; Réfy, D.; Zoletnik, S. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); Czopf, A.; Erdei, G. [Department of Atomic Physics, BME IOP, Budapest (Hungary); Guszejnov, D.; Kovácsik, Á.; Pokol, G. I. [BME NTI, Budapest (Hungary); Nam, Y. U. [National Fusion Research Institute, Daejeon (Korea, Republic of)

2015-07-15

A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.

Status of plasma physics research activities in Egypt

International Nuclear Information System (INIS)

Masoud, M.M.

1997-01-01

The status of plasma physics research activities in Egypt is reviewed. There are nine institutes with plasma research activities. The largest is the Atomic energy Authority (AEA), which has activities in fundamental plasma studies, fusion technology, plasma and laser applications, and plasma simulation. The experiments include Theta Pinches, a Z Pinch, a coaxial discharge, a glow discharge, a CO 2 laser, and the EGYPTOR tokamak. (author)

Tore-Supra: a Tokamak with superconducting toroidal field coils

International Nuclear Information System (INIS)

Turck, B.

1987-07-01

Tore Supra is a tokamak under construction on the site of Cen Cadarache by the Euratom-CEA Association. The machine technology integrates all problems related to the fabrication and the operation of large superconducting coils and of the associated cryogenic system. Tore Supra will provide a significant experience to prepare the next generation of machines for plasma physics and controlled fusion. Tore Supra is specially designed to implement a large physics program. The superconducting coils make possible the study of plasma confinement in long pulses (more than 60s), the impurities and the stability, and the efficiency of additional heating sources (neutral particle beams and radio frequency heating). The opportunity is taken to recall the particular features and requirements of the superconducting coils of the large future tokamaks in order to point out the problems that have to be faced by any new material (superconducting or not)

Theoretical and experimental studies of runaway electrons in the TEXTOR tokamak

International Nuclear Information System (INIS)

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

2016-01-01

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

Theoretical and experimental studies of runaway electrons in the TEXTOR tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-01

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

Modular pulse sequencing in a tokamak system

International Nuclear Information System (INIS)

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

1992-01-01

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

The Joint Institute for Nuclear Research in Experimental Physics of Elementary Particles

Science.gov (United States)

Bednyakov, V. A.; Russakovich, N. A.

2018-05-01

The year 2016 marks the 60th anniversary of the Joint Institute for Nuclear Research (JINR) in Dubna, an international intergovernmental organization for basic research in the fields of elementary particles, atomic nuclei, and condensed matter. Highly productive advances over this long road clearly show that the international basis and diversity of research guarantees successful development (and maintenance) of fundamental science. This is especially important for experimental research. In this review, the most significant achievements are briefly described with an attempt to look into the future (seven to ten years ahead) and show the role of JINR in solution of highly important problems in elementary particle physics, which is a fundamental field of modern natural sciences. This glimpse of the future is full of justified optimism.

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.)

Heating of plasmas in tokamaks by current-driven turbulence

International Nuclear Information System (INIS)

Kluiver, H. de.

1985-10-01

Investigations of current-driven turbulence have shown the potential to heat plasmas to elevated temperatures in relatively small cross-section devices. The fundamental processes are rather well understood theoretically. Even as it is shown to be possible to relax the technical requirements on the necessary electric field and the pulse length to acceptable values, the effect of energy generation near the plasma edge, the energy transport, the impurity influx and the variation of the current profile are still unknown for present-day large-radius tokamaks. Heating of plasmas by quasi-stationary weakly turbulent states caused by moderate increases of the resistivity due to higher loop voltages could be envisaged. Power supplies able to furnish power levels 5-10 times higher than the usual values could be used for a demonstration of those regimes. At several institutes and university laboratories the study of turbulent heating in larger tokamaks and stellarators is pursued

ELM induced tungsten melting and its impact on tokamak operation

Czech Academy of Sciences Publication Activity Database

Coenen, J.W.; Arnoux, G.; Bazylev, B.; Matthews, G. F.; Jachmich, S.; Balboa, I.; Clever, M.; Dejarnac, Renaud; Coffey, I.; Corre, Y.; Devaux, S.; Frassinetti, L.; Gauthier, E.; Horáček, Jan; Knaup, M.; Komm, Michael; Krieger, K.; Marsen, S.; Meigs, A.; Mertens, Ph.; Pitts, R.A.; Puetterich, T.; Rack, M.; Stamp, M.; Sergienko, G.; Tamain, P.; Thompson, V.

2015-01-01

Roč. 463, August (2015), s. 78-84 ISSN 0022-3115. [PLASMA-SURFACE INTERACTIONS 21: International Conference on Plasma-Surface Interactions in Controlled Fusion Devices. Kanazawa, 26.05.2014-30.05.2014] Institutional support: RVO:61389021 Keywords : plasma * tokamak Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015 http://www.sciencedirect.com/science/article/pii/S0022311514005960#

Software development of the KSTAR Tokamak Monitoring System

International Nuclear Information System (INIS)

Kim, K.H.; Lee, T.G.; Baek, S.; Lee, S.I.; Chu, Y.; Kim, Y.O.; Kim, J.S.; Park, M.K.; Oh, Y.K.

2008-01-01

The Korea Superconducting Tokamak Advanced Research (KSTAR) project, which is constructing a superconducting Tokamak, was launched in 1996. Much progress in instrumentation and control has been made since then and the construction phase will be finished in August 2007. The Tokamak Monitoring System (TMS) measures the temperatures of the superconducting magnets, bus-lines, and structures and hence monitors the superconducting conditions during the operation of the KSTAR Tokamak. The TMS also measures the strains and displacements on the structures in order to monitor the mechanical safety. There are around 400 temperature sensors, more than 240 strain gauges, 10 displacement gauges and 10 Hall sensors. The TMS utilizes Cernox sensors for low temperature measurement and each sensor has its own characteristic curve. In addition, the TMS needs to perform complex arithmetic operations to convert the measurements into temperatures for each Cernox sensor for this large number of monitoring channels. A special software development effort was required to reduce the temperature conversion time and multi-threading to achieve the higher performance needed to handle the large number of channels. We have developed the TMS with PXI hardware and with EPICS software. We will describe the details of the implementations in this paper

The OpenAIRE Guide for Research Institutions

Directory of Open Access Journals (Sweden)

Gültekin Gürdal

2013-11-01

Full Text Available This text is transcript of OpenAIRE Guide which is prepared in order to help research institutions was released on 13.04.2011and translated with the cooperation of ANKOS Open Access and Institutional Repositories Grup members and OpenAIREplus project team of Turkey which is coordinated from Izmir Institute of Technology Library. OpenAIRE Project aims to support researchers in complying with the European Commission Seventh Framework Programme Open Access Pilot through a European Helpdesk System; support researchers in depositing their research publications in an institutional or disciplinary repository; build up an OpenAIRE portal and e-infrastructure for repository networks. The project will work in tadem with OpeanAIREplus Project which has the principal goal of creating a robust, participatory service for the cross-linking of peer-reviewed scientific publications and associated datasets.

Preliminary Design of Alborz Tokamak

Science.gov (United States)

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

2012-04-01

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

Auditing as Institutional Research: A Qualitative Focus.

Science.gov (United States)

Fetterman, David M.

1991-01-01

Internal institutional auditing can improve effectiveness and efficiency and protect an institution's assets. Many of the concepts and techniques used to analyze higher education institutions are qualitative in nature and suited to institutional research, including fiscal, operational, data-processing, investigative, management consulting,…

Moving Divertor Plates in a Tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Zhang, H.

2009-01-01

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

Moving Divertor Plates in a Tokamak

Energy Technology Data Exchange (ETDEWEB)

S.J. Zweben, H. Zhang

2009-02-12

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

Radial electric field at the plasma edge on the FT-2 Tokamak in regimes with large gradients

International Nuclear Information System (INIS)

Lashkul, S.; Popov, A.

2001-01-01

The transport barrier formation is widely believed to be the fundamental element of transition into improved confinement regimes (H-mode). Experiments on many tokamaks demonstrate that transport barrier formation is connected with the suppression of turbulent transport by shear of E x B drift. Therefore, the calculation of radial electric field is of great importance. Our work is devoted to progress the neoclassical theory by taking into account electron viscosity and non-linear effects (ion inertia), presented results being valuable for interpretation transition into H-mode at the plasma edge in small tokamaks. Calculations of the electric field profile for FT-2 tokamak (a=8cm, R 0 =55cm, Ioffe Institute, Russia) according found expressions are in the good agreement with experimental results obtained. (orig.)

Upgrade of the COMPASS tokamak real-time control system

Czech Academy of Sciences Publication Activity Database

Janky, Filip; Havlíček, Josef; Batista, A.J.N.; Kudláček, Ondřej; Seidl, Jakub; Neto, A.C.; Pipek, Jan; Hron, Martin; Mikulín, Ondřej; Duarte, A.S.; Carvalho, B.B.; Stöckel, Jan; Pánek, Radomír

2014-01-01

Roč. 89, č. 3 (2014), s. 186-194 ISSN 0920-3796 R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Real-time * Feedback control * Real-time framework * MARTe * COMPASS tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.152, year: 2014 http://www.sciencedirect.com/science/article/pii/S0920379613007564

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Conceptual design of Remote Control System for EAST tokamak

International Nuclear Information System (INIS)

Sun, X.Y.; Wang, F.; Wang, Y.; Li, S.

2014-01-01

Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication

Conceptual design of Remote Control System for EAST tokamak

Energy Technology Data Exchange (ETDEWEB)

Sun, X.Y., E-mail: xysun@ipp.ac.cn; Wang, F.; Wang, Y.; Li, S.

2014-05-15

Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.

Nuclear fusion research and plasma application technologies in SWIP (Southwestern Institute of Physics)

International Nuclear Information System (INIS)

Deng, X.W.

1990-01-01

A brief introduction of nuclear fusion research and plasma application technologies in SWIP is reported in this paper. The SWIP focuses its fusion efforts mainly on Tokamak with mirror as the supplemental experiments and fusion reactor conceptual design as preparation for future application of fusion energy. SWIP is making great efforts on fusion technology spin-off to make contribution towards national economic construction. (Author)

National Nuclear Research Institute Annual Report 2013

International Nuclear Information System (INIS)

2014-01-01

The report highlights the activities of the National Nuclear Research Institute (NNRI) of the Ghana Atomic Energy Commission for the year 2013, grouped under the following headings: Centres under the institute namely Nuclear Reactors Research Centre (NRRC); Accelerator Research Centre (ARC); Engineering Services Centre (ESC); National Radioactive Waste Management Centre (NRWMC); Nuclear Chemistry and Environmental Research Centre (NCERC); Nuclear Applications Centre (NAC) and National Data Centre (NDC). (A. B.)

[Managing a health research institute: towards research excellence through continuous improvement].

Science.gov (United States)

Olmedo, Carmen; Buño, Ismael; Plá, Rosa; Lomba, Irene; Bardinet, Thierry; Bañares, Rafael

2015-01-01

Health research institutes are a strategic commitment considered the ideal environment to develop excellence in translational research. Achieving quality research requires not only a powerful scientific and research structure but also the quality and integrity of management systems that support it. The essential instruments in our institution were solid strategic planning integrated into and consistent with the system of quality management, systematic evaluation through periodic indicators, measurement of key user satisfaction and internal audits, and implementation of an innovative information management tool. The implemented management tools have provided a strategic thrust to our institute while ensuring a level of quality and efficiency in the development and management of research that allows progress towards excellence in biomedical research. Copyright © 2015 SESPAS. Published by Elsevier Espana. All rights reserved.

Scrape-off layer power flux measurements in the Tore Supra tokamak

Czech Academy of Sciences Publication Activity Database

Gunn, J. P.; Dejarnac, Renaud; Devynck, P.; Fedorczak, N.; Fuchs, Vladimír; Gil, C.; Kočan, M.; Komm, Michael; Kubič, M.; Lunt, T.; Monier-Garbet, P.; Pascal, J.-Y.; Saint-Laurent, F.

2013-01-01

Roč. 438, suppl (2013), S184-S188 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/20./. Aachen, 21.05.2012-25.05.2012] Institutional support: RVO:61389021 Keywords : plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.016, year: 2013 http://www.sciencedirect.com/science/article/pii/S0022311513000639#

Final Report: Performance Engineering Research Institute

Energy Technology Data Exchange (ETDEWEB)

Mellor-Crummey, John [Rice Univ., Houston, TX (United States)

2014-10-27

This document is a final report about the work performed for cooperative agreement DE-FC02-06ER25764, the Rice University effort of Performance Engineering Research Institute (PERI). PERI was an Enabling Technologies Institute of the Scientific Discovery through Advanced Computing (SciDAC-2) program supported by the Department of Energy's Office of Science Advanced Scientific Computing Research (ASCR) program. The PERI effort at Rice University focused on (1) research and development of tools for measurement and analysis of application program performance, and (2) engagement with SciDAC-2 application teams.

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.

A tokamak with nearly uniform coil stress based on virial theorem

International Nuclear Information System (INIS)

Tsutsui, H.

2002-01-01

A novel tokamak concept with a new type of toroidal field (TF) coils and a central solenoid (CS) whose stress is much reduced to a theoretical limit determined by the virial theorem has been devised. Recently, we had developed a tokamak with force-balanced coils (FBCs) which are multi-pole helical hybrid coils combining TF coils and a CS coil. The combination reduces the net electromagnetic force in the direction of major radius. In this work, we have extended the FBC concept using the virial theorem. High-field coils should accordingly have same averaged principal stresses in all directions, whereas conventional FBC reduces stress in the toroidal direction only. Using a shell model, we have obtained the poloidal rotation number of helical coils which satisfy the uniform stress condition, and named the coil as virial-limited coil (VLC). VLC with circular cross section of aspect ratio A=2 reduces maximum stress to 60% compared with that of TF coils. In order to prove the advantage of VLC concept, we have designed a small VLC tokamak Todoroki-II. The plasma discharge in Todoroki-II will be presented. (author)

Evaluation of the state water-resources research institutes

Science.gov (United States)

Ertel, M.O.

1988-01-01

Water resources research institutes, as authorized by the Water Resources Research Act of 1984 (Public Law 98-242), are located in each state and in the District of Columbia, Guam, Puerto Rico , and the Virgin Islands. Public Law 98-242 mandated an onsite evaluation of each of these institutes to determine whether ' . . .the quality and relevance of its water resources research and its effectiveness as an institution for planning, conducting, and arranging for research warrant its continued support in the national interest. ' The results of these evaluations, which were conducted between September 1985 and June 1987, are summarized. The evaluation teams found that all 54 institutes are meeting the basic objectives of the authorizing legislation in that they: (1) use the grant funds to support research that addresses water problems of state and regional concern; (2) provide opportunities for training of water scientists through student involvement on research projects; and (3) promote the application of research results through preparation of technical reports and contributions to the technical literature. The differences among institutes relate primarily to degrees of effectiveness, and most often are determined by the financial, political, and geographical contexts in which the institutes function and by the quality of their leadership. (Lantz-PTT)

Radiant Research. Institute for Energy Technology 1948-98

International Nuclear Information System (INIS)

Njoelstad, Olav

1999-01-01

Institutt for Atomenergi (IFA), or Institute for Atomic Energy, at Kjeller, Norway, was founded in 1948. The history of the institute as given in this book was published in 1999 on the occasion of the institute's 50th anniversary. The scope of the institute was to do research and development as a foundation for peaceful application of nuclear energy and radioactive substances in Norway. The book tells the story of how Norway in 1951 became the first country after the four superpowers and Canada to have its own research reactor. After the completion of the reactor, the institute experienced a long and successful period and became the biggest scientific and technological research institute in Norway. Three more reactors were built, one in Halden and two at Kjeller. Plans were developed to build nuclear powered ships and nuclear power stations. It became clear, however, in the 1970s, that there was no longer political support for nuclear power in Norway, and it was necessary for the institute to change its research profile. In 1980, the institute changed its name to Institutt for energiteknikk (IFE), or Institute for energy technology, to signal the broadened scope. The book describes this painful but successful readjustment and shows how IFE in the 1980s and 1990s succeeded in using its special competence from the nuclear field to establish special competence in new research fields with great commercial potential

The ARIES-I high-field-tokamak reactor: Design-point determination and parametric studies

International Nuclear Information System (INIS)

Miller, R.L.

1989-01-01

The multi-institutional ARIES study has examined the physics, technology, safety, and economic issues associated with the conceptual design of a tokamak magnetic-fusion reactor. The ARIES-I variant envisions a DT-fueled device based on advanced superconducting coil, blanket, and power-conversion technologies and a modest extrapolation of existing tokamak physics. A comprehensive systems and trade study has been conducted as an integral and ongoing part of the reactor assessment in order to identify an acceptable design point to be subjected to detailed analysis and integration as well as to characterize the ARIES-I operating space. Results of parametric studies leading to the identification of such a design point are presented. 15 refs., 6 figs., 2 tabs

Recent developments in the role of atomic processes in future tokamaks

International Nuclear Information System (INIS)

Post, D.

1996-01-01

Since the beginning of magnetic fusion research, reducing the impurity level in experiments has been strongly correlated with successful achievement of high performance plasmas. One of the most important examples of this was the recognition that the use of tungsten as a plasma facing material and the associated high radiative losses were responsible for the poor performance of the ORMAK and PLT tokamaks. Tungsten was replaced with graphite and the central plasma temperature in PLT increased a factor of ten. The magnetic fusion program is now planning on constructing an ignited fusion experiment. One of the major design issues is the reduction of the peak heat loads on the plasma facing components. It appears that the carefully controlled introduction of impurities can lead to a solution of the problem. copyright 1996 American Institute of Physics

Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

Science.gov (United States)

Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

2013-10-01

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

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

International Nuclear Information System (INIS)

Bathke, C.G.

1994-01-01

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

76 FR 11765 - Education Research and Special Education Research Grant Programs; Institute of Education Sciences...

Science.gov (United States)

2011-03-03

... DEPARTMENT OF EDUCATION Education Research and Special Education Research Grant Programs; Institute of Education Sciences; Overview Information; Education Research and Special Education Research.... SUMMARY: The Director of the Institute of Education Sciences (Institute) announces the Institute's FY 2012...

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

Institutional radioactive waste management in the Nuclear Research Institute Rez plc

International Nuclear Information System (INIS)

Kovarik, P.; Svoboda, K.; Podlaha, J.

2008-01-01

Nuclear research institute Rez, plc. (mentioned below as NRI) has had a dominant position in the area of the nuclear research and development in the Czech Republic, the Central and the Eastern Europe. Naturally, the radioactive waste management is an integral part of the nuclear industry, research and development. For that reason, there is Centre of the radioactive waste management (mentioned below as Centre) in the NRI. This Centre is engaged in the radioactive waste treatment, decontamination, characterisation, decommissioning and other relevant activities. This paper describes the system of technology and other information about institutional radioactive waste management in the NRI. (authors)

START: the creation of a spherical tokamak

International Nuclear Information System (INIS)

Sykes, Alan

1992-01-01

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

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

1992-01-01

The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

Electrical conductivity in tokamaks and extended neoclassical theory

International Nuclear Information System (INIS)

Segre, S.E.; Zanza, V.

1992-01-01

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

Grazing incidence EUV study of the Alcator tokamaks

International Nuclear Information System (INIS)

Castracane, J.

1982-01-01

The use of impurity radiation to examine plasma conditions is a well known technique. To gain access, however, to the hot, central portion of the plasma created in the present confinement machines it is necessary to be able to observe radiation from medium and heavy elements such as molybdenum and iron. These impurities radiate primarily in the extreme ultra violet region of the spectrum and can play a role in the power balance of the tokamak. Radiation from highly ionized molybdenum was examined on the Alcator A and C tokamaks using a photometrically calibrated one meter grazing incidence monochromator. On Alcator A, a pseudo-continuum of Mo emissions in the 60 to 100 A ranges were seen to comprise 17% of the radiative losses from the plasma. This value closely matched measurements by a broad band bolometer array. Following these preliminary measurements, the monochromator was transferred to Alcator C for a more thorough examination of EUV emissions. Deviations from predicted scaling laws for energy confinement time vs density were observed on this machine

Modelling dust transport in tokamaks

International Nuclear Information System (INIS)

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

2008-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Compact tokamak reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

Mathematical modeling plasma transport in tokamaks

Energy Technology Data Exchange (ETDEWEB)

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

1997-01-01

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

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%

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

Danish Space Research Institute

International Nuclear Information System (INIS)

1991-01-01

The present report presents a description of the activities and finances of the Danish Space Reserach Institute during 1989 and 1990. The research deals with infrared astronomy (ISOPHOT), X-ray astronomy (EXPECT/SODART), hard X-ray astronomy (WATCH), satellite projects and sounding rocket experiments. (CLS)

ITER tokamak device

International Nuclear Information System (INIS)

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

1991-01-01

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

Overview of research potential of Institute for Nuclear Research

International Nuclear Information System (INIS)

Ciocanescu, Marin

2007-01-01

The main organizations involved in nuclear power production in Romania, under supervision of Presidency, Prime Minister and Parliament are: CNCAN (National Commission for Nuclear Activities Control), Nuclear Agency, Ministry of Economy and Commerce, ANDRAD (Waste Management Agency), SNN (Nuclearelectrica National Society), RAAN (Romanian Authority for Nuclear Activities), ICN (Institute for Nuclear Research - Pitesti), SITON (Center of Design and Engineering for Nuclear Projects- Bucharest); ROMAG-PROD (Heavy Water Plant), CNE-PROD (Cernavoda Nuclear Power Plant - Production Unit), CNE-INVEST (Cernavoda Nuclear Power Plant -Investments Unit), FCN (Nuclear Fuel Factory). The Institute for Nuclear Research, Pitesti INR, Institute for Nuclear Research, Pitesti is endowed with a TRIGA Reactor, Hot Cells, Materials Laboratories, Nuclear Fuel, Nuclear Safety Laboratories, Nuclear Fuel, Nuclear Safety. Waste management. Other research centers and laboratories implied in nuclear activities are: ICIT, National Institute for cryogenics and isotope technologies at Rm Valcea Valcea. with R and D activity devoted to heavy water technologies, IFIN, Institute for nuclear physics and engineering, Bucharest, as well as the educational institutions involved in atomic energy applications and University research, Politechnical University Bucharest, University of Bucharest, University of Pitesti, etc. The INR activity outlined, i.e. the nuclear power research as a scientific and technical support for the Romanian nuclear power programme, mainly dedicated to the existing NPP in the country (CANDU). Focused with priority are: - Nuclear Safety (behavior of plant materials, components, installations during accident conditions and integrity investigations); - Radioactive Waste Management Radioactive; - Radioprotection; Product and services supply for NPP. INR Staff numbers 320 R and D qualified and experienced staff, 240 personnel in devices and prototype workshops and site support

Booklet of the Research Institute of Clinical Medicine

International Nuclear Information System (INIS)

Todua, F.; Jgamadze, N.; Todua, N.; Beriashvili, Z.; Chelishvili, M.; Todua, I.; Chovelidze, Sh. et al.

2012-01-01

Research Institute of Clinical Medicine is one of the biggest university diagnostic and treatment centre in Georgia with unique modern diagnostic and treatment apparatus. The institute is acknowledged as a leader in various trends of radiology and surgery. The Research Institute of Clinical Medicine was founded in 1991. It is the leading scientific establishment in the field of medicine. The scientific-research work of the Institute is coordinated by the National Academy of Sciences of Georgia. The main scientific trend of the Institute is the Early Complex Diagnostics and Treatment. The scientific activity of the Institute is led by the Scientific Council. Institute achieved remarkable success since its foundation: It has been defended 56 theses for Candidate of Medical Sciences and 16 for Doctor of Medical Sciences; About 30 post-graduate students and more than 200 radiologists have taken training courses in radiology. Nowadays they work in different regions of Georgia, 21 inventions took out patents. It has been published 2000 scientific works and 9 monographs. (authors)

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

International Nuclear Information System (INIS)

Fitzpatrick, R.

1998-01-01

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

Carbon deposition and hydrogen retention in tokamak

International Nuclear Information System (INIS)

Tanabe, Tetsuo

2006-01-01

The results of measurements on co-deposition of hydrogen isotopes and wall materials, hydrogen retention, redeposition of carbon and deposition of hydrogen on PMI of JT-60U are described. From above results, selection of plasma facing material and ability of carbon wall is discussed. Selection of plasma facing materials in fusion reactor, characteristics of carbon materials as the plasma facing materials, erosion, transport and deposition of carbon impurity, deposition of tritium in JET, results of PMI in JT-60, application of carbon materials to PFM of ITER, and future problems are stated. Tritium co-deposition in ITER, erosion and transport of carbon in tokamak, distribution of tritium deposition on graphite tile used as bumper limiter of TFTR, and measurement results of deposition of tritium on the Mark-IIA divertor tile and comparison between them are described. (S.Y.)

Predictions of fast wave heating, current drive, and current drive antenna arrays for advanced tokamaks

International Nuclear Information System (INIS)

Batchelor, D.B.; Baity, F.W.; Carter, M.D.

1994-01-01

The objective of the advanced tokamak program is to optimize plasma performance leading to a compact tokamak reactor through active, steady state control of the current profile using non-inductive current drive and profile control. To achieve these objectives requires compatibility and flexibility in the use of available heating and current drive systems--ion cyclotron radio frequency (ICRF), neutral beams, and lower hybrid. For any advanced tokamak, the following are important challenges to effective use of fast waves in various roles of direct electron heating, minority ion heating, and current drive: (1) to employ the heating and current drive systems to give self-consistent pressure and current profiles leading to the desired advanced tokamak operating modes; (2) to minimize absorption of the fast waves by parasitic resonances, which limit current drive; (3) to optimize and control the spectrum of fast waves launched by the antenna array for the required mix of simultaneous heating and current drive. The authors have addressed these issues using theoretical and computational tools developed at a number of institutions by benchmarking the computations against available experimental data and applying them to the specific case of TPX

Institutional shared resources and translational cancer research

Directory of Open Access Journals (Sweden)

De Paoli Paolo

2009-06-01

Full Text Available Abstract The development and maintenance of adequate shared infrastructures is considered a major goal for academic centers promoting translational research programs. Among infrastructures favoring translational research, centralized facilities characterized by shared, multidisciplinary use of expensive laboratory instrumentation, or by complex computer hardware and software and/or by high professional skills are necessary to maintain or improve institutional scientific competitiveness. The success or failure of a shared resource program also depends on the choice of appropriate institutional policies and requires an effective institutional governance regarding decisions on staffing, existence and composition of advisory committees, policies and of defined mechanisms of reporting, budgeting and financial support of each resource. Shared Resources represent a widely diffused model to sustain cancer research; in fact, web sites from an impressive number of research Institutes and Universities in the U.S. contain pages dedicated to the SR that have been established in each Center, making a complete view of the situation impossible. However, a nation-wide overview of how Cancer Centers develop SR programs is available on the web site for NCI-designated Cancer Centers in the U.S., while in Europe, information is available for individual Cancer centers. This article will briefly summarize the institutional policies, the organizational needs, the characteristics, scientific aims, and future developments of SRs necessary to develop effective translational research programs in oncology. In fact, the physical build-up of SRs per se is not sufficient for the successful translation of biomedical research. Appropriate policies to improve the academic culture in collaboration, the availability of educational programs for translational investigators, the existence of administrative facilitations for translational research and an efficient organization

Institutional shared resources and translational cancer research.

Science.gov (United States)

De Paoli, Paolo

2009-06-29

The development and maintenance of adequate shared infrastructures is considered a major goal for academic centers promoting translational research programs. Among infrastructures favoring translational research, centralized facilities characterized by shared, multidisciplinary use of expensive laboratory instrumentation, or by complex computer hardware and software and/or by high professional skills are necessary to maintain or improve institutional scientific competitiveness. The success or failure of a shared resource program also depends on the choice of appropriate institutional policies and requires an effective institutional governance regarding decisions on staffing, existence and composition of advisory committees, policies and of defined mechanisms of reporting, budgeting and financial support of each resource. Shared Resources represent a widely diffused model to sustain cancer research; in fact, web sites from an impressive number of research Institutes and Universities in the U.S. contain pages dedicated to the SR that have been established in each Center, making a complete view of the situation impossible. However, a nation-wide overview of how Cancer Centers develop SR programs is available on the web site for NCI-designated Cancer Centers in the U.S., while in Europe, information is available for individual Cancer centers. This article will briefly summarize the institutional policies, the organizational needs, the characteristics, scientific aims, and future developments of SRs necessary to develop effective translational research programs in oncology.In fact, the physical build-up of SRs per se is not sufficient for the successful translation of biomedical research. Appropriate policies to improve the academic culture in collaboration, the availability of educational programs for translational investigators, the existence of administrative facilitations for translational research and an efficient organization supporting clinical trial recruitment

Southern Universities Nuclear Institute

International Nuclear Information System (INIS)

Anon.

1977-01-01

The Southern Universities Nuclear Institute was created in 1961 to provide postgraduate research and teaching facilities for the universities of Cape Town and Stellenbosch. The main research tool is the 6,0 MV Van de Graaff accelerator installed in 1964. Developments and improvements over the years have maintained the Institute's research effectiveness. The work of local research groups has led to a large number of M Sc and doctorate degrees and numerous publications in international journals. Research at the Institute includes front-line studies of basic nuclear and atomic physics, the development and application of nuclear analytical techniques and the application of radioisotope tracers to problems in science, industry and medicine. The Institute receives financial support from the two southern universities, the Department of National Education, the CSIR and the Atomic Energy Board

Microwave Tokamak Experiment: Overview and status

International Nuclear Information System (INIS)

1990-05-01

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

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-01-01

This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

Magnetic ''islandography'' in tokamaks

International Nuclear Information System (INIS)

Callen, J.D.; Waddell, B.V.; Hicks, H.R.

1978-09-01

Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

Fabrication of the vacuum vessel for JT-60 machine upgrade

International Nuclear Information System (INIS)

Uchikawa, T.; Takanabe, K.; Tsujimura, S.; Ue, K.; Oka, K.; Kuri, S.; Ioki, K.; Namiki, K.; Suzuki, Y.; Horliike, H.; Ninomiya, H.; Yamamoto, M.; Neyatani, Y.; Ando, T.; Matsukawa, M.

1992-01-01

The JT-60 tokamak was upgraded to double the plasma current to 6 MA. In the JT-60 machine upgrade (JT-60U), the vacuum vessel and poloidal field (PF) coils were renewed. The new vacuum vessel features a three-dimensionally curved, thin double-skin torus with multi-arc D-shaped cross section. The double-skin structure is strengthened with square pipes placed in between the outer and inner skins. Fabrication and site installation of the vessel was smoothly completed in February, 1991. This paper describes an overview of the JT-60U vacuum vessel construction

Large power supply facilities for fusion research

International Nuclear Information System (INIS)

Miyahara, Akira; Yamamoto, Mitsuyoshi.

1976-01-01

The authors had opportunities to manufacture and to operate two power supply facilities, that is, 125MVA computer controlled AC generator with a fly wheel for JIPP-T-2 stellerator in Institute of Plasma Physics, Nagoya University and 3MW trial superconductive homopolar DC generator to the Japan Society for Promotion of Machine Industry. The 125MVA fly-wheel generator can feed both 60MW (6kV x 10kA) DC power for toroidal coils and 20MW (0.5kV x 40kA) DC power for helical coils. The characteristic features are possibility of Bung-Bung control based on Pontrjagin's maximum principle, constant current control or constant voltage control for load coils, and cpu control for routine operation. The 3MW (150V-20000A) homopolar generator is the largest in the world as superconductive one, however, this capacity is not enough for nuclear fusion research. The problems of power supply facilities for large Tokamak devices are discussed

Annual report of Naka Fusion Research Establishment. From April 1, 1996 to March 31, 1997

Energy Technology Data Exchange (ETDEWEB)

Shimizu, Masatsugu; Ide, Shunsuke; Matsukawa, Makoto; Kurihara, Ryoichi; Koizumi, Koichi; Takahashi, Ichiro [eds.] [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1997-10-01

This report provides an overview of research and development activities at Naka Fusion Research Establishment, JAERI, during the period from April 1, 1996 to March 31, 1997. The activities in Naka Fusion Research Establishment are highlighted by high temperature plasma research in JT-60 and JFT-2M, and progress in ITER-EDA, including technology development. The objectives of the JT-60 project are to contribute to the ITER physics R and D and to establish the physics basis for a steady state tokamak fusion reactor like SSTR. Objectives of the JFT-2M program are (1) advanced and basic researches for the development of high-performance plasmas for nuclear fusion and (2) contribution to the physics R and D for ITER, with a merit of flexibility of a medium-size device. The Detailed Design Report (DDR) of ITER was issued by the Director in November 1996, as the basis of the Final Design Report (FDR). After the formal review by the Technical Advisory Committee (TAC), the DDR was officially accepted by the ITER Council at its 11th Meeting held in December 1996. The DDR is composed of various technical documents on the detailed design of plasma parameters, tokamak components, plant system and tokamak building. The major results of safety analyses described in the Non-site Specific Safety Report (NSSR)-1 was also included in the DDR. The FDR will be prepared by the end of 1997 for presentation at the ITER Council. (J.P.N.)

Central Institute for Nuclear Research (1956 - 1979)

International Nuclear Information System (INIS)

Flach, G.; Bonitz, M.

1979-12-01

The Central Institute for Nuclear Research (ZfK) of the Academy of Sciences of the GDR is presented. This first overall survey covers the development of the ZfK since 1956, the main research activities and results, a description of the departments responsible for the complex implementation of nuclear research, the social services for staff and the activities of different organizations in the largest central institute of the Academy of Sciences of the GDR. (author)

Institutional Research's Role in Strategic Planning

Science.gov (United States)

Voorhees, Richard A.

2008-01-01

Institutions that have organized and centralized their data enjoy an obvious advantage in grappling with strategic planning and other issues. As the drumbeat for accountability, planning, and demonstrating effectiveness to internal and external stakeholders intensifies, the stature and importance of institutional research offices on most campuses…

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

International Nuclear Information System (INIS)

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

1986-05-01

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

Two-body similarity and its violation in tokamak edge plasmas

International Nuclear Information System (INIS)

Catto, P.J.; Knoll, D.A.; Krasheninnikov, S.I.

1996-01-01

Scaling laws found under the assumption that two-body collisions dominate can be effectively used to benchmark complex multi-dimensional codes dedicated to investigating tokamak edge plasmas. The applicability of such scaling laws to the interpretation of experimental data, however, is found to be restricted to the relatively low plasma densities ( 19 m -3 ) at which multistep processes, which break the two-body collision approximation, are unimportant. copyright 1996 American Institute of Physics

Retooling Institutional Support Infrastructure for Clinical Research

Science.gov (United States)

Snyder, Denise C.; Brouwer, Rebecca N.; Ennis, Cory L.; Spangler, Lindsey L.; Ainsworth, Terry L.; Budinger, Susan; Mullen, Catherine; Hawley, Jeffrey; Uhlenbrauck, Gina; Stacy, Mark

2016-01-01

Clinical research activities at academic medical centers are challenging to oversee. Without effective research administration, a continually evolving set of regulatory and institutional requirements can detract investigator and study team attention away from a focus on scientific gain, study conduct, and patient safety. However, even when the need for research administration is recognized, there can be struggles over what form it should take. Central research administration may be viewed negatively, with individual groups preferring to maintain autonomy over processes. Conversely, a proliferation of individualized approaches across an institution can create inefficiencies or invite risk. This article describes experiences establishing a unified research support office at the Duke University School of Medicine based on a framework of customer support. The Duke Office of Clinical Research was formed in 2012 with a vision that research administration at academic medical centers should help clinical investigators navigate the complex research environment and operationalize research ideas. The office provides an array of services that have received high satisfaction ratings. The authors describe the ongoing culture change necessary for success of the unified research support office. Lessons learned from implementation of the Duke Office of Clinical Research may serve as a model for other institutions undergoing a transition to unified research support. PMID:27125563

Current drive by spheromak injection into a tokamak

International Nuclear Information System (INIS)

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

1990-01-01

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

78 FR 2678 - Proposed Collection; Comment Request (60-Day FRN): The National Cancer Institute (NCI...

Science.gov (United States)

2013-01-14

... Request (60-Day FRN): The National Cancer Institute (NCI) SmokefreeTXT (Text Message) Program Evaluation..., Behavioral Scientist/ Health Science Administrator, Division of Cancer Control and Population Sciences, 6130... text message smoking cessation intervention designed for young adult smokers ages 18-29. The Smokefree...

Čínský česnek nechci....ale čínské cívky pro tokamak ano

Czech Academy of Sciences Publication Activity Database

Řípa, Milan

2014-01-01

Roč. 7, prosinec (2014), s. 14-15 Institutional support: RVO:61389021 Keywords : fusion * superconducting tokamak * ITER * EAST Subject RIV: BL - Plasma and Gas Discharge Physics http://3pol.cz/1563-cinsky-cesnek-nechci

Alligator Rivers Region Research Institute: annual research summary 1989-1990

International Nuclear Information System (INIS)

1991-01-01

The Alligator Rivers Region Research Institute (ARRRI) research activities are associated with an assessment of environmental effect of mining in the region. While emphasis on baseline research is now much reduced, some projects are still necessary because of significant changes in the Magela Creek system, because new areas of proposed mining have been identified (e.g. Coronation Hill) and because the emphasis now being placed on rehabilitation research requires a sound knowledge of the Region's flora. The ARRRI rehabilitation research program has concentrated on the Ranger mine site, principally because it is at a critical planning stage where detailed research information is required. With regard to the development of techniques, research at the Institute has led to the development of specific analytical methods or protocols that can be used in assessing environmental impact. 39 tabs., 42 figs

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

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

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

1985-01-01

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

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

[Hospital biomedical research through the satisfaction of a Health Research Institute professionals].

Science.gov (United States)

Olmedo, C; Plá, R; Bellón, J M; Bardinet, T; Buño, I; Bañares, R

2015-01-01

A Health Research Institute is a powerful strategic commitment to promote biomedical research in hospitals. To assess user satisfaction is an essential quality requirement. The aim of this study is to evaluate the professional satisfaction in a Health Research Institute, a hospital biomedical research centre par excellence. Observational study was conducted using a satisfaction questionnaire on Health Research Institute researchers. The explored dimensions were derived from the services offered by the Institute to researchers, and are structured around 4 axes of a five-year Strategic Plan. A descriptive and analytical study was performed depending on adjustment variables. Internal consistency was also calculated. The questionnaire was completed by 108 researchers (15% response). The most valued strategic aspect was the structuring Areas and Research Groups and political communication and dissemination. The overall rating was 7.25 out of 10. Suggestions for improvement refer to the need for help in recruitment, and research infrastructures. High internal consistency was found in the questionnaire (Cronbach alpha of 0.9). So far research policies in health and biomedical environment have not been sufficiently evaluated by professionals in our field. Systematic evaluations of satisfaction and expectations of key stakeholders is an essential tool for analysis, participation in continuous improvement and advancing excellence in health research. Copyright © 2015 SECA. Published by Elsevier Espana. All rights reserved.

The role of the spherical tokamak in clarifying tokamak physics

International Nuclear Information System (INIS)

Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.

1999-01-01

The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)

World's largest DC flywheel generator for the toroidal field power supply of JAERI's JFT-2M Tokamak nuclear fusion reactor

International Nuclear Information System (INIS)

Tani, Takashi; Nakanishi, Yuji; Horita, Tsuyoshi; Kawase, Chiharu; Oyabu, Isao; Kishimoto, Takeshi.

1996-01-01

Mitsubishi Electric has delivered the world's largest DC generator for the toroidal field coil power supply of the JFT-2M Tokamak at the Japan Atomic Energy Research Institute. The unit rotates at 225 or 460 rpm, providing a maximum rated output of 2,700 V, 19,000 A and 51.3 MW. The toroidal field is a DC field, so use of a DC generator permits a simpler design consuming less floor space than an AC drive system. The generator was manufactured following extensive studies on commutation, mechanical strength and insulation. (author)

Conditioning of the vacuum chamber of the Tokamak Novillo

International Nuclear Information System (INIS)

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

1992-03-01

The obtained experimental results of the implementation of two techniques of present time for the conditioning of the internal wall of the chamber of discharges of the Tokamak Novillo are presented, which has been designed, built and put in operation in the Laboratory of Plasma Physics of the National Institute of Nuclear Research (ININ). These techniques are: the vacuum baking and the low energy pulsed discharges, which were applied after having reached an initial pressure of the order of 10 -7 Torr. with a system of turbomolecular pumping previous preparation of surfaces and vacuum seals. The analysis of residual gases was carried out with a mass spectrometer before and after conditioning. The obtained results show that the vacuum baking it was of great effectiveness to reduce the value of the initial pressure in short time, in more of a magnitude order and the low energy discharges reduced the oxygen at worthless levels with regard to the initial values. (Author)

Models for Predicting Boundary Conditions in L-Mode Tokamak Plasma

International Nuclear Information System (INIS)

Siriwitpreecha, A.; Onjun, T.; Suwanna, S.; Poolyarat, N.; Picha, R.

2009-07-01

Full text: The models for predicting temperature and density of ions and electrons at boundary conditions in L-mode tokamak plasma are developed using an empirical approach and optimized against the experimental data obtained from the latest public version of the International Pedestal Database (version 3.2). It is assumed that the temperature and density at boundary of L-mode plasma are functions of engineering parameters such as plasma current, toroidal magnetic field, total heating power, line averaged density, hydrogenic particle mass (A H ), major radius, minor radius, and elongation at the separatrix. Multiple regression analysis is carried out for these parameters with 86 data points in L-mode from Aug (61) and JT60U (25). The RMSE of temperature and density at boundary of L-mode plasma are found to be 24.41% and 18.81%, respectively. These boundary models are implemented in BALDUR code, which will be used to simulate the L-mode plasma in the tokamak

Diagnostics and control for the steady state and pulsed tokamak DEMO

Czech Academy of Sciences Publication Activity Database

Orsitto, F.P.; Villari, R.; Moro, F.; Todd, T.N.; Lilley, S.; Jenkins, I.; Felton, R.; Biel, W.; Silva, A.; Scholz, M.; Rzadkiewicz, J.; Ďuran, Ivan; Tardocchi, M.; Gorini, G.; Morlock, C.; Federici, G.; Litnovsky, A.

2016-01-01

Roč. 56, č. 2 (2016), č. článku 026009. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : measurement systems, fusion reactor, fusion plasma diagnostics * fusion reactor * fusion plasma diagnostics * DEMO * Hall sensors * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/2/026009

Academic Libraries’ Role in Improving Institutions Research Impact

KAUST Repository

Tamarkin, Molly

2015-11-11

In the changing landscape of scientific research and scholarly communication, importance of “quality in research”, “reviewed research” and “reviewed publications” in qualifying for the ratings and rankings are widely discussed. While publishing the research pieces in peer-reviewed and highly ranked journals are increasingly important, there are different methods and tools to be in place at Institutional level to increase researchers’ profile and the ranking of the institutions. As a young research based university created in 2009, King Abdullah University of Science and Technology (KAUST) focuses on the bibliometrics and altemetrics tools, author affiliations, author naming and plug-ins to different search engines, research evaluation systems as well as to research repositories. The University has launched an institutional repository in September 2012 as a home for the intellectual outputs of KAUST researchers, and then adopted the first institutional open access mandate in the Arab region effective June 31, 2014. Integration with ORCID became a key element in this process and the best way to ensure data quality for researcher’s scientific contributions systematically. We will present the inclusion and creation of ORCID identifiers in the existing systems as an institutional member to ORCID, and the creation of dedicated integration tools with Current Research Information System (CRIS) as a standardized common resource to monitor KAUST research outputs. We will also present our experiences in awareness programs, trainings, outreach, implementation of systems and tools like PlumX, as well as our approach in improving the research impact and profiling our Institution’s research to the world.

Academic Libraries’ Role in Improving Institutions Research Impact

KAUST Repository

Tamarkin, Molly; Vijayakumar, J.K.; Baessa, Mohamed A.; Grenz, Daryl M.

2015-01-01

In the changing landscape of scientific research and scholarly communication, importance of “quality in research”, “reviewed research” and “reviewed publications” in qualifying for the ratings and rankings are widely discussed. While publishing the research pieces in peer-reviewed and highly ranked journals are increasingly important, there are different methods and tools to be in place at Institutional level to increase researchers’ profile and the ranking of the institutions. As a young research based university created in 2009, King Abdullah University of Science and Technology (KAUST) focuses on the bibliometrics and altemetrics tools, author affiliations, author naming and plug-ins to different search engines, research evaluation systems as well as to research repositories. The University has launched an institutional repository in September 2012 as a home for the intellectual outputs of KAUST researchers, and then adopted the first institutional open access mandate in the Arab region effective June 31, 2014. Integration with ORCID became a key element in this process and the best way to ensure data quality for researcher’s scientific contributions systematically. We will present the inclusion and creation of ORCID identifiers in the existing systems as an institutional member to ORCID, and the creation of dedicated integration tools with Current Research Information System (CRIS) as a standardized common resource to monitor KAUST research outputs. We will also present our experiences in awareness programs, trainings, outreach, implementation of systems and tools like PlumX, as well as our approach in improving the research impact and profiling our Institution’s research to the world.

A study on the fusion reactor - Development of x-ray spectrometer for diagnosis of tokamak plasma

Energy Technology Data Exchange (ETDEWEB)

Jang, Hong Young; Choi, Duk In; Seo, Sung Hun; Kwon, Gi Chung; Jun, Sang Jin; Heo, Sung Hoi; Lee, Chan Hui [Korea Advanced Institute of Science and Technolgoy, Taejon (Korea, Republic of)

1996-09-01

This report of research is on the development of X-ray Photo-Electron Spectrometer (PES) for diagnosis of tokamak plasma. The spectrometer utilizes the fact that the energy of photo-electron is given by the difference between the energy of X-ray and the binding energy of materials. In the research of this year, we constructed two spectrometers; one is operated in KAIST tokamak and the other in KT1 tokamak. In addition, we reviewed the characteristics of the x-ray filter, the photo-electric effect of carbon foils and the detection efficiency of MCP and x-ray radiation of plasma. We measured the x-ray radiation in tokamak and diagnosed the qualitative plasma parameters from the analysis of data. The major interesting plasma parameters, which we can diagnose with the spectrometer, are the electron temperature, Z{sub eff}, the spatial distribution of x-ray radiation and etc. 27 refs., 2 tabs., 20 figs. (author)

Institutional research and development, FY 1987

International Nuclear Information System (INIS)

Struble, G.L.; Lawler, G.M.; Crawford, R.B.; Kirvel, R.D.; Peck, T.M.; Prono, J.K.; Strack, B.S.

1987-01-01

The Institutional Research and Development program at Lawrence Livermore National Laboratory fosters exploratory work to advance science and technology, disciplinary research to develop innovative solutions to problems in various scientific fields, and long-term interdisciplinary research in support of defense and energy missions. This annual report describes research funded under this program for FY87

Information Science Research Institute. Quarterly progress report

Energy Technology Data Exchange (ETDEWEB)

Nartker, T.A.

1994-06-30

This is a second quarter 1194 progress report on the UNLV Information Science Research Institute. Included is symposium activity; staff activity; document analysis program; text retrieval program; institute activity; and goals.

Fusion research in Hungary

International Nuclear Information System (INIS)

Zoletnik, S.

2004-01-01

Hungarian fusion research started in the 1970s, when the idea of installing a small tokamak experiment emerged. In return to computer equipment a soviet tokamak was indeed sent to Hungary and started to operate as MT-1 at the Central Research Institute for Physics (KFKI) in 1979. Major research topics included diagnostic development, edge plasma studies and investigation of disruptions. Following a major upgrade in 1992 (new vacuum vessel, active position control and PC network based data acquisition system) the MT-1M tokamak was used for the study of transport processes with trace impurity injection, micropellet ablation studies, X-ray tomography and laser blow-off diagnostic development. Although funding ceased in the middle of the 90's the group was held alive by collaborations with EU fusion labs: FZ -Juelich, IPP-Garching and CRPP-EPFL Lausanne. In 1998 the machine was dismantled due to reorganization of the Hungarian Academy of Sciences. New horizons opened to fusion research from 1999, when Hungary joined EURATOM and a fusion Association was formed. Since then fusion physics studies are done in collaboration with major EU fusion laboratories, Hungarian researchers also play an active role in JET diagnostics upgrade and ITER design. Major topics are pellet ablation studies, plasma turbulence diagnosis using Beam Emission Spectroscopy and other techniques, tomography and plasma diagnostics using various neutral beams. In fusion relevant technology R and D Hungary has less records. Before joining EURATOM some materials irradiation studies were done at the Budapest Research Reactor at KFKI-AEKI. The present day fusion technology programme focuses still on irradiation studies, nuclear material database and electromagnetic testing techniques. Increasing the fusion technology research activities is a difficult task, as the competition in Hungarian industry is very strong and the interest of organizations in long-term investments into R and D is rather weak and

Predictions of of fast wave heating, current drive, and current drive antenna arrays for advanced tokamaks

International Nuclear Information System (INIS)

Batchelor, D.B.; Baity, F.W.; Carter, M.D.

1995-01-01

The objective of the advanced tokamak program is to optimize plasma performance leading to a compact tokamak reactor through active, steady state control of the current profile using non-inductive current drive and profile control. To achieve this objective requires compatibility and flexibility in the use of available heating and current drive systems - ion cyclotron radio frequency (ICRF), neutral beams, and lower hybrid. For any advanced tokamak, the following are important challenges to effective use of fast waves in various role of direct electron heating, minority ion heating, and current drive: (1) to employ the heating and current drive systems to give self-consistent pressure and current profiles leading to the desired advanced tokamak operating modes; (2) to minimize absorption of the fast waves by parasitic resonances, which limit current drive; (3) to optimize and control the spectrum of fast waves launched by the antenna array for the required mix of simultaneous heating and current drive. The paper addresses these issues using theoretical and computational tools developed at a number of institutions by benchmarking the computations against available experimental data and applying them to the specific case of TPX. (author). 6 refs, 3 figs

Heavy Neutral Beam Probe for edge plasma analysis in tokamaks

International Nuclear Information System (INIS)

1991-01-01

The Heavy Neutral Beam Probe project presented in this document is part of an international collaboration in magnetic confinement fusion energy research sponsored by the US Department of Energy, Office of Energy Research (Confinement Systems Division) and the Centre Canadian de Fusion Magnetique. The overall objective of the effort is to apply a neutral particle beam to the study of edge plasma dynamics in discharges on the Tokamak de Varennes facility in Montreal, Canada. To achieve this goal, a research and development project was started in December, 1990 to produce the necessary hardware to make such measurements and meet the scheduling requirements of the program. At present, satisfactory progress has been achieved. The ion gun is fully operational with the neutralizer in the final assembly stage in preparation for testing. The beam diagnostics have been completed and mounted in the computer automated test stand. The analyzer design and detailed trajectory calculations are nearing completion to allow for the vacuum interface construction. The CAMAC based data acquisition system hardware was integrated into the test stand. Part of this hardware is a component of the Tokamak de Varennes' contribution to the collaboration. Next steps on the critical path include the beginning of the neutralization tests and the start of the analyzer construction. Anticipated installation of the diagnostic on the tokamak is Spring 1992

Microwave Tokamak Experiment: An overview of the construction and checkout phase

International Nuclear Information System (INIS)

Lang, L.L.; Bell, H.H.

1989-01-01

At Lawrence Livermore National Laboratory (LLNL) we constructed and presently operate the Microwave Tokamak Experiment (MTX) to demonstrate the feasibility of using microwave pulses produced from a free electron laser (FEL) to provide electron cyclotron heating (ECH) for use in tokamaks, particularly high-field machines. The MTX consists primarily of the ALCATOR C tokamak and power supplies that were documented and disassembled at the Massachusetts Institute of Technology (MIT) and shipped to LLNL in April 1987. We made many additions, including a new primary power system from the magnetic Fusion Test Facility (MFTF) substation, a new commutation system, substantially upgraded seismic support system for earthquake loading, a fast controls system for use with the FEL, a new data-acquisition system, and a new vault facility. We checked out these systems and put them into operation in October 1988; we achieved the first plasma in November 1988. We have also constructed and installed the microwave transmission system and the local microwave system to be used with the FEL. These systems transmit the microwaves to MTX quasi-optically through an evacuated tube. The ongoing plasma operations, both with and without FEL heating, are described in a companion paper. 12 refs., 2 figs., 2 tabs

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)

Computational studies of tokamak plasmas

International Nuclear Information System (INIS)

Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

1981-02-01

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

Report of results of joint research using facilities in Japan Atomic Energy Research Institute in fiscal year 1987

International Nuclear Information System (INIS)

1988-06-01

The total themes of the joint research in fiscal year 1987 were 127. These are shown being classified into the general joint research in Tokai and Takasaki, neutron diffraction research and cooperative research. The general joint research is the standard utilization form using research reactors JRR-2 and JRR-4, Co-60 gamma irradiation facilities in Tokai and Takasaki, an electron beam irradiation facility in Takasaki, an electron beam linear accelator and hot laboratories, which are opened for common utilization by Japan Atomic Energy Research Institute. The cooperative research is carried out by concluding research cooperation contracts between the researchers of universities and JAERI. In the general joint research, radioactivation analysis, radiation chemistry, irradiation effect, neutron diffraction and so on are the main themes, and in the cooperative research, reactor technology, reactor materials, nuclear physics measurement and others are the main themes. The total number of visitors was 2629 man-day, and decreased due to the stop of JRR-2. Also other activities are reported. The abstracts of respective reports are collected in this book. (Kako, I.)

Combined confinement system applied to tokamaks

International Nuclear Information System (INIS)

Ohkawa, Tihiro

1986-01-01

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

Software problems in magnetic fusion research

International Nuclear Information System (INIS)

Gruber, R.

1982-01-01

The main world effort in magnetic fusion research involves studying the plasma in a Tokamak device. Four large Tokamaks are under construction (TFTR in USA, JET in Europe, T15 in USSR and JT60 in Japan). To understand the physical phenomena that occur in these costly devices, it is generally necessary to carry out extensive numerical calculations. These computer simulations make use of sophisticated numerical methods and demand high power computers. As a consequence they represent a substantial investment. To reduce software costs, the computer codes are more and more often exhanged among scientists. Standardization (STANDARD FORTRAN, OLYMPUS system) and good documentation (CPC program library) are proposed to make codes exportable. Centralized computing centers would also help in the exchange of codes and ease communication between the staff at different laboratories. (orig.)

Institutional research and development, FY 1987

Energy Technology Data Exchange (ETDEWEB)

Struble, G.L.; Lawler, G.M.; Crawford, R.B.; Kirvel, R.D.; Peck, T.M.; Prono, J.K.; Strack, B.S. (eds.)

1987-01-01

The Institutional Research and Development program at Lawrence Livermore National Laboratory fosters exploratory work to advance science and technology, disciplinary research to develop innovative solutions to problems in various scientific fields, and long-term interdisciplinary research in support of defense and energy missions. This annual report describes research funded under this program for FY87. (DWL)

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2004

International Nuclear Information System (INIS)

Weiss, F.P.; Rindelhardt, U.

2005-01-01

The Institute of Safety Research (ISR) is one of the six Research Institutes of Forschungszentrum Rossendorf e.V. (FZR e.V.) which is a member institution of the Wissenschaftsgemeinschaft Gottfried Wilhelm Leibniz (Leibniz Association). Together with the Institute of Radiochemistry, ISR constitutes the research programme ''Safety and Environment'' which is one from three scientific programmes of FZR. In the framework of this research programme, the institute is responsible for the two subprogrammes ''Plant and Reactor Safety'' and ''Thermal Fluid Dynamics'', respectively. We also provide minor contributions to the sub-programme ''Radio-Ecology''. Moreover, with the development of a pulsed photo-neutron source at the radiation source ELBE (Electron linear accelerator for beams of high brilliance and low emittance), we are involved in a networking project carried out by the FZR Institute of Nuclear and Hadron Physics, the Physics Department of TU Dresden, and ISR. (orig.)

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#

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

Marketing based on knowledge as a basis for strategy of research institution – on the example of the Packaging Research Institute

Directory of Open Access Journals (Sweden)

Stanislaw Tkaczyk

2013-09-01

Full Text Available Basis for marketing activities of COBRO – Packaging Research Institute are two main issues. First of all, as a small research and development centre, COBRO has no funds to carry out specialized marketing department. On the other hand, due to huge growth of packaging market, all needs of stakeholders – companies but also other research institutions seeking consortium members – cannot be entirely identified or forecasted, and practical solutions are created in the course of cooperation. For all that reasons Institute has developed its own concept of the knowledge-based marketing, which means more flexible use of the potential of academics and research employees.

THE CONTRIBUTION OF RESEARCH INSTITUTES IN EUREKA PROJECTS

NARCIS (Netherlands)

VANROSSUM, W; CABO, PG

1995-01-01

Technological cooperation between industrial firms and research institutes is studied at the project level. The various forms of cooperation, and the instances in which they are advantageous, are discussed. The authors then focus on situations in which the research institute acts as 'knowledge

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)

The density limit in Tokamaks

International Nuclear Information System (INIS)