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Sample records for intor tokamak

  1. TFTR/JET INTOR workshop on plasma transport tokamaks

    International Nuclear Information System (INIS)

    Singer, C.E.

    1985-01-01

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

  2. INTOR-a European view

    International Nuclear Information System (INIS)

    Grieger, G.

    1985-01-01

    This paper discusses INTOR, an international cooperative project which started in 1979. The project demonstrates Tokamak reactor physics and it serves as a test bed for the development of fusion reactor technology. Particular discussion is devoted to the INTOR physics data base and on the strategy of their generation. It is concluded that INTOR offers a chance to go to the next step within a reasonable time during which the data base will be improved. The INTOR design concept is shown to be very useful as a reference system for assessing the impact of innovative concepts

  3. Engineering features of the INTOR conceptual design

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1981-01-01

    The INTOR engineering design has been strongly influenced by considerations for assembly and maintenance. A maintenance philosophy was established at the outset of the conceptual design to insure that the tokamak configuration would be developed to accommodate maintenance requirements. The main features of the INTOR design are summarized in this paper with primary emphasis on the impact of maintenance considerations

  4. INTOR cost approximation

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1980-01-01

    A simplified cost approximation for INTOR parameter sets in a narrow parameter range is shown. Plausible constraints permit the evaluation of the consequences of parameter variations on overall cost. (orig.) [de

  5. Japanese contributions to IAEA INTOR Workshop, Phase Two A, Part 3

    International Nuclear Information System (INIS)

    Mori, Shigeru; Tomabechi, Ken; Fujisawa, Noboru

    1988-02-01

    This report corresponds to Chapter I and II of Japanese contribution report to IAEA INTOR Workshop, Phase Two A, Part 3. The major objectives of the INTOR Workshop Part 3 are to evaluate innovative ideas which would significantly improve the tokamak concept, to study new-selected critical technical issues, that affect the feasibility or practicability of the INTOR concept, to assess scientific and technical data bases, and to conduct critical analyses of the existing INTOR-like designs with the aim of preparing a useful information base for future design work for ITER. To perform the innovations, critical issues, and data base, the following six groups are organized; (A) Impurity control, (B) Operational limits and confinement, (C) Current drive and heating, (D) Electromagnetics, (E) Configuration and maintenance, (F) Blanket and first wall. In addition to those groups, the two disciplinary groups are also organized to do the critical analyses of the existing INTOR-like designs; (G) Physics and (H) Engineering. (author)

  6. Engineering aspects of the INTOR design

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1983-01-01

    The INTOR engineering design has been strongly influenced by considerations for assembly and maintenance. A maintenance philosophy was established at the outset of the conceptual design to insure that the tokamak configuration would be developed to accommodate maintenance requirements. The main features of the INTOR design are summarized in this paper with primary emphasis on the impact of maintenance considerations. The most apparent configuration design feature is the access provided for torus maintenance. Particular attention was given to the size and location of superconducting magnets and the location of vacuum boundaries. All of the poloidal field (PF) coils are placed outside of the bore of the toroidal field (TF) coils and located above and below an access opening between adjacent TF coils through which torus sectors are removed. A magnet structural configuration consisting of mechanically attached reinforcing members has been designed which facilitates the open access space for torus sector removal. For impurity control, a single null poloidal divertor was selected over a double null design in order to maintain sufficient access for pumping and maintenance of the collector. A double null divertor was found to severely limit access to the torus with the addition of divertor collectors and pumping at the top. For this reason, a single null concept was selected in spite of the more difficult design problems associated with the required asymmetric PF system and higher particle loadings. Tokamak support systems and the reactor building and facilities are also important to the overall design evolution and were included in the conceptual design effort. However, this paper discusses only the primary tokamak systems. (author)

  7. Design of a materials testing experiment for the INTOR

    International Nuclear Information System (INIS)

    Vogel, M.A.; Opperman, E.K.

    1981-01-01

    The United States, Japan, USSR and the European community are jointly participating in the design of an International Tokamak Reactor called INTOR. In support of the US contribution to the INTOR design, the features of an experiment for bulk neutron irradiation damage studies were developed. It is anticipated that materials testing will be an important part of the programmatic mission of INTOR and consequently the requirements for materials testing in INTOR must be identified early in the reactor design to insure compatibility. The design features of the experiment, called a Channel Test, are given in this paper. The major components of the channel test are the water cooled heat sink (channel module) and the specimen capsule. The temperature within each of the 153 specimen capsules is predetermined by engineering the thermal barrier between the specimen capsule and heat sink. Individual capsules can be independently accessed and are designed to operate at a predetermined temperature within the range of 50 to 700 0 C. The total irradiation volume within a single channel test is 45 liters. Features of the channel test that result in experimental versatility and simplified remote access and handling are discussed

  8. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

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

    1981-08-01

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

  9. International tokamak reactor conceptual design overview

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.

    1981-01-01

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

  10. Japanese contributions to IAEA INTOR Workshop, phase IIA

    International Nuclear Information System (INIS)

    Tomabechi, Ken; Hiraoka, Tohru; Fujisawa, Noboru; Nishio, Satoshi; Sawada, Yoshio; Kobayashi, Takeshi.

    1982-11-01

    Japanese contributions at INTOR Phase 2A are summarized as a first three chapters of the report, Japanese Contributions to INTOR Workshop, Phase 2A. This report contains Introduction, Summary of the total report and INTOR Concepts. (author)

  11. Conceptual design of the INTOR first-wall system

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Majumdar, S.; Mattas, R.F.; Turner, L.; Jung, J.; Abdou, M.A.; Bowers, D.; Trachsel, C.; Merrill, B.

    1981-10-01

    The design concept and performance characteristics of the first-wall design for the phase-1 INTOR (International Tokamak Reactor) study is described. The reference design consists of a water-cooled stainless steel panel. The major uncertainty regarding the performance of the bare stainless steel wall relates to the response of a thin-melt layer predicted to form on limited regions during a plasma disruption. A more-complex backup design, which incorporates radiatively cooled graphite tiles on the inboard wall, is briefly described.

  12. Conceptual design of the INTOR first-wall system

    International Nuclear Information System (INIS)

    Smith, D.L.; Majumdar, S.; Mattas, R.F.; Turner, L.; Jung, J.; Abdou, M.A.; Bowers, D.; Trachsel, C.; Merrill, B.

    1981-10-01

    The design concept and performance characteristics of the first-wall design for the phase-1 INTOR (International Tokamak Reactor) study is described. The reference design consists of a water-cooled stainless steel panel. The major uncertainty regarding the performance of the bare stainless steel wall relates to the response of a thin-melt layer predicted to form on limited regions during a plasma disruption. A more-complex backup design, which incorporates radiatively cooled graphite tiles on the inboard wall, is briefly described

  13. IAEA INTOR workshop report, group 8

    International Nuclear Information System (INIS)

    Tamura, Sanae; Shimada, Ryuichi; Miya, Naoyuki; Shinya, Kichiro; Kishimoto, Hiroshi

    1979-10-01

    This report provides material for discussion in Group 8, Power Supply and Transfer, of the IAEA Workshop on INTOR. A new system for the poloidal field power supply for INTOR is proposed and its overall system design is described. The results of simulation calculation of the system are also given. (author)

  14. On materials problems in INTOR

    International Nuclear Information System (INIS)

    Schiller, P.

    1981-01-01

    In INTOR, an attempt has been made to define more precisely the performance limits of important parts of the device. For the first wall, it appears that under the assumed conditions, the erosion by the plasma will be the most important problem. Part of this problem may disappear with the advancement of plasma physics, but other erosion mechanisms will remain and therefore limit the lifetime of this component and influence heavily the burning of the plasma. The breeding blanket requires particular attention in order to achieve a reasonable breeding factor and to maintain at a low level the tritium inventory. The necessity to keep the tritium inventory down can be met better by temperatures which are high enough to allow an easy outgasing of the tritium from the breeder material. This temperature may bring the canning material in a range where swelling is important and lifetime limiting. Radiation damage in the superconducting coils will be low since otherwise the radiation heating would be difficult to be eliminated by the liquid helium, but the damage may be high enough to put seriously in question the integrity of the insulating material, especially under cyclic conditions. The INTOR exercise may be considered an excellent way to show up where to-days materials have their limits in a fusion reactor. It leads inevitably to the conclusion that, for the construction of economic power reactors, it will be necessary to start immediately an alloy development programme. (orig./GG)

  15. IAEA INTOR workshop report, groups 2, 5, 7, 9, 10 and 15

    International Nuclear Information System (INIS)

    1980-02-01

    In order to prove scientific feasibility of magnetic confinement fusion, large fusion devices are under construction in several countries (JT-60 in Japan, T-15 in U.S.S.R., TFTR in U.S.A. and JET in EC). International Tokamak Reactor (INTOR) Workshop was organized by the International Atomic Energy Agency (IAEA) to identify roles, objectives and characteristics of the next generation fusion device. This report is a compilation of the home task reports of six groups on INTOR engineering aspects by Japan Atomic Energy Research Institute for workshop sessions 2 and 3 held in 1979. Tasks of the respective groups are group 2: first wall/blanket/shield, group 5: magnetics, group 7: systems integration and structure, group 9: assembly and remote maintenance, group 10: radiation shielding and personnel access, group 15: safety and environment. (author)

  16. Thermal hydraulics of the impurity control system for FED/INTOR

    International Nuclear Information System (INIS)

    Cha, Y.S.; Mattas, R.F.; Abdou, M.A.; Haines, J.R.

    1983-01-01

    This paper addresses two important aspects of thermal hydraulics related to the design of the impurity control system (limiter and divertor) of the Fusion Engineering Device (FED) and the International Tokamak Reactor (INTOR). The first part of the paper is devoted to the determination of temperature distributions in various combinations of the coating/structural materials proposed for the limiter/divertor of FED and INTOR. The second part of the paper describes the analysis of the tangential motion of the melt layer under the influence of magnetic force during plasma disruption. The results of both analysis provide inputs to the determination of the life time of the limiter (or divertor) which is the most critical problem for the impurity control system as far as engineering and materials consideration is concerned

  17. Key features of INTOR nuclear systems

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1981-05-01

    The conceptual design effort for INTOR was broadly defined in three areas: (1) Plasma Physics, (2) Engineering, and (3) Nuclear Systems. This paper is devoted to a summary of the Nuclear Systems effort. The emphasis is placed on the First Wall, Breeding Blanket, and Divertor

  18. Tritium-management requirements for D-T fusion reactors (ETF, INTOR, FED)

    International Nuclear Information System (INIS)

    Finn, P.A.; Clemmer, R.G.; Misra, B.

    1981-10-01

    The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design

  19. Japanese contributions to IAEA INTOR Workshop, Phase Two A, Part 3

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kobayashi, Takeshi; Yamada, Masao

    1988-02-01

    This report corresponds to Chapter VII of Japanese contribution report to IAEA INTOR Workshop, Phase two A, Part 3. The discussions on the following tasks are reported. Task 1. Comparative study of vertical and horizontal access configuration. Task 2. Shape memory alloy. Task 3. Ferromagnetic inserts for ripple reduction. Task 4. PF coil redundancy. Task 5. Rapid replacement of divertor and first wall. Task 6. Containment of tritium and activated dust. In the above tasks, Task 2, 3 and 4 come from the Specialist Meeting on Tokamak Concept Innovation in January, 1986. (author)

  20. Tokamak

    International Nuclear Information System (INIS)

    Meglicki, Z.

    1995-01-01

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs

  1. Engineering testing requirements in FED/INTOR

    International Nuclear Information System (INIS)

    Abdou, M.A.; Nygren, R.E.; Morgan, G.D.; Trachsel, C.A.; Wire, G.; Oppermann, E.; Puigh, R.; Gold, R.E.

    1982-10-01

    The FED/INTOR critical issues activity has addressed three key testing requirements that have the largest impact on the design, operation and cost of FED/INTOR. These are: (1) the total testing time (fluence) during the device lifetime, (2) the minimum number of back-to-back cycles, and (3) the neutron wall load (power density in the first wall/blanket). The testing program activities were structured into three tasks in order to define the benefits, and in some cases, costs and risks of these testing requirements. The three tasks were carried out with wide participation of experts from a number of organizations in the United States. Similar effort was performed by Japan, the European Community and the Soviet Union

  2. INTOR impurity control and first wall system

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1983-04-01

    The highlights of the recent INTOR effort on examining the key issues of the impurity control/first wall system are summarized. The emphasis of the work was an integrated study of the edge-region physics, plasma-wall interaction, materials, engineering and magnetic considerations associated with the poloidal divertor and pump limiter. The development of limiter and divertor collector plate designs with an acceptable lifetime was a major part of the work

  3. IAEA INTOR workshop report, group 11

    International Nuclear Information System (INIS)

    Takeda, Tatsuoki; Kurita, Gen-ichi; Matsuura, Toshihiko; Ninomiya, Hiromasa; Tsunematsu, Toshihide

    1980-01-01

    Data base assessments of INTOR are presented. Included are the theoretical beta -limits, experimental results of current disruption in DIVA, wide range of parameter study of plasma shaping, effect of conducting shell on positional and ballooning mode, and the method of disruption control. Good shell effect of the SS blanket is examined. Also shown is the possibility of the equilibrium control of minor radius expansion. Design of hybrid poloidal coil system with exterior coils has been carried out. (author)

  4. IAEA INTOR workshop report, group 6

    International Nuclear Information System (INIS)

    Shiraishi, Kensuke; Kondo, Tatsuo; Nasu, Shoichi; Takamura, Saburo; Oku, Tatsuo

    1979-10-01

    For materials data base assessment, selected properties of the materials important in INTOR are evaluated on the basis of data obtained in Japan. Materials covered are austenitic steels, non-magnetic steels, and titanium alloys as structural materials; helium and water for coolant; lithium oxide for breeder; aluminum and copper for stabilizer and organic insulator as magnet materials; graphite for non-metallic materials. (author)

  5. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-01-01

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

  6. Radiological safety assessment of a reference INTOR facility

    International Nuclear Information System (INIS)

    Khan, T.A.; Stasko, R.R.; Watts, R.T.; Shaw, G.; Morrison, C.A.; Russell, S.; Kempe, T.; Zimmerman, R.

    1985-03-01

    This report consists of a number of separate studies all of which were performed in support of INTOR Critical Issue D: Tritium Containment and Personnel Access vs Remote Maintenance. The common thread running through these studies is the radiological safety element in the design and operation of the INTOR facility. The intent is to help establish a firm basis for comparisons between a reactor cell maintenance option which requires personnel access, and one which involves completely remote maintenance

  7. Design studies of an aluminum first wall for INTOR

    International Nuclear Information System (INIS)

    Powell, J.R.; Fillo, J.A.; Yu, W.S.; Hsieh, S.Y.; Pearlman, H.; Kramer, R.; Franz, E.; Craig, A.; Farrell, K.

    1980-01-01

    Besides the high erosion rates (including evaporation) expected for INTOR, there may also be high heat fluxes to the first wall, e.g., approx. 9 (Case I) to 24 (Case II) W/cm 2 , from two sources - radiation and charge exchange neutrals. There will also be internal heat generation by neutron and gamma deposition. An aluminum first wall design is analyzed, which substantially reduces concerns about survivability of the first wall during INTOR's operating life

  8. Self-pumping inpurity control systems for INTOR

    International Nuclear Information System (INIS)

    Brooks, J.N.; Mattas, R.F.; Smith, D.L.; Hassanein, A.M.

    1987-01-01

    Two self-pumping systems have been examined for use as the INTOR impurity control system. The systems work by trapping helium in freshly deposited metal surface layers on or near the divertor plate. A slot divertor concept using vanadium or other trapping material appears to be both feasible and mechanically simple, and offers significant advantages in cost, reduced complexity, and helium pumping efficiency for the INTOR design

  9. Impurity and particle control for INTOR

    International Nuclear Information System (INIS)

    Post, D.

    1985-02-01

    The INTOR impurity control system studies have been focused on the development of an impurity control system which would be able to provide the necessary heat removal and He pumping while satisfying the requirements for (1) minimum plasma contamination by impurities, (2) reasonable component lifetime (approx. 1 year), and (3) minimum size and cost. The major systems examined were poloidal divertors and pumped limiters. The poloidal divertor was chosen as the reference option since it offered the possibility of low sputtering rates due to the formation of a cool, dense plasma near the collector plates. Estimates of the sputtering rates associated with pumped limiters indicated that they would be too high for a reasonable system. Development of an engineering design concept was done for both the poloidal divertor and the pumped limiter

  10. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  11. Comparison of tokamak burn cycle options

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  12. Engineering aspects of the JAERI proposal for INTOR, (2)

    International Nuclear Information System (INIS)

    1979-11-01

    To assess the data base of INTOR, a design study of the INTOR-J reactor system was carried out. Purpose of the study to evaluate the following five could be attained: (1) validity of the guiding parameters (especially those related to reactor size), (2) feasibility of a reactor with poloidal divertors, (3) design problems and research/development items in respective components, (4) possibility of repair and maintenance, and (5) design of the test sections for tritium breeding and electric power generation. (Author)

  13. Heat removal in INTOR via a toroidal limiter

    International Nuclear Information System (INIS)

    Mioduszewski, P.

    1981-01-01

    In the present paper the potential of removing about 100 MW of thermal plasma power via a toroidal limiter in INTOR is studied. The heat flux distributions on various limiter configurations are calculated and the thermal response of a graphite tile limiter is estimated on the base of a one-dimensional heat conduction approach. The evaporation rates which have to be expected for the given energy flux densities and radiation cooled graphite tiles are evaluated. According to the present understanding it should be possible to remove 100 MW power from the INTOR plasma via a radiation cooled toroidal limiter. (author)

  14. Shielding and maintainability in an experimental tokamak

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  15. Japanese contributions to IAEA INTOR workshop, phase IIA

    International Nuclear Information System (INIS)

    Fujisawa, Noboru; Sugihara, Masayoshi; Shimada, Michiya; Saito, Seiji.

    1982-11-01

    This report corresponds to Chapter VI of Japanese contribution report to IAEA INTOR workshop, Phase IIA. Special emphasis is placed on pumped limiter analysis for comparative studies between limiter and divertor concepts. Pumping characteristics of divertor/limiter and radiation cooling of diverted plasmas by impurities are also intensively studied. (author)

  16. Comparative analysis of INTOR-like designs: Status and preliminary results

    International Nuclear Information System (INIS)

    Flanagan, C.A.; Galambos, J.D.; Peng, P.K.M.; Reid, R.L.; Strickler, D.J.

    1987-07-01

    This document provides an interim status report on the INTOR tasks related to the critical analysis of INTOR-like designs. At the INTOR Workshop in December 1986, and updated at the INTOR related Specialists' Meeting in March 1987, a series of comparative analysis studies was established. Initial results were targeted for completion for the July INTOR meeting and final results completed for the November INTOR meeting. The goal of the analytic comparisons is to evaluate and determine the specific impact of how different choices affect the design. The results of such an analytic determination should allow the international community to identify and agree upon which aspects of choice have what influence on the International Thermonuclear Experimental Reactor (ITER) design

  17. Japanese contributions to IAEA INTOR workshop, phase two A, part 2 chapter VII: technical benefit

    International Nuclear Information System (INIS)

    Itoh, Shin-ichi; Iida, Hiromasa; Tomabechi, Ken; Tone, Tatsuzo; Fujisawa, Noboru.

    1985-07-01

    This report corresponds to Chapter VII of Japanese contribution report to IAEA INTOR Workshop, Phase Two A, Part 2. The purpose of technical benefit study is to examine the implications of having different manufacturers fabricate components of a major system of INTOR. A systematic examinations of advantages and disadvantages of designing and fabricating major INTOR components in the frame of one international joint projects is performed. (author)

  18. Japanese contributions to IAEA INTOR workshop, phase two A, part 2, chapter X: nuclear

    International Nuclear Information System (INIS)

    Tomabechi, Ken; Iida, Hiromasa; Honda, Tsutomu

    1985-07-01

    This report corresponds to Chapter X of Japanese contribution report to IAEA INTOR Workshop, Phase Two A, Part 2. Data base of the nuclear technology for INTOR is assessed focusing on the four fields; Blanket, Shield, Tritium and Safty. Ongoing R and D programs in these fields are also reviewed and required new programs are discussed. (author)

  19. Japanese contributions to IAEA INTOR workshop, phase two A, part 2, chapter IX: engineering

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Seki, Masahiro; Sawada, Yoshio

    1985-07-01

    This report corresponds to Chapter IX of Japanese contribution report to IAEA INTOR Workshop, Phase Two A, Part 2. Data base assessment are made for systems engineering, magnet systems, torus systems, and NBI heating systems. R and D programme and impact on INTOR design are also specified. In addition to the data base assessment, studies have been made for several new tasks. (author)

  20. Japanese contributions to IAEA INTOR workshop, phase IIA

    International Nuclear Information System (INIS)

    Hiraoka, Toru; Fujisawa, Noboru; Nishio, Satoshi

    1982-11-01

    This report is one of the chapters of Japanese National Report of INTOR Workshop, Phase 2A. At this Phase 2A, the pumped limiter has been investigated as impurity control concept. The curved-surface double-edged pumped limiter was employed as results of various features. Surface materials are assessed from various aspects such as sputtering, behavior at plasma disruption, thermal characteristics, electro-magnetics, attachment method to substrate materials. Application of newly developed SiC with high thermal conductivity is also investigated. (author)

  1. High current density magnets for INTOR and TIBER

    International Nuclear Information System (INIS)

    Miller, J.R.; Henning, C.D.; Kerns, J.A.; Slack, D.S.; Summers, L.T.; Zbasnik, J.P.

    1986-12-01

    The adoption of high current density, high field, superconducting magnets for INTOR and TIBER would prove beneficial. When combined with improved radiation tolerance of the magnets to minimize the inner leg shielding, a substantial reduction in machine dimensions and capital costs can be achieved. Fortunately, cable-in-conduit conductors (CICC) which are capable of the desired enhancements are being developed. Because conductor stability in a CICC depends more on the trapped helium enthalpy, rather than the copper resistivity, higher current densities of the order of 40 A/mm 2 at 12 T are possible. Radiation damage to the copper stabilizer is less important because the growth in resistance is a second-order effect on stability. Such CICC conductors lend themselves naturally to niobium-tin utilization, with the benefits of the high current-sharing temperature of this material being taken to advantage in absorbing radiation heating. When the helium coolant is injected at near the critical pressure, Joule-Thompson expansion in the flow path tends to stabilize the fluid temperature at under 6 K. Thus, higher fields, as well as higher current densities, can be considered for INTOR or TIBER

  2. Conceptual design of an electricity generating tritium breeding blanket sector for INTOR/NET

    International Nuclear Information System (INIS)

    Bond, A.

    1984-01-01

    A study is made of a fusion reactor power blanket and its associated equipment with the objective of producing a conceptual design for a blanket sector of INTOR, or one of its national variants (e.g. NET), from which electricity could be generated simultaneously with the breeding of tritium. (author)

  3. Computational studies in tokamak equilibrium and transport

    International Nuclear Information System (INIS)

    Braams, B.J.

    1986-01-01

    This thesis is concerned with some problems arising in the magnetic confinement approach to controlled thermonuclear fusion. The work address the numerical modelling of equilibrium and transport properties of a confined plasma and the interpretation of experimental data. The thesis is divided in two parts. Part 1 is devoted to some aspects of the MHD equilibrium problem, both in the 'direct' formulation (given an equation for the plasma current, the corresponding equilibrium is to be determined) and in the 'inverse' formulation (the interpretation of measurements at the plasma edge). Part 2 is devoted to numerical studies of the edge plasma. The appropriate Navier-Stokes system of fluid equations is solved in a two-dimensional geometry. The main interest of this work is to develop an understanding of particle and energy transport in the scrape-off layer and onto material boundaries, and also to contribute to the conceptual design of the NET/INTOR tokamak reactor experiment. (Auth.)

  4. A comparison of burn cycle options for Tokamak reactors

    International Nuclear Information System (INIS)

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

    1985-01-01

    Consideration of burn cycle options for commercial tokamaks shows that there is substantial motivation to achieve steady state operation. This is partly due to longer replacement periods for first wall and impurity control components, but, in addition, large cost savings are found when magnets, power supplies, and the energy transfer system are not frequently pulsed. The hybrid burn cycle, with a combination of ohmic and noninductive current drive, does not significantly improve the economics of ohmically-driven commercial reactors with large major radius. However, an INTOR-class device has a critically small hole in the doughnut, and we find for this size tokamak that the hybrid cycle is preferred over ohmically-driven operation

  5. INTOR critical issue D: maintainability. Tritium containment and personnel access vs remote maintenance, Chapter VI of the US INTOR report for Phase Two A, Part 2

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Finn, P.A.; Gohar, Y.

    1984-01-01

    The purpose of this study is to compare the benefits and costs associated with personnel access mmaintenance procedures compared to those of all-remote maintenance procedures. The INTOR Phase Two A, Part I configuration was used to make this comparison. For both approaches, capital and operating costs were considered to first order, maintenance equipment requirements were investigated, maintenance requirements common to both approaches and unique to each were identified, tritium handling requirements were outlined, and maintenance scenarios and device downtime were developed for both. In addition, estimates of person-rem exposure were made for the personnel access approach

  6. Thermal and structural limitations for impurity-control components in FED/INTOR

    International Nuclear Information System (INIS)

    Majumdar, S.; Cha, Y.; Mattas, R.; Abdou, M.; Cramer, B.; Haines, J.

    1983-02-01

    The successful operation of the impurity-control system of the FED/INTOR will depend to a large extent on the ability of its various components to withstand the imposed thermal and mechanical loads. The present paper explores the thermal and stress analyses aspects of the limiter and divertor operation of the FED/INTOR in its reference configuration. Three basic limitations governing the design of the limiter and the divertor are the maximum allowable metal temperature, the maximum allowable stress intensity and the allowable fatigue life of the structural material. Other important design limitations stemming from sputtering, evaporation, melting during disruptions, etc. are not considered in the present paper. The materials considered in the present analysis are a copper and a vanadium alloy for the structural material and graphite, beryllium, beryllium oxide, tungsten and silicon carbide for the coating or tile material

  7. 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

  8. Japanese contributions to IAEA INTOR workshop, phase two A, part 2, chapter VI: maintainability

    International Nuclear Information System (INIS)

    Tomabechi, Ken; Iida, Hiromasa; Honda, Tsutomu

    1985-07-01

    This report corresponds to Chapter VI of Japanese contribution report to IAEA INTOR Workshop, Phase Two A, Part 2. In this report, we have compared two different reactor concepts, one is based on the personnel access concept where the personnel access is possible to the outside of the torus for maintenance one day after shutdown and the other where the personnel access is not necessary and maintenance is performed in full-remote manner. The results are described from various view points such as reactor configuration, tritium confinement, safety, shielding, maintenance scenario and cost. Data base assessments of maintenance equipment are also reported. (author)

  9. Tokamak power reactor ignition and time dependent fractional power operation

    International Nuclear Information System (INIS)

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-06-01

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

  10. Near-wall plasma in a tokamak thermonuclear reactor

    International Nuclear Information System (INIS)

    Pistunovich, V.I.

    1991-01-01

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

  11. Tokamak formation and sustainment by tokamak injection

    International Nuclear Information System (INIS)

    Farengo, R.; Jarboe, T.R.

    1991-01-01

    The authors propose here a new helicity injection method for tokamak formation and sustainment that has high efficiency, conserves toroidal symmetry and is inductively driven. The basic idea is to inject a small tokamak (source tokamak) into a larger tokamak (steady tokamak). This current drive scheme eliminates the need for the ohmic heating transformer in the steady tokamak allowing the formation of very small aspect ratio tokamaks (Spherical Tori). Thus, steady state operation and high beta can be realized simultaneously. The method can also be applied to a larger aspect ratio tokamak and used in conjunction with the standard inductive formation technique. In order to allow for translation the ohmic heating coil used to produce the source tokamaks must be fed from one end (as in the CSS device) and the toroidal field coil must link both tokamaks. After formation the source tokamaks are accelerated towards the steady tokamak by a mirror field and the tension of the field lines that wrap around both tokamaks (producing a doublet type configuration). In a tokamak the helicity is proportional to the current. This indicates that (assuming helicity is conserved during the merging process) a steady state situation will result if the helicity supplied by the source tokamaks is equal to the helicity dissipated by the steady tokamak. Assuming that source tokamaks of helicity K s are injected with frequency f, the steady state condition can be written as: fK s = 2V t Ψ t = K t /τ K where V t , Ψ t , K t and τ K are the ohmic loop voltage, toroidal flux, helicity and helicity decay time of the steady tokamak. A simple calculation shows that the DIII-D tokamak could be sustained by injecting source tokamaks with R = 1.20 m, a = 0.23 m and I = 151 kA at a frequency of 120 Hz. 1 ref

  12. 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

  13. Japanese contributions to IAEA INTOR Workshop, Phase Two A, Part 3

    International Nuclear Information System (INIS)

    Kobayashi, Takeshi; Iida, Hiromasa; Abe, Tadashi

    1988-01-01

    This report includes Japanese contributions presented at INTOR Workshop Phase 2A, Part 3 relevant to the studies of technical critical issues of blanket, first wall, divertor plate and shield designs, and new database concerning these areas. New database for ceramic breeder materials, high heat flux materials, high heat flux experiments and fabrication experience are included. A comparison study of blanket concepts and design studies of candidate blankets have been carried out. The blanket with lithium oxide breeder is selected as the leading blanket concept because of its high breeding performance and good tritium release behavior. Two types of first wall concepts, bare stainless steel first wall concept and protected first wall concepts, are evaluated. For disruptions, the behaviors of crack propagation are examined based on J integrals. The local guard limiter concept with easy maintainability is selected as a reference considering currently rather uncertain disruption scenarios. Tungsten is selected as the reference armor material of divertor plate because of its good thermal and sputtering resistance. The inboard shield thickness has been optimized both for stainless steel material and tungsten material. Blanket engineering test program of FER is also described in this report. Major basic engineering tests on blanket performances such as neutronics, thermal-hydraulics, tritium recovery etc. can be performed by test elements and modules. (author)

  14. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

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

  15. 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.)

  16. 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

  17. 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

  18. Status of tokamak research

    Energy Technology Data Exchange (ETDEWEB)

    Rawls, J.M. (ed.)

    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. (MOW)

  19. Particle and momentum confinement in tokamak plasmas with unbalanced neutral beam injection and strong rotation

    International Nuclear Information System (INIS)

    Malik, M.A.

    1988-01-01

    There is a self-consistent theory of the effects of neutral beam injection on impurity transport in tokamak plasmas. The theory predicts that co-injection drives impurities outward and that counter-injection enhances the normally inward flow of impurities. The theory was applied to carry out a detailed analysis of the large experimental database from the PLT and the ISX-B tokamaks. The theory was found to generally model the experimental data quite well. It is, therefore, concluded that neutral beam co-injection can drive impurities outward to achieve clean central plasmas and a cool radiating edge. Theoretical predictions for future thermonuclear reactors such as INTOR, TIBER II, and ITER indicated that neutral beam driven flow reversal might be an effective impurity control method if the rate of beam momentum deposited per plasma ion is adequate. The external momentum drag, which is a pivotal concept in impurity flow reversal theory, is correctly predicted by the gyroviscous theory of momentum confinement. The theory was applied to analyze experimental data from the PLT and the PDX tokamaks with exact experimental conditions. The theory was found to be in excellent agreement with experiment over a wide range of parameters. It is, therefore, possible to formulate the impurity transport theory from first principles, without resort to empiricism

  20. 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

  1. 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.

  2. 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

  3. 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)

  4. Approach to ignition of tokamak reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sigmar, D.J.

    1981-02-01

    Recent transport modeling results for JET, INTOR, and ETF are reviewed and analyzed with respect to existing uncertainties in the underlying physics, the self-consistency of the very large numerical codes, and the margin for ignition. The codes show ignition to occur in ETF/INTOR-sized machines if empirical scaling can be extrapolated to ion temperatures (and beta values) much higher than those presently achieved, if there is no significant impurity accumulation over the first 7 s, and if the known ideal and resistive MHD instabilities remain controllable for the evolving plasma profiles during ignition startup.

  5. 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)

  6. 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.

  7. 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

  8. Research using small tokamaks

    International Nuclear Information System (INIS)

    1989-07-01

    These proceedings of the IAEA-sponsored meeting held in Nice, France 10-11 October, 1988, contain the manuscripts of the 21 reports dealing with research using small tokamaks. The purpose of this meeting was to highlight some of the achievements of small tokamaks and alternative magnetic confinement concepts and assess the suitability of starting new programs, particularly in developing countries. Papers presented were either review papers, or were detailed descriptions of particular experiments or concepts. Refs, figs and tabs

  9. Tokamak simulation code manual

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

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

  10. Tokamak simulation code manual

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. Joint research using small tokamaks

    International Nuclear Information System (INIS)

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

    2005-01-01

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  12. Cyclotron radiation as Tokamak diagnostics

    International Nuclear Information System (INIS)

    Fiedler-Ferrari, N.

    1985-01-01

    A brief introduction to the use of Electron Cyclotron Emission as diagnostics in tokamaks is made. The utilization feasibility of this dignostics in the TBR-1 and TTF2A tokamaks is discussed. (L.C.) [pt

  13. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-03-01

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

  14. 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

  15. 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

  16. 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

  17. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

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

  18. High beta tokamak instabilities

    International Nuclear Information System (INIS)

    Bateman, G.

    1977-01-01

    Theoretical predictions using the ideal MHD model indicable that large-scale ballooning modes should appear when the average beta is raised about 1 to 2% in present-day tokamak geometries or 5 to 10% in more optimized geometries. The onset of instability is predicted to be sudden and the behavior of ballooning modes to be strikingly different from the saw-tooth and Mirnov oscillations experimentally observed at low beta. Conditions close to the predicted onset were achieved in ORMAK with no noticeable change in plasma behavior. Experiments are planned for the ISX tokamak to test the beta limit. 15 references, 3 figures

  19. Long Pulse Technology Tokamak

    International Nuclear Information System (INIS)

    Jernigan, T.C.

    1978-01-01

    The LPTT tokamak is a non-circular tokamak (R = 1.5 m, a = .45 m) proposed by ORNL for extended pulse operation at high β (5%) and reactor level wall power loading (40 w/cm 2 ). The toroidal field coils are superconducting and a super-conducting bundle divertor is proposed for active impurity control. All systems are designed for continuous operation which will provide pulse lengths > 20 seconds with a 6 to 10 weber flux swing. Experimental access and flexibility in operation are primary design goals

  20. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Pare, V.K.

    1983-01-01

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

  1. 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

  2. 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

  3. 50 years of tokamaks

    Czech Academy of Sciences Publication Activity Database

    Mlynář, Jan; Řípa, Milan

    2008-01-01

    Roč. 2, č. 2 (2008), s. 7-7 ISSN 1818-5355 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * history Subject RIV: BL - Plasma and Gas Discharge Physics http://www.efda.org/news_and_events/downloads/efda_newsletter/nl_2008_12.pdf

  4. 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)

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

    International Nuclear Information System (INIS)

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

    1988-09-01

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

  6. 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

  7. Density limits in Tokamaks

    International Nuclear Information System (INIS)

    Tendler, M.

    1984-06-01

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

  8. Energy confinement in tokamaks

    International Nuclear Information System (INIS)

    Sugihara, M.; Singer, C.

    1986-08-01

    A straightforward generalization is made of the ohmic heating energy confinement scalings of Pfeiffer and Waltz and Blackwell et. al. The resulting model is systematically calibrated to published data from limiter tokamaks with ohmic, electron cyclotron, and neutral beam heating. With considerably fewer explicitly adjustable free parameters, this model appears to give a better fit to the available data for limiter discharges than the combined ohmic/auxiliary heating model of Goldston

  9. 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

  10. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    NARCIS (Netherlands)

    Box, F. M. A.; Howard, J.; VandeKolk, E.; Meijer, F. G.

    1997-01-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines.

  11. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1994-01-01

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  12. The density limit in Tokamaks

    International Nuclear Information System (INIS)

    Alladio, F.

    1985-01-01

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

  13. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

  14. Analysis of the tritium inventory and permeation in a Li17Pb83 blanket proposed for INTOR and extrapolation to a power reactor

    International Nuclear Information System (INIS)

    Proust, E.

    1984-01-01

    A transient tritium hold-up and permeation model is developed and applied to a simplified conceptual design of a water cooled Li 17 Pb 83 blanket. Tritium inventories in the blanket associated with diffusivity, solubility and trapping effects are estimated. The model is applied to the INTOR/NET LiPb blanket design. Assuming a daily LiPb reprocessing frequency a tritium production rate of 64 grams per day yields a total tritium inventory in the blanket comparable to that of the tritium system or trapped in the first wall. The diffusion-limited permeation rate (neglecting oxide layers effects) reaches 4.2 g/day. The extrapolation of these results to reactor relevant conditions aggravates the permeation and the associated problems. (author)

  15. Nonlinear gyrokinetic tokamak physics

    International Nuclear Information System (INIS)

    Brizard, A.J.

    1990-01-01

    The gyrokinetic reduced description of low-frequency and small-perpendicular-wavelength nonlinear tokamak dynamics is presented in three different versions: the reduced dynamical description of test particles moving in electromagnetic fields; the reduced gyrokinetic description of the self-consistent interaction of particles and fields through the Maxwell-Vlasov equations; and the reduced description of nonlinear fluid motion. The unperturbed tokamak plasma is described in terms of a noncanonical Hamiltonian guiding-center theory. The unperturbed guiding-center tokamak plasma is then perturbed by gyrokinetic electromagnetic fields and consequently the perturbed guiding-center dynamical system acquires new gyrophase dependence. The perturbation analysis that follows makes extensive use of Lie-transform perturbation techniques. Because the electromagnetic perturbations affect both the Hamiltonian and the Poisson-bracket structure, the Phase-space Lagrangian Lie perturbation method is used. The description of the reduced test-particle dynamics is given in terms of a non-canonical Hamiltonian gyrocenter theory. The description of the reduced kinetic dynamics is concerned with the self consistent response of the guiding-center plasma and is described in terms of the nonlinear gyrokinetic Maxwell-Vlasov equations. It is also shown that the gyrokinetic Maxwell-Vlasov system possesses a gyrokinetic energy adiabatic invariant and that, in both the linear and nonlinear (quadratic) approximations, the corresponding energy invariants are exact. The description of the reduced fluid dynamics is concerned with the derivation of a closed set of reduced fluid equations. Three generations of reduced fluid models are presented: the reduced MHD equations; the reduced FLR-MHD equations; and the gyrofluid equations

  16. Tokamak instrumentation and controls

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-02-01

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

  17. Tokamak instrumentation and controls

    International Nuclear Information System (INIS)

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

    1979-02-01

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

  18. The TFR-600 Tokamak

    International Nuclear Information System (INIS)

    1977-11-01

    The new step of the Tokamak TFR, TFR 600, is described with its different aspects: physical objectives, modifications of the vacuum chamber and of the poloidal circuit, additionnal heatings. The nominal characteristics are: R=98 cm; a 0 or D 0 at 40 keV (power transmitted to the plasma); - ion cyclotron radiofrequency heating: 600 kW in the bandwidth 55-83 MHz; - and cluster injection: 100 KW at 600 keV (average mass of the H 0 clusters: 100-200 A.MU) [fr

  19. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    Minardi, E.

    1989-01-01

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

  20. Tokamak burn control

    International Nuclear Information System (INIS)

    Sager, G.T.

    1988-06-01

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  1. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

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

  2. Topology of tokamak orbits

    International Nuclear Information System (INIS)

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

    1978-09-01

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

  3. 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

  4. Canonical profiles in tokamaks

    International Nuclear Information System (INIS)

    Dnestrovskij, Yu.N.

    2002-01-01

    We consider the problem of the canonical profiles for tokamak plasma with arbitrary cross-section, taking into account two principles: 1) the free plasma energy minimum with the constraint of total current conservation and 2) the profile consistency. We deduce the Euler differential equation for the canonical profile of μ=1/q with two types of the boundary conditions: soft and stiff. The soft conditions correspond to the Kadomtsev solution for the circular cylinder. The stiff conditions describe a fast response of the plasma over the whole cross-section on the edge impact. Using the canonical profile of the current density, we calculate the critical gradients for the temperature, and create the transport model for the electron and ion temperatures and density. We show that, when the aspect ratio is diminished, or when the elongation increases, the canonical profiles become flatten. The similar tendency for the real profiles of the electron temperature was found in analysis of JET and START experiments. The obtained critical gradients were used to analysis of the experiments in tokamaks with moderate and tight aspect ratios. (author)

  5. 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

  6. JUST: Joint Upgraded Spherical Tokamak

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  7. 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.

  8. 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...

  9. 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)

  10. 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)

  11. 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)

  12. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  13. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    McWilliams, R.

    1988-01-01

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

  14. The ETE spherical Tokamak project

    International Nuclear Information System (INIS)

    Ludwig, Gerson Otto; Andrade, Maria Celia Ramos de; Barbosa, Luis Filipe Wiltgen

    1999-01-01

    This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)

  15. The ETE spherical Tokamak project

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Andrade, Maria Celia Ramos de; Barbosa, Luis Filipe Wiltgen [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] [and others]. E-mail: ludwig@plasma.inpe.br

    1999-07-01

    This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)

  16. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  17. Plasma turbulence in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  18. The Varennes tokamak

    International Nuclear Information System (INIS)

    Gregory, B.C.; Bolton, R.A.; Pacher, H.D.

    1983-01-01

    This article is a progress report on the Varennes Tokamak (TdeV), which is the main element in the Canadian research program on magnetic confinement fusion. The project is led by a group of five institutions: the Hydro-Quebec Research Institute (IREQ), the National Research Council - Energy, the University of Montreal, CANATOM Ltd., and MPB Technologies Inc. The TdeV will cost about 40 million dollars and will be built in a large hall at the IREQ high energy laboratory in Varennes. Operation in a quasi-stationary regime has been adopted as one of the primary research areas for the TdeV. First plasma is expected at the end of 1984 [fr

  19. 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

  20. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-09-01

    A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  1. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-01-01

    A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  2. 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

  3. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    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{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  4. Beta limits of a completely bootstrapped tokamak

    International Nuclear Information System (INIS)

    Weening, R.H.; Bondeson, A.

    1992-03-01

    A beta limit is given for a completely bootstrapped tokamak. The beta limit is sensitive to the achievable Troyon factor and depends directly upon the strength of the tokamak bootstrap effect. (author) 16 refs

  5. Bibliography of fusion product physics in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-09-01

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

  6. 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.

  7. 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

  8. 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)

  9. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

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

  10. 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

  11. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  12. Tokamak SST-1: an over-view

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2002-01-01

    Steady State Tokamak SST-1 is in advanced stage of fabrication at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak with superconducting magnets. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas for 1000 s with significant elongation (K) and triangularity (δ). The choice of the parameters is dictated by the physics and technology goals viz. (a) to control and study strongly shaped single and double null divertor plasma, (b) explore advanced tokamak plasma regimes, (c) steady state particle and heat removal from the device, (d) design and operation of large volume superconducting magnets, (e) non-inductive steady state current drive, (f) methods of plasma heating and (g) material technologies

  13. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  14. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...

  15. Tokamak Plasmas: Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...

  16. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    1975-06-01

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  17. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...wavelength polariza- tion field produced by the curvature and field gradient drifts [15]. The growth rate is y = Vs[k/R 2 = [T(eV)/X(cm)J 2 3.3 x 105 sec

  18. TECHNOLOGIES TO OPTIMIZE ADVANCED TOKAMAK

    Energy Technology Data Exchange (ETDEWEB)

    SIMONEN, TC

    2004-01-01

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

  19. 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

  20. Tokamak impurity-control techniques

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1980-01-01

    A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)

  1. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254 ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  2. An enhanced tokamak startup model

    Science.gov (United States)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

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

  3. Computer predictions for future Tokamaks

    International Nuclear Information System (INIS)

    Duechs, D.F.

    1978-01-01

    Proceeding from a reasonable agreement with existing experimental results, this lecture presents radial particle and energy transport computations which extrapolate to large (up to reactor dimensions) future Tokamaks. Special consideration is given to the behavior of alpha-particles, the influence of high-z impurities, and the thermal stability of the plasma

  4. 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

  5. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

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

    1988-04-01

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

  6. Engineering analysis of new Brazilian Tokamak

    International Nuclear Information System (INIS)

    Tuszel, A.G.

    1990-01-01

    The engineering basic headlines are described. A project for the construction of a new tokamak is being developed at the Institute of Physics, University of Sao Paulo. The tokamak named TBR-II will be a medium size tokamak using two high power generators of 15 MW each and concepted as a versatile device for plasma physics research of interest for thermonuclear fusion studies. (Author)

  7. A numerical study of tokamak edge turbulence

    International Nuclear Information System (INIS)

    Hu Shuanghui; Huang Lin; Qiu Xiaoming

    1993-01-01

    The tokamak edge turbulence which contains resistivity and impurity gradients and impurity radiation driven sources is studied numerically. The effect of ohmic dissipation on the evolution and saturation of this turbulence is investigated. The ohmic effect drops the saturation levels of fluctuations efficiently in high density tokamaks (such as Alcator), indicating that the ohmic effect plays an important role in the evolution of tokamak edge turbulence in high density devices

  8. 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

  9. Comprehensive numerical modelling of tokamaks

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  10. Advanced commercial Tokamak optimization studies

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  11. Flux driven turbulence in tokamaks

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  12. Starfire: a commercial tokamak reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  13. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Kieras, C.E.

    1982-12-01

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

  14. Equilibrium Reconstruction in EAST Tokamak

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  15. Dipole Map For Divertor Tokamaks

    International Nuclear Information System (INIS)

    Ali, Halima; Punjabi, Alkesh; Boozer, Allen

    2003-01-01

    Heat flux impinging on the collector plates of divertor tokamaks can be prodigious. Therefore, the problem of spreading the heat flux on plates is a crucial issue for divertor tokamaks such as ITER. Here we use method of maps /1,2/ to investigate this problem. Magnetic field lines in non-axisymmetric divertor tokamaks are a one and a half degree of freedom Hamiltonian system /1-3/. We represent the unperturbed magnetic topology by the Symmetric Simple Map (SSM) /4/ given by yn+1 = yn + 2kxn - 2k2yn (1 - yn), xn+1 = xn - kyn (1 - yn) - 2k2yn+1 (1 - yn+1). The effects of a current carrying coil placed externally across from X-point is represented by Dipole Map (DP) /4,5/ given by x n+1 = x n + 2δs 3 x n+1 (y n - y s + s/[x n+1 2 + (y n - y s + s) 2 ] 2 ), y n+1 = y n + δs 3 x n+1 ((y n - y s + s) 2 - x n+1 2 /[x n+1 2 + (y n - y s + s) 2 ] 2 ) δ is amplitude of high MN magnetic perturbation, s is the distance of coil from last good surface across from X point, and is the y coordinate of last good surface where it crosses the axis joining X point and O point across from X point. We fix k=0.3 and s = (1/2)|y s |. We calculate the increase in width of stochastic layer and area of footprint of field lines on divertor plate as δ is increased. We also calculate how connection length, toroidal and poloidal circuits and their fractal structures, the number, location and density of hot spots change with δ. Finally, we make conclusions about how the heat flux can be possibly controlled and reduced by applying external magnetic perturbation in divertor tokamaks

  16. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

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

  17. Tokamak plasma position dynamics and feedback control

    International Nuclear Information System (INIS)

    Burenko, L.; Bailey, J.M.

    1983-01-01

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

  18. Mercier criterion for high-β tokamaks

    International Nuclear Information System (INIS)

    Galvao, R.M.O.

    1984-01-01

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

  19. Magnetic confinement by Tokamak: physical aspects

    International Nuclear Information System (INIS)

    Tachon, J.

    1980-01-01

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

  20. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

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

  1. TGV, hutě a tokamak ITER

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    Leden (2017) ISSN 2464-7888 Institutional support: RVO:61389021 Keywords : fusion * ITER * tokamak * TGV * Pulse Power Electrical Network * Steady State Electrical Network Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.3pol.cz/cz/rubriky/jaderna-fyzika-a-energetika/1954-tgv-hute-a-tokamak-iter

  2. 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)

  3. Plasma detachment in divertor tokamaks

    Science.gov (United States)

    Leonard, A. W.

    2018-04-01

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

  4. The microwave Tokamak experiment (MTX)

    International Nuclear Information System (INIS)

    Thomassen, K.I.; Cohen, B.I.; Hooper, E.B.; Lang, D.D.; Nevins, W.M.

    1987-01-01

    A new experimental facility is being assembled at the Lawrence Livermore National Laboratory (LLNL) for studying microwave propagation and absorption in high density plasmas. A unique feature of the facility is the free electron laser (FEL) used to generate high peak power microwaves at 250 GHz, at a repetition rate so as to produce up to 2 MW of average power for up to 30 s. Called the Microwave Tokamak Experiment (MTX), the facility will be used for studies of plasma heating, current drive, and confinement

  5. Cluster storage for COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

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

    2012-01-01

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

  6. 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.)

  7. Channels in tokamak reactor shields

    International Nuclear Information System (INIS)

    Shchipakin, O.L.

    1981-01-01

    The results of calculations of neutron transport through the channels in the tokamak reactor radiation shields, obtained by the Monte Carlo method and by the method of discrete ordinates, are considered. The given data show that the structural materials of the channel and that of the blanket and shields in the regions close to it are subjected to almost the same irradiation as the first wall and therefore they should satisfy the technical requirements. The radiation energy release in the injector channel wall, caused by neutron shooting, substantially depends on the channel dimensions. At the channel large diameter (0.7-10 m) this dependence noticeably decreases. The investigation of the effect of the injector channel cross section form on the neutron flux density through the channel, testifies to weak dependence of shooting radiation intensity on the form of the channel cross section. It is concluded that measures to decrease unfavourable effect of the channels on the safety of the power tokamak reactor operation and maintenance cause substantial changes in reactor design due to which the channel protection must be developed at first stages. The Monte Carlo method is recommended to be used for variant calculations and when calculating the neutron flux functionals in specific points of the system the discrete ordinate method is preferred [ru

  8. CAT-D-T tokamaks

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  9. 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)

  10. Estimation of Zeff in Novillo Tokamak

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  11. 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

  12. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

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

  13. Research into controlled fusion in tokamaks

    International Nuclear Information System (INIS)

    Zacek, F.

    1992-01-01

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

  14. 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

  15. Definition of total bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    Ross, D.W.

    1995-01-01

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

  16. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    1981-01-01

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

  17. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    Caldas, I.L.; Vannucci, A.

    1985-01-01

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

  18. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

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

  19. 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

  20. 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

  1. Preliminary results of the TBR small tokamak

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  2. 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)

  3. Contributions to the INTOR workshop

    International Nuclear Information System (INIS)

    Miller, J.R.

    1987-01-01

    This paper discusses the use of Nb 3 Sn conductors as coils for toroidal fields in thermonuclear reactors. In particular, the mechanical properties are investigated for these conductors as well as substitute materials. 8 refs

  4. On estimation of reliability of a nuclear power plant with tokamak reactor

    International Nuclear Information System (INIS)

    Klemin, A.I.; Smetannikov, V.P.; Shiverskij, E.A.

    1982-01-01

    The results of the analysis of INTOR plant reliability are presented. The first stage of the analysis consists in the calculation of the INTOR plant structural reliability factors (15 ibs main systems have been considered). For each system the failure flow parameter (W(1/h)) and operational readiness Ksub(r) have been determined which for the plant as a whole besides these factors-technological utilization coefficient Ksub(TU) and mean-cycles-between failures Tsub(o). The second stage of the reliability analysis consists in investigating methods of improving its reliability factors reratively to the one calculated at the first level stage. It is shown that the reliability of the whole plant to the most essential extent is determined by the power supply system reliability. The following as to the influence extent on the INTOR plant reliability is the cryogenic system. Calculations of the INTOR plant reliability factors have given the following values: W=4,5x10 -3 1/h. Tsub(o)=152 h, Ksub(r)=0,71, Ksub(TU)=o,4 g

  5. Particle and energy balances in tokamak plasmas

    International Nuclear Information System (INIS)

    Tazima, Teruhiko

    1978-06-01

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

  6. Analysis of tokamak plasma confinement modes using the fast

    Indian Academy of Sciences (India)

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

  7. 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.)

  8. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Furth, H.P.

    1985-10-01

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

  9. Three novel tokamak plasma regimes in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Furth, H.P.

    1985-10-01

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

  10. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J.A.

    1994-11-30

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

  11. 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

  12. Tokamak disruption heat flux simulator

    International Nuclear Information System (INIS)

    Langhoff, M.; Hess, G.; Gahl, J.; Ingram, R.

    1990-01-01

    A coaxial plasma gun system, operating in the deflagration mode, has been built and fired at the University of New Mexico. This system, powered by a 100 kJ capacitor bank, was designed to give a variable pulse length of approximately 50-100 us. The gun is intended to deliver to a target an energy deposition density of 1 kJ per cm 2 via impact with a deuterium plasma possessing a highly directed energy. This system should simulate on the target, over an area of approximately 10 cm 2 , the heat flux of a tokamak plasma disruption on plasma facing components. Current diagnostics for the system are rather rudimentary but sufficient for determination of plasma pulse characteristics and energy transfer to target. Electrical measurements include bank voltage measured via resistive voltage dividers, and bank current measured via Rogowski coil. The shape of the plasma, its position relative to the target area, and the final impact area, is determined via open-shutter photography and the use of witness plates. Total energy deposited onto targets will be determined through simple calorimetry and careful target mass measurements. Preliminary results describing the ablation of carbon targets exposed to disruption like heat fluxes will be presented as well as a description of the experimental apparatus

  13. Neoclassical MHD equations for tokamaks

    International Nuclear Information System (INIS)

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

    1986-03-01

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

  14. Tokamak rotation and charge exchange

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. Stability analysis of tokamak plasmas

    International Nuclear Information System (INIS)

    Bourdelle, C.

    2000-10-01

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

  16. Erosion and deposition in tokamaks

    International Nuclear Information System (INIS)

    Staudenmaier, G.

    1985-01-01

    The flow of metal impurities from the wall and limiter to the plasma, and back towards the wall, is investigated using surface collection probes and subsequent surface analysis in order to understand impuritiy generation and impurity transport. Impurity fluxes and their scrapeoff lengths have been investigated for several years in a large number of tokamaks. The results are summarized and discussed. Erosion exceeding deposition was first observed to occur at limiterlike structures closest to the plasma edge. Recently, a new probe has been developed to measure quantitatively the erosion in ASDEX. Subsequent quantitative surface analysis is performed in situ by electron induced x-ray analysis. Erosion caused either by ions (limiter) or charge exchange neutrals (wall) can be investigated separately. The erosion at the wall is about two orders of magnitude smaller than the erosion at limiterlike structures, which is of the order of one monolayer per discharge. Simultaneous measurements of deposition and erosion have been performed to elucidate net values of deposition and erosion

  17. Activation analysis of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Selcow, E.C.

    1986-01-01

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

  18. Helicity content and tokamak applications of helicity

    International Nuclear Information System (INIS)

    Boozer, A.H.

    1986-05-01

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

  19. Effect of impurity radiation on tokamak equilibrium

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  20. Time - resolved thermography at Tokamak T-10

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. A Fast Shutdown Technique for Large Tokamaks

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  2. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-12-01

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

  3. Proposed tokamak poloidal field system development program

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-05-01

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

  4. Periodic disruptions in the MT-1 tokamak

    International Nuclear Information System (INIS)

    Zoletnik, S.

    1988-11-01

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

  5. Gas blanket fueling of a tokamak reactor

    International Nuclear Information System (INIS)

    Gralnick, S.L.

    1978-01-01

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

  6. Thermonuclear ignition in the next generation tokamaks

    International Nuclear Information System (INIS)

    Johner, J.

    1989-04-01

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

  7. Overview of spherical tokamak research in Japan

    Science.gov (United States)

    Takase, Y.; Ejiri, A.; Fujita, T.; Fukumoto, N.; Fukuyama, A.; Hanada, K.; Idei, H.; Nagata, M.; Ono, Y.; Tanaka, H.; Uchida, M.; Horiuchi, R.; Kamada, Y.; Kasahara, H.; Masuzaki, S.; Nagayama, Y.; Oishi, T.; Saito, K.; Takeiri, Y.; Tsuji-Iio, S.

    2017-10-01

    Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.

  8. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    Downum, W.B.

    1981-01-01

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

  9. Tokamak Engineering Technology Facility scoping study

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-03-01

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

  10. Can better modelling improve tokamak control?

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  11. Helicity injection experiment in the SINP tokamak

    International Nuclear Information System (INIS)

    Bhattacharyya, Krishnendu; Ray, Nihar Ranjan

    2000-01-01

    The current drive or sustainment in magnetized toroidal resistive plasmas can be though of as a 'balance' between helicity injection and dissipation. In the present work, the mechanisms of the 'balance' in the fluctuating magnetized resistive plasmas of the SINP tokamak, have been studied experimentally. The result shows that the oscillatory vertical magnetic field and oscillatory plasmas' velocity in a definite phase relationship causes the balancing effect between helicity injection and dissipation and thus sustainment of plasma current for a longer period of time has been observed in the resistive plasmas of the SINP tokamak. (author)

  12. Development of Operation Scenario for Spherical Tokamak at SNU

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

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

  14. 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

  15. Preliminary measurements on Tokamak KT-5

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  16. Fluctuation measurements with emissive probes in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Adámek, Jiří; Ďuran, Ivan; Hron, Martin; Stöckel, Jan; Balan, P.; Schrittwieser, R.; Ionita, C.; Martines, E.; Tichý, M.; Van Oost, G.

    2002-01-01

    Roč. 52, č. 10 (2002), s. 1115-1120 ISSN 0011-4626. [Workshop Role of Electric Fields in Plasma Confinement and Exhaust/5th./. Montreux, 23.06.2002-24.06.2002] Institutional research plan: CEZ:AV0Z2043910 Keywords : tokamak, electron-emissive Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.311, year: 2002

  17. MHD stability of vertically asymmetric tokamak equilibria

    International Nuclear Information System (INIS)

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

    1981-03-01

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

  18. Investigation of Tokamak Solid Divertor Target Options.

    Science.gov (United States)

    1981-05-26

    to fatigue nor is penetrated by sputtering. Norem and Bowers (47) report that a thickness of just 10 microns of beryllium should be sufficient to...Tokamak Surfaces", Journal of Nuclear Mater- ials, v.53, 1974, pp.107-110. 47. Norem , J. and D.A. Bowers, "Thin Low Z Coatings for Plasma Devices", ANL

  19. 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

  20. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1980-11-01

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

  1. Observations of arcing in the ISX tokamak

    International Nuclear Information System (INIS)

    Mioduszewski, P.; Clausing, R.E.; Heatherly, L.

    1979-01-01

    Arcing has been proposed as a major source of metal impurities in tokamak plasmas. Arc tracks have been observed in the ISX tokamak on the limiter, the inner-wall surface, and on the samples from the surface analysis station. Linear as well as fern-like arc tracks have been observed. From optical and SEM analysis of the tracks, it was estimated that about 10 16 to 10 17 atoms were released per arc. To study the influence of arcing on the tokamak discharge, an experiment was set up to measure electrical and optical signals of arcing in situ. In well controlled tokamak discharges, arcing was observed only during the initial breakdown of the plasma and during the quenching phase at the end of the discharge. In disrupted discharges, each plasma disruption was accompanied by arcing. The pulse-length of one single unipolar arc was measured to be about 50 μs and the current amplitude was typically about 20 A

  2. Microinstabilities in weak density gradient tokamak systems

    International Nuclear Information System (INIS)

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

    1986-04-01

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

  3. Experimental methods to study tokamak plasma stability

    International Nuclear Information System (INIS)

    Perez-Navarro, A.

    1978-01-01

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

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

    Indian Academy of Sciences (India)

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

  5. [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

  6. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

  7. Maintenance considerations of the STARFIRE commercial tokamak

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  8. UCLA Tokamak Program Close Out Report.

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Robert John [UCLA/retired

    2014-02-04

    The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

  9. 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

  10. 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.

  11. Runaway electrons in the SINP tokamak

    Indian Academy of Sciences (India)

    The highly energised runaway electrons (П 100 keV) is practically inevitable in the presence of an electric field. In tokamak discharges, these, on one hand, are a source of concern causing damages to the limiter and torus walls [1] whereas, on the other hand, it can be used as a diagnostic tool to determine the properties of ...

  12. The ARIES tokamak fusion reactor study

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  13. Plea for stellarator funding raps tokamaks

    International Nuclear Information System (INIS)

    Blake, M.

    1992-01-01

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

  14. 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.

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

    International Nuclear Information System (INIS)

    Bing-Ren, Shi

    2010-01-01

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

  16. First divertor operation on the HL-2A tokamak

    International Nuclear Information System (INIS)

    Yang Qingwei; Ding Xuantong; Yan Longwen; Xuan Weimin; Liu Dequan; Chen Liaoyuan; Song Xianming; Yuan Baoshan; Zhang Jinhua; Cao Zeng; Li Xiaodong; Mao Weicheng; Zhou Caipin; Wang Enyao; Yan Jiancheng; Liu Yong

    2004-01-01

    HL-2A device is the first divertor tokamak in China. One of its main subjects is to study the features of the divertor plasma. In the last campaign, the first divertor configuration has been achieved and sustained on the HL-2A tokamak. Here authors give a brief description about the HL-2A tokamak, diagnostics arrangements, and the equilibrium analysis results on divertor configuration. The main results of divertor experiments are also presented. (author)

  17. Summary of the 1982 small tokamak users meeting

    International Nuclear Information System (INIS)

    Sprott, J.C.

    1982-11-01

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

  18. Eddy currents in the Alcator Tokamak

    International Nuclear Information System (INIS)

    Schram, D.C.; Rem, J.

    1975-03-01

    A one-dimensional model of an aircore transformer has been developed through which it is possible to analyze the effect of eddy currents in the primary windings and of similar currents in the field coils for the toroidal magnetic field, on the time dependence of the current in a Tokamak experiment. The model is applied to the 'Alcator' Tokamak at MIT and its accuracy is tested by comparing analytical results for the harmonic behaviour of the transformer, with experimental data. The time-dependent behaviour of the plasma current for a constant plasma resistance shows that eddy currents in the primary windings will lead to a reduction of 8% of the current maximum. The eddy currents in the 'Bitter' coils are found to affect predominantly the initial current rise; they lead to a steepening of the current rise. Finally, the influence of the time dependence of the plasma resistance is investigated

  19. Microinstability theory in tokamaks: a review

    Energy Technology Data Exchange (ETDEWEB)

    Tang, W.M.

    1977-06-01

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

  20. Comparison between stellarator and tokamak divertor transport

    International Nuclear Information System (INIS)

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

    2010-11-01

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

  1. Assembly study for JT-60SA tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  2. Start of the international tokamak physics activity

    International Nuclear Information System (INIS)

    Campbell, D.

    2001-01-01

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

  3. Metal impurity release in diverted tokamak discharges

    International Nuclear Information System (INIS)

    Staudenmaier, G.; Wampler, W.R.

    1986-01-01

    Plasma materials interaction at the wall of the main plasma chamber of the divertor tokamak ASDEX was investigated by a combined probe, allowing simultaneous measurements of the erosion rate by neutral particles, and the flux and average energy of neutrals. The erosion was measured by collecting part of the released material on a carbon strip. Subsequent surface analysis was performed by electron induced x-ray analysis. Flux and energy of the impinging neutral particles were measured after each single discharge with an energy resolving carbon resistance probe. Such combined measurements yield the erosion yield being characteristic for the erosion process. Data for ohmic discharges in deuterium and helium are compared. It turns out that the carbon resistance probe is a simple but powerful means to study the metal impurity release from the tokamak walls by charge exchange neutrals

  4. Tokamak Fusion Core Experiment maintenance study

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  5. Boundary Plasma Turbulence Simulations for Tokamaks

    International Nuclear Information System (INIS)

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

    2008-05-01

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

  6. Runaway-ripple interaction in Tokamaks

    International Nuclear Information System (INIS)

    Laurent, L.; Rax, J.M.

    1989-08-01

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

  7. Neural net prediction of tokamak plasma disruptions

    International Nuclear Information System (INIS)

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

    1994-10-01

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

  8. Lower hybrid current drive in shaped tokamaks

    International Nuclear Information System (INIS)

    Kesner, J.

    1993-01-01

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

  9. The physics of tokamak start-up

    International Nuclear Information System (INIS)

    Mueller, D.

    2013-01-01

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

  10. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  11. The tokamak - an imperfect frame of refernce

    International Nuclear Information System (INIS)

    Schmitter, K.U.

    1981-03-01

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

  12. Tensor pressure tokamak equilibrium and stability

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, W.A.

    1981-03-01

    We investigate the equilibrium and magnetohydrodynamic (MHD) stability of tokamaks with tensor pressure and examine, in particular, the effects of anisotropies induced by neutral beam injection. Perpendicular and parallel beam pressure components are evaluated by taking moments of a distribution function obtained from the solution of a Fokker-Planck equation that models the injection of high-energy neutral beams into a tokamak. We numerically generate D-shaped beam-induced tensor pressure equilibria. A double adiabatic energy principle is derived from a modified version of the guiding center plasma energy principle. Finally, we apply the tensor pressure ballooning mode equation to computed equilibria that model experimentally determined ISX-B discharge profiles with high-power neutral beam injection. We predict that the plasma is unstable to flutelike modes in the central core of the discharge as a result of the pressure profile peakedness induced by the beams.

  13. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

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

    1980-10-01

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

  14. Microinstabilities in weak density gradient tokamak systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-04-01

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

  15. Microwave correllation reflectometry for tokamak CASTOR

    Czech Academy of Sciences Publication Activity Database

    Nanobashvili, S.; Žáček, František; Zajac, Jaromír

    2005-01-01

    Roč. 55, č. 6 (2005), s. 701-719 ISSN 0011-4626 R&D Projects: GA AV ČR IAA1043101 Grant - others:GA EU(EU) INTAS ´2001 1B-2056 Institutional research plan: CEZ:AV0Z20430508 Keywords : microwaves * tokamak * plasma * turbulence * reflectometry Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.360, year: 2005

  16. Confinement scaling and ignition in tokamaks

    International Nuclear Information System (INIS)

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

    1985-10-01

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

  17. Electron cyclotron emission from the PLT tokamak

    International Nuclear Information System (INIS)

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

    1977-07-01

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

  18. MHD stability of an almost circular tokamak

    International Nuclear Information System (INIS)

    Roy, A.

    1990-10-01

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

  19. Neoclassical tearing modes in a tokamak

    International Nuclear Information System (INIS)

    Hahm, T.S.

    1988-08-01

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

  20. Axisymmetric instability in a noncircular tokamak

    International Nuclear Information System (INIS)

    Lipschultz, B.

    1979-10-01

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

  1. Tokamak with liquid metal toroidal field coil

    International Nuclear Information System (INIS)

    Ohkawa, T.; Schaffer, M.J.

    1981-01-01

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

  2. Discrete Alfven waves in the TORTUS tokamak

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  3. Multiple mode model of tokamak transport

    International Nuclear Information System (INIS)

    Singer, C.E.; Ghanem, E.S.; Bateman, G.; Stotler, D.P.

    1989-07-01

    Theoretical models for radical transport of energy and particles in tokamaks due to drift waves, rippling modes, and resistive ballooning modes have been combined in a predictive transport code. The resulting unified model has been used to simulate low confinement mode (L-mode) energy confinement scalings. Dependence of global energy confinement on electron density for the resulting model is also described. 26 refs., 1 fig., 2 tabs

  4. Physics parameter space of tokamak ignition devices

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  5. Modeling tokamak discharges with current holes

    International Nuclear Information System (INIS)

    Jensen, T.H.

    2002-01-01

    Tokamaks with current holes [T.S. Taylor, et al., Bull. Am. Phys. Soc. 43 (1998) 1783; N.C. Hawkes, et al., Phys. Rev. Lett. 87 (2001) 115001; T. Fujita, et al., Phys. Rev. Lett. 87 (2001) 245001] are interesting, in part, because discharges with true current holes do not consume poloidal flux. The modeling of this Letter suggests that under steady-state conditions their currents may be driven by radial flow of plasma resulting from neutral beam injection

  6. User's manual of Tokamak Simulation Code

    International Nuclear Information System (INIS)

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

    1992-12-01

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

  7. Coherent structures in tokamak plasmas workshop: Proceedings

    International Nuclear Information System (INIS)

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

    1992-08-01

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

  8. 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#

  9. High beta plasmas in the PBX tokamak

    International Nuclear Information System (INIS)

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

    1986-04-01

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

  10. Multiple mode model of tokamak transport

    Energy Technology Data Exchange (ETDEWEB)

    Singer, C.E.; Ghanem, E.S.; Bateman, G.; Stotler, D.P.

    1989-07-01

    Theoretical models for radical transport of energy and particles in tokamaks due to drift waves, rippling modes, and resistive ballooning modes have been combined in a predictive transport code. The resulting unified model has been used to simulate low confinement mode (L-mode) energy confinement scalings. Dependence of global energy confinement on electron density for the resulting model is also described. 26 refs., 1 fig., 2 tabs.

  11. High beta plasmas in the PBX tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-04-01

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

  12. Small-scale tearing mode in tokamaks

    International Nuclear Information System (INIS)

    Ivanov, N.V.

    1983-01-01

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

  13. 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)

  14. Current drive for spherical tokamak plasmas

    International Nuclear Information System (INIS)

    Storer, R.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Azizov, E A

    2012-01-01

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

  16. Accessibility of high. beta. tokamak states

    Energy Technology Data Exchange (ETDEWEB)

    Hogan, J. T.

    1978-05-01

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

  17. Liquid tin limiter for FTU tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  18. 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

  19. 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

  20. On the density limit of Tokamaks

    International Nuclear Information System (INIS)

    Lehnert, B.

    1982-12-01

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

  1. Initial DEMO tokamak design configuration studies

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. The spheric tokamak programme at Culham

    International Nuclear Information System (INIS)

    Sykes, A.

    1999-01-01

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

  3. Experimental and theoretical basis for advanced tokamaks

    International Nuclear Information System (INIS)

    Chan, V.S.

    1994-09-01

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

  4. 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.)

  5. Electron cyclotron emission from tokamak plasmas

    International Nuclear Information System (INIS)

    Sillen, R.M.J.

    1986-01-01

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

  6. Deposit of thin films for Tokamaks conditioning

    International Nuclear Information System (INIS)

    Valencia A, R.

    2006-01-01

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

  7. Toroidal and poloidal momentum transport studies in Tokamaks

    DEFF Research Database (Denmark)

    Tala, T.; Andrew, Y.; Giroud, C.

    2007-01-01

    The present status of understanding of toroidal and poloidal momentum transport in tokamaks is presented in this paper. Similar energy confinement and momentum confinement times, i.e. τE/τφ ≈ 1 have been reported on several tokamaks. It is more important though, to study the local transport both ...

  8. 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)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem)

    2013-01-01

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

  11. Lower hybrid heating experiments in tokamaks: an overview

    International Nuclear Information System (INIS)

    Porkolab, M.

    1985-10-01

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

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

    Indian Academy of Sciences (India)

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

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

    International Nuclear Information System (INIS)

    Ignat, D.W.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1982-01-01

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

  15. 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

  16. Control strategy for plasma equilibrium in a tokamak

    International Nuclear Information System (INIS)

    Miskell, R.V.

    1975-08-01

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

  17. Toroidal and poloidal momentum transport studies in tokamaks

    DEFF Research Database (Denmark)

    Tala, T.; Crombé, K.; Vries, P.C. de

    2007-01-01

    The present status of understanding of toroidal and poloidal momentum transport in tokamaks is presented in this paper. Similar energy confinement and momentum confinement times, i.e. τE/τφ ≈ 1 have been reported on several tokamaks. It is more important though, to study the local transport both ...

  18. System assessment of helical reactors in comparison with tokamaks

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. The physics of an ignited tokamak

    International Nuclear Information System (INIS)

    Troyon, F.

    1990-10-01

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

  20. 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%

  1. Iron forbidden lines in tokamak discharges

    International Nuclear Information System (INIS)

    Suckewer, S.; Hinnov, E.

    1979-03-01

    Several spectrum lines from forbidden transitions in the ground configurations of highly ionized atoms have been observed in the PLT tokamak discharges. Such lines allow localized observations, in the high-temperature regions of the plasma, of ion-temperatures, plasma motions, and spatial distributions of ions. Measured absolute intensities of the forbidden lines have been compared with simultaneous observations of the ion resonance lines and with model calculations in order to deduce the mechanism of level populaions by means of electron collisions and radiative transitions

  2. Iron forbidden lines in tokamak discharges

    Energy Technology Data Exchange (ETDEWEB)

    Suckewer, S.; Hinnov, E.

    1979-03-01

    Several spectrum lines from forbidden transitions in the ground configurations of highly ionized atoms have been observed in the PLT tokamak discharges. Such lines allow localized observations, in the high-temperature regions of the plasma, of ion-temperatures, plasma motions, and spatial distributions of ions. Measured absolute intensities of the forbidden lines have been compared with simultaneous observations of the ion resonance lines and with model calculations in order to deduce the mechanism of level populaions by means of electron collisions and radiative transitions.

  3. 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

  4. Differential and Integral Models of TOKAMAK

    Directory of Open Access Journals (Sweden)

    Ivo Dolezel

    2004-01-01

    Full Text Available Modeling of 3D electromagnetic phenomena in TOKAMAK with typically distributed main and additional coils is not an easy business. Evaluated must be not only distribution of the magnetic field, but also forces acting in particular coils. Use of differential methods (such as FDM or FEM for this purpose may be complicated because of geometrical incommensurability of particular subregions in the investigated area or problems with the boundary conditions. That is why integral formulation of the problem may sometimes be an advantages. The theoretical analysis is illustrated on an example processed by both methods, whose results are compared and discussed.

  5. Present status of TCA/BR Tokamak

    International Nuclear Information System (INIS)

    Nascimento, I.C.; Galvao, R.M.O.; Tuszel, A.G.

    1997-01-01

    The TCA tokamak is being partially reconstructed and reassembled in the Plasma Laboratory of The University of Sao Paulo, and afterwards it will be named TCA/BR. The first discharges are expected by June/July of next year. The main scientific objectives envisaged for the machine are: Alfven wave heating and current drive, confinement improvement, disruptions and turbulence. In this paper we also describe: (i) the present status of the project; (ii) the diagnostic system; (iii) the control and data acquisition system; (iv) the RF system for the excitation of Alfven waves, that are being developed, and also the results of predictive transport simulations of its performance. (author)

  6. Viscosity in the edge of tokamak plasmas

    International Nuclear Information System (INIS)

    Stacey, W.M.

    1993-05-01

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

  7. Particle and heat transport in Tokamaks

    International Nuclear Information System (INIS)

    Chatelier, M.

    1984-01-01

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

  8. Alpha transport and blistering in tokamaks

    International Nuclear Information System (INIS)

    Bauer, W.; Wilson, K.L.; Bisson, C.L.; Haggmark, L.G.; Goldston, R.J.

    1978-12-01

    The particle flux and angular distribution of 3.5 MeV alpha particles impinging on the first wall from uncontained banana orbits in an axisymmetric tokamak reactor have been calculated. The resulting helium concentration profiles in the first wall can give rise to surface exfoliation under specified conditions. The major mitigating factor is the simultaneous surface recession due to sputtering by the D-T charge exchange neutral flux. For the parameters used in these calculations blistering in high sputtering rate materials such as Be is unlikely whereas in low sputtering rate materials such as Nb, He induced surface deformation is quite probable

  9. Modelling of neutron sawteeth in Tokamaks

    International Nuclear Information System (INIS)

    Anderson, D.; Hamnen, H.; Lisak, M.

    1990-01-01

    A model is developed to relate the drop in fusion neutron emission during sawtooth discharges in Tokamaks to the properties of the ion temperature and density sawteeth. In particular, the ion profile characteristics are shown to play an important role. The model determines the ion temperature profile exponent and the central ion temperature drop from the drop in neutron emission and the observed radius of inversion for the electron temperature. An extension is also made to line integrated neutron emission measurements as well as to neutron emission from neutral beam heated discharges where the dominating contribution to the neutron emission comes from beam-plasma reactions

  10. The spherical tokamak fusion power plant

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  11. 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. science direct.com/ science /article/pii/S0920379617300650

  12. 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

  13. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1997-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  14. Magnetic field structure of experimental high beta tokamak equilibria

    International Nuclear Information System (INIS)

    Deniz, A.V.

    1986-01-01

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

  15. Development of atomic beam probe for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Berta, M., E-mail: bertam@sze.hu [Széchenyi István University, EURATOM Association, Győr (Hungary); Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, EURATOM Association, Budapest (Hungary); Havlícek, J.; Háček, P. [Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics (Czech Republic)

    2013-11-15

    Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies.

  16. Electromagnetic simulations of tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Michael; Mishchenko, Alexey [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Assoziation, Wendelsteinstrasse 1, 17491 Greifswald (Germany)

    2014-07-01

    A practical fusion reactor will require a plasma β of around 5%. In this range Alfvenic effects become important. Since a practical reactor will also produce energetic alpha particles, the interaction between Alfvenic instabilities and fast ions is of particular interest. We have developed a fluid electron, kinetic ion hybrid model that can be used to study this problem. Compared to fully gyrokinetic electromagnetic codes, hybrid codes offer faster running times and greater flexibility, at the cost of reduced completeness. The model has been successfully verified against the worldwide ITPA Toroidal Alfven Eigenmode (TAE) benchmark, and the ideal MHD code CKA for the internal kink mode in a tokamak. Use of the model can now be turned toward cases of practical relevance. Current work focuses on simulating fishbones in a tokamak geometry, which may be of relevance to ITER, and producing the first non-perturbative self-consistent simulations of TAE in a stellarator, which may be of relevance both to Wendelstein 7-X and any future stellarator reactor. Preliminary results of these studies are presented.

  17. Alfven wave studies on a tokamak

    International Nuclear Information System (INIS)

    Kortbawi, D.

    1987-10-01

    The continuum modes of the shear Alfven resonance are studied on the Tokapole II device, a small tokamak operated in a four node poloidal divertor configuration. A variety of antenna designs and the efficiency with which they deliver energy to the resonant layer are discussed. The spatial structure of the driven waves is studied by means of magnetic probes inserted into the current channel. In an attempt to optimize the coupling of energy in to the resonant layer, the angle of antenna currents with respect to the equilibrium field, antenna size, and plasma-to-antenna distance are varied. The usefulness of Faraday shields, particle shields, and local limiters are investigated. Antennas should be well shielded, either a dense Faraday shield or particle shield being satisfactory. The antenna should be large and very near to the plasma. The wave magnetic fields measured show a spatial resonance, the position of which varies with the value of the equilibrium field and mass density. They are polarized perpendicular to the equilibrium field. A wave propagates radially in to the resonant surface where it is converted to the shear Alfven wave. The signal has a short risetime and does not propagate far toroidally. These points are all consistent with a strongly damped shear Alfven wave. Comparisons of this work to theoretical predictions and results from other tokamaks are made

  18. Theoretical scaling law for ohmically heated tokamaks

    International Nuclear Information System (INIS)

    Minardi, E.

    1981-06-01

    The electrostatic drift instability arising from the reduction of shear damping, due to toroidal effects, is assumed to be the basic source of the anomalous electron transport in tokamaks. The Maxwellian population of electrons constitutes a medium whose adiabatic nonlinear reaction to the instability (described in terms of an effective dielectric constant of the medium) determines the stationary electrostatic fluctuation level in marginally unstable situations. The existence of a random electrostatic potenial implies a fluctuating current of the Maxwellian electrons which creates a random magnetic field and a stocasticization of a magnetic configuration. The application of recent results allows the calculation of the realted radial electron transport. It is found that the confinement time under stationary ohmic conditions scales as n Tsub(i)sup( - 1/2) and is proportional roughly to the cube of the geometric dimenisions. Moreover, it is deduced that the loop voltage is approximateley the same for all tokamaks, irrespective of temperature and density and to a large extent, also of geometrical conditions. Thes results are characteristic of the ohmic stationary regime and can hardly be extrapolated to order heating regimes. (orig.)

  19. Sources of nonadiabaticity in tokamak turbulence

    International Nuclear Information System (INIS)

    Thyagaraja, A.; Haas, F.A.

    1993-01-01

    The two-fluid equations governing the nonlinear evolution and saturation of drift wave-like turbulence and transport in tokamaks under quasi-neutral conditions in periodic cylinder geometry are investigated. Using experiment as guide and employing appropriate orderings, two non-adiabaticity parameters, Υ es and Υ em are derived as functions of the reduced frequency ωa/v thi and wave number ρ i k r characteristic of the turbulent fluctuation spectrum. These parameters correspond respectively to the electrostatic limit and the general electromagnetic case. It is shown that they must be O(1) if significant particle and ion energy transport are to be expected from the turbulence. In other words, they are measures of the departure from neo-classical particle and ion energy transport due to the turbulence. These analytic results are complementary to, and serve as guidelines for, any future direct numerical simulations of the set of seven nonlinear partial differential equations which must be solved with suitable sources of particles, momentum and energy to determine the turbulence evolution and resultant saturated power spectra of density, pressure, electrostatic potential and magnetic field. The nonadiabaticity parameters discussed suggest possible qualitative explanations of the isotope effect and reduction of anomalous transport noted in H-mode tokamak discharges. (orig.)

  20. Theory of tokamak resistive fishbone modes

    International Nuclear Information System (INIS)

    Shi Bingren; Sui Guofang

    1995-12-01

    A special kind of internal kink mode, the fishbone, can be excited by the energetic particles in tokamak plasmas. Theoretical analyses of fishbone modes based on the ideal MHD framework have predicted that two branches of modes exists. One is the Chen-White branch with ω∼ω-bar dm , corresponding to a higher threshold in β h ; the other is the Coppis branch with ω∼ω *i , and a much lower threshold in β h . The latter mode would put a rather unfavourable restriction on heating efficiency and on plasma confinement. However. It is found that the resistivity effect is essential for this mode. In this paper, a new resistive fishbone mode analysis is carried out. In the (γ mhd ,β H ) space, the stability diagram shows complicate structure, the Coppis branch is replaced by a weakly unstable mode and there is no longer closed stable region. The growth rate of this mode varies with β h , its peak value is still very low compared to other internal modes. The implications of these results to future tokamak experiments are discussed. (8 figs.)

  1. Tokamak advanced pump limiter experiments and analysis

    International Nuclear Information System (INIS)

    Conn, R.W.

    1983-06-01

    Experiments with pump limiter modules on several operating tokamaks establish such limiters as efficient collectors of particles and has demonstrated the importance of ballistic scattering as predicted theoretically. Plasma interaction with recycling neutral gas appears to become important as the plasma density increases and the effective ionization mean free path within the module decreases. In limiters with particle collection but without active internal pumping, the neutral gas pressure is found to vary nonlinearly with the edge plasma density at the highest densities studies. Both experiments and theory indicate that the energy spectrum of gas atoms in the pump ducting is non-thermal, consistent with the results of Monte Carlo neutral atom transport calculations. The distribution of plasma power over the front surface of such modules has been measured and appears to be consistent with the predictions of simple theory. Initial results from the latest experiment on the ISX-B tokamak with an actively pumped limiter module demonstrates that the core plasma density can be controlled with a pump limiter and that the scrape-off layer plasma can partially screen the core plasma from gas injection. The results from module pump limiter experiments and from the theory and design analysis of advanced pump limiters for reactors are used to suggest the major features of a definitive, axisymmetric, toroidal belt pump limiter experiment

  2. Experimental investigations at the Soviet tokamaks

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  3. Overview on Chinese tokamak experimental progress

    International Nuclear Information System (INIS)

    Xie, J.K.; Li, J.; Liu, Y.; Wen, Y.Z.; Wang, L.

    2001-01-01

    Tokamak experiment research in China has made important progress. The main efforts subjected to quasi-steady state operation, LHCD, plasma heating with ICRF, IBW, NBI, ECRH, fueling with pellet and supersonic molecular beam, first wall conditioning technique. Plasma parameters in experiments were much improved, such as n e =8x10 19 m -3 , plasma pulse >10Sec. ICRF boronization and conditioning made Z eff close to unit. Steady state full LH wave current drive has been achieved for more than 3 seconds. LHCD ramp up and recharge have also been demonstrated. The Best η CD exp ∼0.5(1+0.085 exp(4.8(B T -1.45))n e I CD R p /P LH =10 19 m -2 A/W. Quasi steady state H-mode like plasma with density close to Greenwald limit was obtained by LHCD, in which energy confinement time was nearly 5 times longer than the Ohmic case. The synergy between IBW, pellet and LHCD was tested. Research on the mechanism of macro-turbulence has been extensively carried out experimentally. Ac operation of tokamak was successfully demonstrated. (author)

  4. Anomalous transport in the tokamak edge

    International Nuclear Information System (INIS)

    Vayakis, G.

    1991-04-01

    The tokamak edge has been studied with arrays of Langmuir and magnetic probes on the DITE and COMPASS-C devices. Measurements of plasma parameters such as density, temperature and radial magnetic field were taken in order to elucidate the character, effect on transport and origin of edge fluctuations. The tokamak edge is a strongly-turbulent environment, with large electrostatic fluctuation levels and broad spectra. The observations, including direct correlation measurements, are consistent with a picture in which the observed magnetic field fluctuations are driven by the perturbations in electrostatic parameters. The propagation characteristics of the turbulence, investigated using digital spectral techniques, appear to be dominated by the variation of the radial electric field, both in limiter and divertor plasmas. A shear layer is formed, associated in each case with the last closed flux surface. In the shear layer, the electrostatic wavenumber spectra are significantly broader. The predictions of a drift wave model (DDGDT) and of a family of models evolving from the rippling mode (RGDT group), are compared with experimental results. RGDT, augmented by impurity radiation effects, is shown to be the most reasonable candidate to explain the nature of the edge turbulence, only failing in its estimate of the wavenumber range. (Author)

  5. Modelling and control of a tokamak plasma

    International Nuclear Information System (INIS)

    Bremond, S.

    1995-01-01

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

  6. Sawtooth driven particle transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Nicolas, T.

    2013-01-01

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

  7. Thomson scattering on the PRETEXT Tokamak

    International Nuclear Information System (INIS)

    McCool, S.C.

    1982-03-01

    Ruby laser Thomson scattering was performed on the PRETEXT tokamak. A 10 Joule Q-switched laser and a 1 meter 10 channel polychromator were used to diagnose the electron temperature and density profiles in the PRETEXT plasma. These parameters were measured as a function of time and radial position on a shot to shot basis. The density measurement was calibrated by Rayleigh and Raman scattering and by comparison with data from a 4 mm microwave interferometer. Electron densities ranging from 1 x 10 12 cm -3 to 2 x 10 13 cm -3 and temperatures ranging from 3 eV to 400 eV were observed. Detailed measurements were made throughout the 40 ms discharge with particular emphasis on the current rise phase. The Thomson scattering data was used as input to a one dimensional magnetic diffusion code. This code modelled the evolution of the current density and safety factor profiles. The results of this analysis were compared with existing theories of tokamak current penetration. The growth of resitive MHD tearing modes was proposed as a likely explanation for the anomalously rapid current penetration observed in PRETEXT

  8. Physics evaluation of compact tokamak ignition experiments

    International Nuclear Information System (INIS)

    Uckan, N.A.; Houlberg, W.A.; Sheffield, J.

    1985-01-01

    At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/ 2 /q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Considering both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs

  9. Hydrogen recycle modeling and measurements in tokamaks

    International Nuclear Information System (INIS)

    Howe, H.C.

    1980-01-01

    A model for hydrogen recycling developed for use in a tokamak transport code is described and compared with measurements on ISX-B and DITE. The model includes kinetic reflection of charge-exchange neutrals from the wall and deposition, thermal diffusion, and desorption processes in the wall. In a tokamak with a limiter, the inferred recycle coefficient of 0.9-1.0 is due primarily to reflection (0.8-0.9) with the remainder (0.1-0.2) being due to desorption. Laboratory experiments supply much of the data for the model and several areas are discussed where additional data are needed, such as reflection from hydrogen-loaded walls at low (approx. equal to100 eV) energy. Simulation of ISX-B shows that the recently observed density decrease with neutral beam injection may be partially due to a decrease in recycling caused by hardening of the charge-exchange flux incident on the wall from the plasma. Modeling of isotopic exchange in DITE indicates the need for an ion-induced desorption process which responds on a timescale shorter than the wall thermal diffusion time. (orig.)

  10. Tokamak power systems studies at ANL

    International Nuclear Information System (INIS)

    Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

    1986-01-01

    A number of advances in plasma physics and engineering promise to greatly improve the reactor prospects of tokamaks. The following features, in particular, are examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 7 T; (c) low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields produced in the plasma. In addition to matching desirable high-beta equilibria, this method is capable of producing a large variety of new equilibria, many of which look attractive. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors

  11. 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

  12. Hard X-ray studies on the Castor tokamak

    International Nuclear Information System (INIS)

    Mlynar, J.

    1990-04-01

    The electron runaway processes in tokamaks are discussed with regard to hard X radiation measurements. The origin and confinement of runaway electrons, their bremsstrahlung spectra and the influence of lower hybrid current drive on the distribution of high-energy electrons are analyzed for the case of the Castor tokamak. The hard X-ray spectrometer designed for the Castor tokamak is also described and preliminary qualitative results of hard X-ray measurements are presented. The first series of integral measurements made it possible to map the azimuthal dependence of the hard X radiation

  13. Nonneutralized charge effects on tokamak edge magnetohydrodynamic stability

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  14. 3He functions in tokamak-pumped laser systems

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1986-10-01

    3 He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the 3 He(n,p)T reaction, and thereby excite gaseous lasants mixed with the 3 He while simultaneously breeding tritium. The total 3 He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak

  15. /sup 3/He functions in tokamak-pumped laser systems

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.

    1986-10-01

    /sup 3/He placed in an annular cell around a tokamak fusion generator can convert moderated fusion neutrons to energetic ions by the /sup 3/He(n,p)T reaction, and thereby excite gaseous lasants mixed with the /sup 3/He while simultaneously breeding tritium. The total /sup 3/He inventory is about 4 kg for large tokamak devices. Special configurations of toroidal-field magnets, neutron moderators and beryllium reflectors are required to permit nearly uniform neutron current into the laser cell with minimal attenuation. The annular laser radiation can be combined into a single output beam at the top of the tokamak.

  16. High pressure tokamaks. [Review of equilibrium and stability problems

    Energy Technology Data Exchange (ETDEWEB)

    Bateman, G.

    1978-05-01

    The successful development of the neutral beam injection method of heating tokamaks has opened up a new range of theoretically predicted phenomena to be explored. This article, intended for the nonspecialist, reviews the existing experimental observations and theoretical understanding of tokamak equilibrium and large scale stability. Then a survey is presented of the new phenomena, such as flux conserving sequences of equilibria and pressure-driven ballooning modes, that are expected to accompany the significantly enhanced plasma pressure to be produced in tokamaks now under construction.

  17. Transient electromagnetic analysis in tokamaks using TYPHOON code

    International Nuclear Information System (INIS)

    Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.

    1996-01-01

    The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)

  18. Analysis of EAST tokamak cryostat anti-seismic performance

    International Nuclear Information System (INIS)

    Chen Wei; Kong Xiaoling; Liu Sumei; Ni Xiaojun; Wang Zhongwei

    2014-01-01

    A 3-D finite element model for EAST tokamak cryostat is established by using ANSYS. On the basis of the modal analysis, the seismic response of the EAST tokamak cryostat structure is calculated according to an input of the design seismic response spectrum referring to code for seismic design of nuclear power plants. Calculation results show that EAST cryostat displacement and stress response is small under the action of earthquake. According to the standards, EAST tokamak cryostat structure under the action of design seismic can meet the requirements of anti-seismic design intensity, and ensure the anti-seismic safety of equipment. (authors)

  19. Equilibrium reconstruction in the TCA/Br tokamak; Reconstrucao do equilibrio no tokamak TCA/BR

    Energy Technology Data Exchange (ETDEWEB)

    Sa, Wanderley Pires de

    1996-12-31

    The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author) 68 refs., 31 figs., 16 tabs.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  1. 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.)

  2. Comments on density fluctuations and confinement in Tokamak

    International Nuclear Information System (INIS)

    Olivain, J.; Andreoletti, J.; Gervais, F.; Quemeneur, A.; Barkley, H.

    1987-04-01

    Anomalous transport imposes a serious limitation on Tokamak performance. Here, one possible cause of the transport ''anomaly'' is commented, the low frequency turbulence observed through density fluctuations. Recent progress gained on this issue are analysed

  3. Joint Czechoslovak-Soviet workshop on current drive in tokamaks

    International Nuclear Information System (INIS)

    1985-10-01

    At the Joint Czechoslovak-Soviet Workshop on Current Drive in Tokamaks, five papers dealing with issues of general interest were presented. In a theoretical paper by Klima and Pavlo a one-dimensional model of the lower-hybrid current drive is described and the results of its analysis are used in a numerical simulation using T-7 tokamak parameters. In the second theoretical paper by Vojtsekhovich, Parail and Pereverzev the influence of the LH wave spectrum on the efficiency of the current drive is studied. Two papers deal with a new microwave system designed for experiments on LHCD in the T-7 tokamak. In particular, the power spectra of new four-waveguide grills are computed. In the last paper the non-inductive start-up of the discharge in the T-7 tokamak by means of electron cyclotron waves is investigated. (J.U.)

  4. Tokamak reactor cost model based on STARFIRE/WILDCAT costing

    International Nuclear Information System (INIS)

    Evans, K. Jr.

    1983-03-01

    A cost model is presented which is useful for survey and comparative studies of tokamak reactors. The model is heavily based on STARFIRE and WILDCAT costing guidelines, philosophies, and procedures and reproduces the costing for these devices quite accurately

  5. A survey of radio frequency heating in tokamaks

    International Nuclear Information System (INIS)

    Bhatti, Z.R.

    1998-01-01

    A brief summary is given of the plasma physics of radio frequency heating in tokamaks. The general features common to all schemes are described. The three main methods, ion cyclotron electron cyclotron, and lower hybrid are also discussed. (author)

  6. Calculation of triton confinement and burn-up in tokamaks

    International Nuclear Information System (INIS)

    Anderson, D.; Battistoni, P.

    1987-01-01

    An analytical investigation is made of the confinement and subsequent burn-up of fusion produced tritons in a deuterium Tokamak plasma. Explicit approximations are obtained for the triton confinement factor, clearly displaying the scaling with physical parameters. The importance of pitch angle scattering losses during the triton slowing down is also estimated. A comparison with experiments and numerical calculations on the FT Tokamak slows good qualitative agreement. (authors)

  7. Parametric study of ohmic discharges in the TCA tokamak

    International Nuclear Information System (INIS)

    De Chambrier, A.; Collins, G.A.; Heym, A.; Hofmann, F.; Hollenstein, Ch.; Joye, B.; Keller, R.; Lietti, A.; Lister, J.B.; Moret, J.-M.; Nowak, S.; O'Rourke, J.; Pochelon, A.; Simm, W.

    1983-01-01

    The study of the energy confinement in a tokamak is an important aspect in the characterisation of its performance. The TCA tokamak has been in operation now for more than two years and the state of the machine and of its diagnostics have permitted such work to be performed. The authors describe the proper method for this type of approach and then present the results concerning the energy confinement of the electrons and ions. (Auth./G.T.H.)

  8. Design of a microwave calorimeter for the microwave tokamak experiment

    International Nuclear Information System (INIS)

    Marinak, M.

    1988-01-01

    The initial design of a microwave calorimeter for the Microwave Tokamak Experiment is presented. The design is optimized to measure the refraction and absorption of millimeter rf microwaves as they traverse the toroidal plasma of the Alcator C tokamak. Techniques utilized can be adapted for use in measuring high intensity pulsed output from a microwave device in an environment of ultra high vacuum, intense fields of ionizing and non-ionizing radiation and intense magnetic fields. 16 refs

  9. Operating tokamaks with steady-state toroidal current

    International Nuclear Information System (INIS)

    Fisch, N.J.

    1981-04-01

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

  10. Automated Fault Detection for DIII-D Tokamak Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Walker, M.L.; Scoville, J.T.; Johnson, R.D.; Hyatt, A.W.; Lee, J.

    1999-11-01

    An automated fault detection software system has been developed and was used during 1999 DIII-D plasma operations. The Fault Identification and Communication System (FICS) executes automatically after every plasma discharge to check dozens of subsystems for proper operation and communicates the test results to the tokamak operator. This system is now used routinely during DIII-D operations and has led to an increase in tokamak productivity.

  11. Recycling in gettered and diverted discharges in DITE tokamak

    International Nuclear Information System (INIS)

    Fielding, S.J.; McCracken, G.M.; Stott, P.E.

    1978-01-01

    A model of recycling in tokamak is described which considers the plasma to consist of three interacting components: ions, fast neutrals and slow neutrals. The model describes the behaviour, during a discharge of the total population of each of these components, together with the fourth component, neutrals trapped in the wall. The model is applied to DITE tokamak and its predictions are compared with data obtained from D/H recycling experiments, in standard, gettered and diverted discharges. (Auth.)

  12. An emerging understanding of H-mode discharges in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Groebner, R.J.

    1992-12-01

    A remarkable degree of consistency of experimental results from tokamaks throughout the world has developed with regard to the phenomenology of the transition from L-mode to H-mode confinement in tokamaks. The transition is initiated in a narrow layer at the plasma periphery where density fluctuations are suppressed and steep gradients of temperature and density form in a region with large first and second radial derivatives in the [upsilon][sub E][sup [yields

  13. Fast wave current drive in reactor scale tokamak plasmas

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  14. Radial propagation of microturbulence in tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Laurent, L.; Roubin, J.P.; Samain, A.

    1992-01-01

    Energy confinement time in tokamaks exhibits a clear dependence on global plasma parameters. This is not the case for transport coefficients; their dependence on local plasma parameters cannot be precisely established. The aim of the present paper is to give a possible explanation of this behaviour; turbulence propagates radially because of departure from cylindrical geometry. This implies that the turbulence level at a given point and hence transport coefficients are not only functions of local plasma parameters. A quantitative estimate of the propagation velocity is derived from a Lagrangian formalism. Two cases are considered: the effect of toroidicity and the effect of non linear mode-mode coupling. The consequences of this model are discussed. This process does not depend on the type of instability. For the sake of simplicity only electrostatic perturbations are considered

  15. Real time analysis of tokamak discharge parameters

    International Nuclear Information System (INIS)

    Ferron, J.R.; Strait, E.J.

    1992-03-01

    The techniques used in implementing two applications of real time analysis of data from the DIII-D tokamak are described. These tasks, which are demanding in both the speed of data acquisition and the speed of computation, execute on hardware capable of acquiring 40 million data samples per second and executing 80 million floating point operations per second. In the first case, a feedback control algorithm executing at a 10 kHz cycle frequency is used to specify the current in the poloidal field coils in order to control the discharge shape. In the second, fast Fourier transforms of Mirnov probe data are used to find the amplitude and frequency of each of eight toroidal mode numbers as a function of time during the discharge. Data sampled continuously at 500 kHz are used to produce results at 2 msec intervals

  16. Advantages of iron core in a tokamak

    International Nuclear Information System (INIS)

    Bettis, E.S.; Ballou, J.K.; Becraft, W.R.; Peng, Y.K.M.; Watts, H.L.

    1977-01-01

    A quantitative comparison of the iron core vs air core concepts was carried out on a preliminary basis by using a representative tokamak reactor design with the following self-consistent reference parameters. In the area of plasma engineering, poloidal field and MHD equilibrium considerations with an unsaturated iron core is discussed. The question of proper poloidal field coils to maintain D-shaped plasmas of relatively high anti β (7%) with a saturated iron core is also discussed. Estimates of the required iron core size, volt seconds, magnetic flux and its influence on force loading on the superconducting toroidal field coils are shown. Conceptual designs of the mechanical structure of an iron core device are presented. Favorable impacts on the OH power supply cost and complexity are indicated

  17. Safety factor profile control in a tokamak

    CERN Document Server

    Bribiesca Argomedo, Federico; Prieur, Christophe

    2014-01-01

    Control of the Safety Factor Profile in a Tokamak uses Lyapunov techniques to address a challenging problem for which even the simplest physically relevant models are represented by nonlinear, time-dependent, partial differential equations (PDEs). This is because of the  spatiotemporal dynamics of transport phenomena (magnetic flux, heat, densities, etc.) in the anisotropic plasma medium. Robustness considerations are ubiquitous in the analysis and control design since direct measurements on the magnetic flux are impossible (its estimation relies on virtual sensors) and large uncertainties remain in the coupling between the plasma particles and the radio-frequency waves (distributed inputs). The Brief begins with a presentation of the reference dynamical model and continues by developing a Lyapunov function for the discretized system (in a polytopic linear-parameter-varying formulation). The limitations of this finite-dimensional approach motivate new developments in the infinite-dimensional framework. The t...

  18. 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.)

  19. Magnetic measurements on the TCV tokamak

    International Nuclear Information System (INIS)

    Moret, J.M.; Buehlmann, F.; Fasel, D.; Hofmann, F.; Tonetti, G.

    1996-12-01

    The TCV Tokamak was designed to create a large variety of plasma shapes. Such a large flexibility requires high precision magnetic measurements with a good spatial coverage. This paper gives a detailed description of the magnetic sensor geometry, fabrication, calibration, the associated electronics and the diagnostic operation and monitoring. A substantial effort has been made to quantify the precision in the measurements and a novel method has been developed to derive corrections in the sensor position and calibration which optimise the consistency of the entire measurement set. Accuracy of 0.5 mWb in the poloidal flux and 1 mT in the magnetic field with a position error of a few mm have been achieved. (author) figs., tabs., refs

  20. Simulation of impurity transport in tokamaks, 1

    International Nuclear Information System (INIS)

    Amano, T.; Mizuno, J.; Kako, M.

    1982-11-01

    A computer code to simulate impurity transport in tokamaks are described. The code solves the coupled rate and diffusion equations for a set of plasma ions, hydrogen isotopes plus several charge states of one or more impurity elements. Neoclassical transport for all ion species including both density gradient and temperature gradient effects is used. Impurity ions and plasma ions can be either in Pfirsch-Schluter or plateau-banana regime. Anomalous transport is also considered. Several models are used for atomic rates. The source of impurity is calculated from the sputtering of limiter and wall. The rate and diffusion equations are solved by Cranck-Nicholson's implicit scheme. The Crank-Nicholson's method is compared with more accurate Gear's method and a fairly good agreement is found between the two methods. (author)

  1. Ignition probabilities for Compact Ignition Tokamak designs

    International Nuclear Information System (INIS)

    Stotler, D.P.; Goldston, R.J.

    1989-09-01

    A global power balance code employing Monte Carlo techniques had been developed to study the ''probability of ignition'' and has been applied to several different configurations of the Compact Ignition Tokamak (CIT). Probability distributions for the critical physics parameters in the code were estimated using existing experimental data. This included a statistical evaluation of the uncertainty in extrapolating the energy confinement time. A substantial probability of ignition is predicted for CIT if peaked density profiles can be achieved or if one of the two higher plasma current configurations is employed. In other cases, values of the energy multiplication factor Q of order 10 are generally obtained. The Ignitor-U and ARIES designs are also examined briefly. Comparisons of our empirically based confinement assumptions with two theory-based transport models yield conflicting results. 41 refs., 11 figs

  2. Runaway electron generation in tokamak disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Helander, P. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Andersson, F.; Fulop, T.; Smith, T.H.; Anderson, D.; Lisak, M. [Chalmers Univ. of Technology, Dept. of Electromagnetics, Goteborg (Sweden); Eriksson, L.G. [Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    2004-07-01

    The time evolution of the plasma current during a tokamak disruption is calculated by solving the equations for runaway electron production simultaneously with the induction equation for the toroidal electric field. The resistive diffusion time in a post-disruption plasma is typically comparable to the runaway avalanche growth time. Accordingly, the toroidal electric field induced after the thermal quench of a disruption diffuses radially through the plasma at the same time as it accelerates runaway electrons, which in turn back-react on the electric field. When these processes are accounted for in a self-consistent way, it is found that (1) the efficiency and time scale of runaway generation agrees with JET experiments; (2) the runaway current profile typically becomes more peaked than the pre-disruption current profile; and (3) can easily become radially in the shape of filaments. It is also shown that higher runaway electron generation is expected if the thermal quench is sufficiently fast. (authors)

  3. Transport Bifurcation in a Rotating Tokamak Plasma

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  4. Energetics of turbulent transport processes in tokamaks

    International Nuclear Information System (INIS)

    Haas, F.A.; Thyagaraja, A.

    1987-01-01

    The effect of electromagnetic turbulence on electrons and ions under Tokamak conditions is considered using a kinetic description. Taking the magnetic fluctuation spectrum as given, the density fluctuation spectrum is self-consistently calculated taking account of quasi-neutrality. The calculation is valid for arbitrary collisionality and appropriate to low frequencies typical of experiment. In addition to the usual enhancement of the radial electron energy transport, it is found that the turbulent fluctuations can heat the plasma at rates comparable to ordinary ohmic heating under well-defined conditions. Interestingly, electromagnetic turbulence appears to imply only an insignificant correction to the toroidal resistance of the plasma as estimated from Spitzer resistivity. The scalings of anomalous transport, fluctuations and heating with temperature and plasma volume are investigated. The assumption that the magnetic fluctuation spectrum of the turbulence is invariant under a wide range of conditions is shown to result in interesting consequences for JET-like plasmas. (author)

  5. Sliding Mode Control of a Tokamak Transformer

    International Nuclear Information System (INIS)

    Romero, J. A.; Coda, S.; Felici, F.; Moret, J. M.; Paley, J.; Sevillano, G.; Garrido, I.; Le, H. B.

    2012-01-01

    A novel inductive control system for a tokamak transformer is described. The system uses the flux change provided by the transformer primary coil to control the electric current and the internal inductance of the secondary plasma circuit load. The internal inductance control is used to regulate the slow flux penetration in the highly conductive plasma due to the skin effect, providing first-order control over the shape of the plasma current density profile. Inferred loop voltages at specific locations inside the plasma are included in a state feedback structure to improve controller performance. Experimental tests have shown that the plasma internal inductance can be controlled inductively for a whole pulse starting just 30ms after plasma breakdown. The details of the control system design are presented, including the transformer model, observer algorithms and controller design. (Author) 67 refs.

  6. Instrumentation and controls of an ignited tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Becraft, W.R.; Golzy, J.; Houlberg, W.A.; Kukielka, C.A.; Onega R.J.; Raju, G.V.S.; Stone, R.S.

    1980-10-01

    The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented.

  7. Sliding Mode Control of a Tokamak Transformer

    Energy Technology Data Exchange (ETDEWEB)

    Romero, J. A.; Coda, S.; Felici, F.; Moret, J. M.; Paley, J.; Sevillano, G.; Garrido, I.; Le, H. B.

    2012-06-08

    A novel inductive control system for a tokamak transformer is described. The system uses the flux change provided by the transformer primary coil to control the electric current and the internal inductance of the secondary plasma circuit load. The internal inductance control is used to regulate the slow flux penetration in the highly conductive plasma due to the skin effect, providing first-order control over the shape of the plasma current density profile. Inferred loop voltages at specific locations inside the plasma are included in a state feedback structure to improve controller performance. Experimental tests have shown that the plasma internal inductance can be controlled inductively for a whole pulse starting just 30ms after plasma breakdown. The details of the control system design are presented, including the transformer model, observer algorithms and controller design. (Author) 67 refs.

  8. Coupling of tearing modes in tokamaks

    International Nuclear Information System (INIS)

    Finn, J.M.

    1977-01-01

    The simultaneous presence of tearing modes of different helical pitches leads to the destruction of magnetic surfaces, which has been suggested as the mechanism leading to the onset of the disruptive instability in tokamaks. For current profiles in which the m = 2 mode is unstable, but the m = 3 is stable, the coupling of the m = 3 to the m = 2 through the poloidal variation of the toroidal field can drive the m = 3 amplitude psi 3 to order psi 2 times the inverse aspect ratio. Detailed calculations, both analytical and numerical, have been performed for two models for the equilibrium and m = 2 mode structure. A slab model and incompressible m = 3 perturbations are assumed. The m = 3 amplitude increases with shear, up to a point, showing that as the current channel shrinks, overlap of resonances becomes more likely. The results also apply qualitatively to other m, m +- 1 interactions

  9. Auxiliary radiofrequency heating of tokamaks, Task 3

    International Nuclear Information System (INIS)

    Scharer, J.E.

    1991-07-01

    The research performed under this grant during the past three years has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling and heating issues: efficient coupling during the L- to H-mode transition by analysis and computer simulation of ICRF antennas edge plasma profiles; analysis of both dielectric-filled waveguide and coil ICRF antenna coupling to plasma edge profiles; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results; ICRF full-wave field solutions, power conservation and heating analyses; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report. 15 refs

  10. Tokamak transport based on the Braginskii model

    International Nuclear Information System (INIS)

    Weitzner, H.

    1987-01-01

    The two-fluid model of Braginskii is applied to the case of a moderately large tokamak. By estimation of the order of magnitude of the various effects and omission of small terms a somewhat simpler reduced two-fluid Braginskii model is obtained. The model applies on a time scale of order τ e m i /m e , where τ e is the electron-electron collision time, and energy confinement time is of this order. With electron and ion flow velocities no larger than is necessary to obtain the correct equilibrium currents, classical parallel viscosity becomes a dominant dissipative mechanism. The model allows for the slow evolution of equilibrium states. The equilibria, which include static, ideal magnetohydrodynamic equilibria as a special case, are described. Generally the number density, electrostatic potential, and flows are not constant on a flux surface. The procedure for determination of the slow evolution of the equilibrium is sketched. (orig.)

  11. Power oscillator in the Tokamaks training

    International Nuclear Information System (INIS)

    Valencia A, R.

    1994-01-01

    This work reports the results obtained from the cleaning of the Novillo Tokamak Chamber, using an A.F. Taylor Discharge Cleaning (TDC) in H 2 with a power oscillator of 20 k W and 17.5 k Hz. The plasma temperature in the discharge was of one electron-volt (Te ≅ 1 eV) with a moderate electron density n e ≅ 4 x 10 11 cm -3 . This discharge cleaning was found helpful in the removal of C and O via the formation of pumping compounds such as CH 4 and H 2 O. A residual gas analyzer was used to monitor the partial pressure of these and other compounds, indicating removal rates as high as two monolayers/hour at the beginning of the discharge. A value of Z eff = 3 was estimated for a discharge of 7 k A after conditioning. (Author)

  12. Maintenance considerations of the Starfire Commercial Tokamak

    International Nuclear Information System (INIS)

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

    1979-01-01

    An initial reactor design has been developed for the Starfire Tokamak reactor that incorporates maintenance features to permit reactor and plant operation with a 75% availability. All components of the reactor can be replaced utilizing highly automated remote maintenance techniques. Provisions for contact maintenance are included but would only be utilized on a contingency basis since it is believed that allowable dose rates will be decreased in future years. A modular design approach is used for the reactor and auxiliary subsystems to permit efficient use of remote maintenance. The modular approach minimizes the number of different maintenance operations required and can result in use of simple tasks such as push, pull, turn, etc., maneuvers for module removal. Fault isolation is provided for each replaceable module

  13. Equilibrium vertical field in the TBR Tokamak

    International Nuclear Information System (INIS)

    Ueta, A.Y.

    1985-01-01

    An experimental study on the influence of the vertical magnetic field of the TBR tokamak on the stability and equilibrium of plasma column, was done. Magnetic pick-up coils were built to measure plasma current and position, together with active networks, necessary fo the electronic processing of signals. Some measurements were on the space configuration of the vertical field, and on the influence due to the toroidal vessel. From the data obtained it was possible to discuss the influence of the currents induced on the vessel surface, on plasma equilibrium. Theoretical and experimental results of the vertica field, as a function of plasma current were compared, and allowed an evaluation of the plasma kinetic pressure and temperature. (Author) [pt

  14. Compact tokamak reactors part 2 (numerical results)

    International Nuclear Information System (INIS)

    Wiley, J.C.; Wootton, A.J.; Ross, D.W.

    1996-01-01

    The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced

  15. Stability of tearing modes in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-02-01

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

  16. Toroidal microinstability studies of high temperature tokamaks

    International Nuclear Information System (INIS)

    Rewoldt, G.; Tang, W.M.

    1989-07-01

    Results from comprehensive kinetic microinstability calculations are presented showing the effects of toroidicity on the ion temperature gradient mode and its relationship to the trapped-electron mode in high-temperature tokamak plasmas. The corresponding particle and energy fluxes have also been computed. It is found that, although drift-type microinstabilities persist over a wide range of values of the ion temperature gradient parameter η i ≡ (dlnT i /dr)/(dlnn i /dr), the characteristic features of the dominant mode are those of the η i -type instability when η i > η ic ∼1.2 to 1.4 and of the trapped-electron mode when η i ic . 16 refs., 7 figs

  17. Recycling measurements in the Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Rowan, W.L.; Klepper, C.C.; Patterson, D.M.; Richards, B.

    1987-01-01

    Recycling was investigated as a function of density, plasma current, and toroidal field in the Texas Experimental Tokamak, an ohmically heated, edge-fueled device with a full aperture poloidal limiter. Estimates for the recycling contribution to the plasma particle inventory were inferred from a particle accounting which employed measurements of the number of ionization events in the plasma and in the scrape-off layer, the total number of electrons in the plasma, and the neutral gas influx from the gas feed. Isotopic changeover experiments were used to further characterize recycling and to derive the contribution to the plasma particle inventory from the neutrals held both in the wall bulk and on the wall surface. A model was used to estimate the number of neutrals trapped in the wall as well as to examine the interactions of the major plasma participants in recycling - ions, fast neutrals, and slow neutrals. (orig.)

  18. Compact tokamak reactors part 2 (numerical results)

    Energy Technology Data Exchange (ETDEWEB)

    Wiley, J.C.; Wootton, A.J.; Ross, D.W.

    1996-10-21

    The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced.

  19. Real time analysis of tokamak discharge parameters

    International Nuclear Information System (INIS)

    Ferron, J.R.; Strait, E.J.

    1992-01-01

    The techniques used in implementing two applications of real time digital analysis of data from the DIII-D tokamak are described. These tasks, which are demanding in both the speed of data acquisition and the speed of computation, execute on hardware capable of acquiring 40 million data samples per second and executing 80 million floating point operations per second. In the first case, a feedback control algorithm executing at a 10 kHz cycle frequency is used to specify the current in the poloidal field coils in order to control the discharge shape. In the second, fast Fourier transforms of Mirnov probe data are used to find the amplitude and frequency of each of eight toroidal mode numbers as a function of time during the discharge. Data sampled continuously at 500 kHz are used to produce results at 2 ms intervals

  20. Ion cyclotron emission in tokamak plasmas

    International Nuclear Information System (INIS)

    Fraboulet, D.

    1996-01-01

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

  1. Plasma diagnostics for the compact ignition tokamak

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  2. Safety in the ARIES Tokamak Design Study

    International Nuclear Information System (INIS)

    Herring, J.S.; Wong, C.P.-C.; Cheng, E.T.; Grotz, S.

    1989-01-01

    Safety is one of the primary goals of the ARIES Tokamak Design Study. Public safety goals are the achievement passive safety which is demonstrable in tests that could precede operation and the assurance that releases from accidents be passively limited such that no evacuation plan in necessary. Strategies for safety of the plant investment are factory fabrication, short construction times and a design such that no off-normal operational transient results in damage which could not be repaired in routine maintenance. ARIES-I, the first of three 'visions' of potential tokamak reactors, will use He at 5 MPa as a blanket coolant and SiC/composite ceramic for the first wall and blanket materials. Both the coolant and the structural material were chosen for their low activation, both in the short term after accidents and for long term waste management. The breeder, Li 4 SiO 4 , was also chosen for low activation. Contemporary plasma physics and aggressive technology are used in ARIES-I, which results in very high toroidal fields (24 T maximum at the coil). The stored TF energy will be about 130 GJ. A central concern is the safe discharge of this stored energy under electrical fault conditions and prevention of a failure in the magnet set from propagating into systems containing radioactive inventories. The TF coil system consists of 16 coils, each containing two separate windings powered by two independent power supplies. Arcs and shorts between the two power supply systems and across individual windings have been modeled. In addition, delay or failure in circuit breaker opening has been modeled. The safety impacts of LOCA, LOFA and disruptive events have also been evaluated. 8 refs., 4 figs., 7 tabs

  3. Physics design of the HL-1M tokamak

    International Nuclear Information System (INIS)

    Gao Qingdi; Shi Bingren; Liu Yukui; Zhang Jinhua; Xue Siwen; Li Fangzhu

    1999-08-01

    Presented is the physics design of the HL-1M tokamak, which is a machine upgraded from the HL-1 tokamak. Based upon the intensive investigations on the controlled nuclear fusion research in the world, the direction for modifying the HL-1 tokamak was determined, i.e. reconstructing the vacuum chamber without the thick copper shell which is used as an outer vacuum vessel in HL-1, reforming the poloidal magnetic field system and upgrading the power supply so as to be suitable for performing experimental study on high power auxiliary heating and non-inductive current drive. The main physics objectives of HL-1M is to carry out investigations on MWs power auxiliary heating and current drive with lower hybrid wave. Besides this, the other physics objectives are as follows: to perform further experimental study on the ohmic heating plasma with higher parameters so that a database for extrapolating to a larger tokamak device could be obtained, and to accumulate experiences for the construction of next tokamak device, HL-2. By using the extrapolation of the HL 1 experiment results, the tokamak scaling law and numerical computation, the physics parameters of ohmic heating and auxiliary heating plasmas are designed in some details

  4. Resistive edge mode instability in stellarator and tokamak geometries

    Science.gov (United States)

    Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

    2008-09-01

    Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

  5. Conceptual design of a commercial tokamak reactor using resistive magnets

    International Nuclear Information System (INIS)

    LeClaire, R.J. Jr.

    1988-01-01

    The future of the tokamak approach to controlled thermonuclear fusion depends in part on its potential as a commercial electricity-producing device. This potential is continually being evaluated in the fusion community using parametric, system, and conceptual studies of various approaches to improving tokamak reactor design. The potential of tokamaks using resistive magnets as commercial electricity-producing reactors is explored. Parametric studies have been performed to examine the major trade-offs of the system and to identify the most promising configurations for a tokamak using resistive magnets. In addition, a number of engineering issues have been examined including magnet design, blanket/first-wall design, and maintenance. The study indicates that attractive design space does exist and presents a conceptual design for the Resistive Magnet Commercial Tokamak Reactor (RCTR). No issue has been identified, including recirculating power, that would make the overall cost of electricity of RCTR significantly different from that of a comparably sized superconducting tokamak. However, RCTR may have reliability and maintenance advantages over commercial superconducting magnet devices

  6. MHD stability limits in the TCV Tokamak

    International Nuclear Information System (INIS)

    Reimerdes, H.

    2001-07-01

    Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation κ and triangularity δ, with high κ, and low δ leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The observed decrease

  7. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-07-01

    Magnetohydrodynamic (MHD) instabilities can limit the performance and degrade the confinement of tokamak plasmas. The Tokamak a Configuration Variable (TCV), unique for its capability to produce a variety of poloidal plasma shapes, has been used to analyse various instabilities and compare their behaviour with theoretical predictions. These instabilities are perturbations of the magnetic field, which usually extend to the plasma edge where they can be detected with magnetic pick-up coils as magnetic fluctuations. A spatially dense set of magnetic probes, installed inside the TCV vacuum vessel, allows for a fast observation of these fluctuations. The structure and temporal evolution of coherent modes is extracted using several numerical methods. In addition to the setup of the magnetic diagnostic and the implementation of analysis methods, the subject matter of this thesis focuses on four instabilities, which impose local and global stability limits. All of these instabilities are relevant for the operation of a fusion reactor and a profound understanding of their behaviour is required in order to optimise the performance of such a reactor. Sawteeth, which are central relaxation oscillations common to most standard tokamak scenarios, have a significant effect on central plasma parameters. In TCV, systematic scans of the plasma shape have revealed a strong dependence of their behaviour on elongation {kappa} and triangularity {delta}, with high {kappa}, and low {delta} leading to shorter sawteeth with smaller crashes. This shape dependence is increased by applying central electron cyclotron heating. The response to additional heating power is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and consequently, a faster increase of the central pressure shortens the sawteeth, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The

  8. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  11. Superconducting magnets and cryogenics for the steady state superconducting tokamak SST-1

    International Nuclear Information System (INIS)

    Saxena, Y.C.

    2000-01-01

    SST-1 is a steady state superconducting tokamak for studying the physics of the plasma processes in tokamak under steady state conditions and to learn technologies related to the steady state operation of the tokamak. SST-1 will have superconducting magnets made from NbTi based conductors operating at 4.5 K temperature. The design of the superconducting magnets and the cryogenic system of SST-1 tokamak are described. (author)

  12. ECRH and electron heat transport in tokamaks

    International Nuclear Information System (INIS)

    Zou, X.L.; Giruzzi, G.; Dumont, R.J.

    2003-01-01

    It has been observed during the ECRH experiments in tokamaks that the shape of the electron temperature profile in stationary regimes is not very sensitive to the ECRH power deposition i.e. the temperature profile remains peaked at the center even though the ECRH power deposition is off-axis. Various models have been invoked for the interpretation of this profile resilience phenomenon: the inward heat pinch, the critical temperature gradient, the Self-Organized Criticality, etc. Except the pinch effect, all of these models need a specific form of the diffusivity in the heat transport equation. In this work, our approach is to solve a simplified time-dependent heat transport equation analytically in cylindrical geometry. The features of this analytical solution are analyzed, in particular the relationship between the temperature profile resilience and the Eigenmode of the physical system with respect to the heat transport phenomenon. Finally, applications of this analytical solution for the determination of the transport coefficient and the polarization of the EC waves are presented. It has been shown that the solution of the simplified transport equation in a finite cylinder is a Fourier-Bessel series. This series represents in fact a decomposition of the heat source in Eigenmode, which are characterized by the Bessel functions of order 0. The physical interpretation of the Eigenmodes is the following: when the heat source is given by a Bessel function of order 0, the temperature profile has exactly the same form as the source at every time. At the beginning of the power injection, the effectiveness of the temperature response is the same for each Eigenmode, and the response in temperature, having the same form as the source, is local. Conversely, in the later phase of the evolution, the effectiveness of the temperature response for each Eigenmode is different: the higher the order, the lower the effectiveness. In this case the response in temperature appears as

  13. Development of 3D ferromagnetic model of tokamak core with strong toroidal asymmetry

    DEFF Research Database (Denmark)

    Markovič, Tomáš; Gryaznevich, Mikhail; Ďuran, Ivan

    2015-01-01

    Fully 3D model of strongly asymmetric tokamak core, based on boundary integral method approach (i.e. characterization of ferromagnet by its surface) is presented. The model is benchmarked on measurements on tokamak GOLEM, as well as compared to 2D axisymmetric core equivalent for this tokamak, pr...

  14. Characterization of the Tokamak Novillo in cleaning regime

    International Nuclear Information System (INIS)

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

    1992-02-01

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

  15. Power supplies and quench protection for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Neumeyer, C.L.

    1994-01-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). First plasma is scheduled for the year 2000. TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This is a new feature which requires not only a departure from the traditional tokamak power supply schemes but also that ultra-reliable quench protection devices be used to rapidly discharge the stored energy from the magnets in the event of a quench. This paper describes the plan and basis for the adaptation and augmentation of the PPPL/TFTR power system facilities to supply TPX. Following a description of the basic operational requirements, four major areas are addressed, namely the AC power system, the TF power supply, the PF power supply, and quench protection for the TF and PF systems

  16. Total magnetic reconnection during a tokamak major disruption

    International Nuclear Information System (INIS)

    Goetz, J.A.

    1990-09-01

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

  17. Power supplies and quench protection for the Tokamak Physics Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Neumeyer, C.L. [Raytheon Engineers & Constructors, Princeton, NJ (United States). EBASCO Div.

    1994-07-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). First plasma is scheduled for the year 2000. TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This is a new feature which requires not only a departure from the traditional tokamak power supply schemes but also that ultra-reliable quench protection devices be used to rapidly discharge the stored energy from the magnets in the event of a quench. This paper describes the plan and basis for the adaptation and augmentation of the PPPL/TFTR power system facilities to supply TPX. Following a description of the basic operational requirements, four major areas are addressed, namely the AC power system, the TF power supply, the PF power supply, and quench protection for the TF and PF systems.

  18. Multiscale coherent structures in tokamak plasma turbulence

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  19. Ionization balance in EBIT and tokamak plasmas

    Science.gov (United States)

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

    2001-01-01

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

  20. Tokamak blanket design study, final report

    International Nuclear Information System (INIS)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m 2 and a particle heat flux of 1 MW/m 2 . Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma

  1. 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

  2. Demonstration tokamak-power-plant study (DEMO)

    International Nuclear Information System (INIS)

    1982-09-01

    A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material

  3. Plasma startup patterns in tokamak reactors

    International Nuclear Information System (INIS)

    Maki, Koichi; Tone, Tatsuzo.

    1983-01-01

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

  4. Exploration of turbulent optimization in stellarators & tokamaks

    Science.gov (United States)

    Mynick, H.; Pomphrey, N.; Xanthopoulos, P.; Lucia, M.

    2012-03-01

    A methodfootnotetextH.E. Mynick, N. Pomphrey, P. Xanthopoulos, Phys. Rev. Letters, 105, 095004 (2010).^,footnotetextH.E. Mynick, N. Pomphrey, P. Xanthopoulos, Phys. Plasmas, 18, 056101 (2011). recently developed for evolving toroidal configurations to ones with reduced turbulent transport, using the STELLOPT optimization codes and the GENE gyrokinetic code, is being applied and extended. The growing body of results has found that the effectiveness of the current proxy measure Qprox used by STELLOPT to estimate transport levels depends on the class of toroidal device considered. The present proxy works well for quasi-axisymmetric stellarators and tokamaks, modestly for quasi-helically symmetric designs, but not for the W7X quasi-omnigenous/quasi-isodynamic design. We are exploring the origin of this variation, and improving the dependence of the proxy on key geometric factors, extending the proxy to apply to transport channels other than the ITG turbulence it was originally developed for, and are also examining the relative effectiveness of different search algorithms. To help in these efforts, we have adapted STELLOPT to provide a new capability for mapping the topography of the cost function in the search space.

  5. Excitation of Alfvenic instabilities in spherical tokamaks

    International Nuclear Information System (INIS)

    McClements, K.G.; Appel, L.C.; Hole, M.J.; Thyagaraja, A.

    2003-01-01

    Understanding energetic particle confinement in spherical tokamak (STs) is important for optimising the design of ST power plants, and provides a testbed for theoretical modelling under conditions of strong toroidicity and shaping, and high beta. MHD analysis of some recent beam-heated discharges in the MAST ST indicates that high frequency modes observed in these discharges can be identified as toroidal Alfven Eigenmodes (TAEs) and elliptical Alfven Eigenmodes (EAEs). It is possible that such modes could strongly enhance fusion alpha-particle transport in an ST power plant. Computations of TAE growth rates for one particular MAST discharge, made using the HAGIS guiding centre code and benchmarked against analytical estimates, indicate strong drive by sub-Alfvenic neutral beam ions. HAGIS computations using higher mode amplitudes than those observed indicate that whereas co-passing beam ions provide the bulk of he TAE drive, counter-passing ions provide the dominant component of TAE-induced particle losses. Axisymmetric Alfvenic mode activity has been detected during ohmic discharges in MAST. These observations are shown by computational modelling to be consistent with the excitation of global Alfven Eigenmodes (GAEs) with n=0 and low m, driven impulsively by low frequency MHD. (author)

  6. Filterscope diagnostic system on EAST tokamak

    International Nuclear Information System (INIS)

    Xu, Z.; Wu, Z.W.; Gao, W.; Zhang, L.; Huang, J.; Chen, Y.J.; Wu, C.R.; Zhang, P.F.

    2015-01-01

    Filterscope diagnostic system, which is designed for monitoring the line emission in fusion plasma has been widely used on fusion devices such as DIII-D, NSTX, CDX-U, KSTAR etc. On EAST (Experimental Advanced Superconducting Tokamak), a filterscope diagnostic system has been mounted to observe the line emission and visible bremsstrahlung emission in plasma from discharge campaign of 2014. It plays a crucial role in studying Edge Localized Modes (ELM) and H-mode, thanks to its high temporal resolution (0.005ms) and good spatial resolution (∼2cm). Furthermore, multi-channel signals at up to 200kHz sampling rates can be digitized simultaneously. The wavelength covers He II (468.5nm), Li I (670.8nm), Li II (548.3nm), C III (465.0nm), O II (441.5nm), Mo I (386.4nm), W I (400.9nm) and visible bremsstrahlung radiation at 538nm besides Dα (656.1nm) and Dγ (433.9nm) with the corresponding wavelength filters. The new developed filterscope system was operating during the EAST 2014 fall experimental campaign and several types ELMs has been observed. (author)

  7. Spectroscopic study of ohmically heated Tokamak discharges

    International Nuclear Information System (INIS)

    Breton, C.; Michelis, C. de; Mattioli, M.

    1980-07-01

    Tokamak discharges interact strongly with the wall and/or the current aperture limiter producing recycling particles, which penetrate into the discharge and which can be studied spectroscopically. Working gas (hydrogen or deuterium) is usually studied observing visible Balmer lines at several toroidal locations. Absolute measurements allow to obtain both the recycling flux and the global particle confinement time. With sufficiently high resolution the isotopic plasma composition can be obtained. The impurity elements can be divided into desorbed elements (mainly oxygen) and eroded elements (metals from both walls and limiter) according to the plasma-wall interaction processes originating them. Space-and time-resolved emission in the VUV region down to about 20 A will be reviewed for ohmically-heated discharges. The time evolution can be divided into four phases, not always clearly separated in a particular discharge: a) the initial phase, lasting less than 10 ms (the so-called burn-out phase), b) the period of increasing plasma current and electron temperature, lasting typically 10 - 100 ms, c) an eventual steady state (plateau of the plasma current with almost constant density and temperature), d) the increase of the electron density up to or just below the maximum value attainable in a given device. For all these phases the results reported from different devices will be described and compared

  8. Divertor design for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Hill, D.N.; Braams, B.

    1994-05-01

    In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4x L-mode), high beta (β N ≥ 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74 degrees from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m 2 with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities

  9. Metal impurity release in diverted tokamak discharges

    International Nuclear Information System (INIS)

    Staudenmaier, G.; Wampler, W.R.

    1987-01-01

    Plasma-materials interaction at the wall of the main plasma chamber of the divertor tokamak ASDEX was investigated by a combined probe, allowing simultaneous measurements of the erosion rate by neutral particles, and the flux and average energy of neutrals. The erosion was measured by collecting part of the released material which was analysed subsequently by electron induced X-ray analysis. The flux and energy of the impinging neutral particles were measured after each single discharge with an energy resolving carbon resistance probe. The carbon probe is an excellent detector not only for deuterium but also for energetic neutral helium atoms. Wall sputtering by CX neutrals is the dominant iron impurity source in ohmically heated deuterium discharges. However, during discharges in helium not only CX neutral sputtering but also sputtering by helium ions is an important impurity release mechanism. But during auxiliary heating by neutral beams and radio frequency in deuterium, impurity generation by CX neutrals was found to increase with heating power. (orig.)

  10. Neoclassical transport of impurtities in tokamak plasmas

    International Nuclear Information System (INIS)

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

    1981-05-01

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

  11. Electrostatic turbulence in the Tokamak TBR-1

    International Nuclear Information System (INIS)

    Castro, R.M. de.

    1991-01-01

    Characteristics of turbulence at plasma edge of tokamak TBR - 1 are determined from measurements of potentials and density fluctuations, done with a square array of four single Langmuir probes. Two adjacent probes are used to measure the floating potential of the plasma in either poloidal or toroidal directions, the remaining two probes are used to measure saturation current also in poloidal and toroidal directions. Using multiple shot data from the four probe array the radial fluctuation density (n ∼ ) and floating potential (φ ∼ ) profiles are estimated. Analysing the fluctuations spectra the wavenumber-frequency spectrum S(k,ω) from two points measurements is determined. An extension of the cross-correlation concept to a three points correlations leads to the estimation of the fluctuation induced particle flux, from which the particle diffusion coefficient and the convected heat flux can be estimated. All this measurements were performed with and without a resonant magnetic field to verify the eventual influence of this field on the data already mentioned. It was verified that the particle flux is outward and due to electrostatic fluctuations with frequencies lower than 150 khz. (author)

  12. Tokamak L/H mode transition

    Energy Technology Data Exchange (ETDEWEB)

    Tsui, K. H.; Navia, C. E. [Instituto de Fisica-Universidade Federal Fluminense, Campus da Praia Vermelha, Av. General Milton Tavares de Souza s/n, Gragoata, 24.210-346, Niteroi, Rio de Janeiro (Brazil)

    2012-01-15

    Through the non field-aligned rotational tokamak equilibrium of a divergence-free plasma flow with a pair of transformed plasma variables w-vector{sub *}=({mu}{rho}){sup 1/2}{nu}-vector and {mu}p{sub *}=({mu}p+w{sub *}{sup 2}/2)[K. H. Tsui, Phys. Plasmas 18, 072502 (2011)], a preliminary understanding of the L/H equilibrium transition is proposed through a feedback cycle, where the higher plasma flux due to external drives enters the rotational Grad-Shafranov equation through the velocity dependent poloidal plasma {beta} to generate the H equilibrium. This H rotational mode has the characteristics of higher normal electric field and plasma pressure. Coupled to the transport properties of E-vector x B-vector drift transport barrier leading to a higher plasma pressure, this makes the H mode a self-sustained equilibrium. The higher plasma {beta} then feeds back to the equilibrium and completes the feedback loop.

  13. Overview of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Flanagan, C.A.

    1986-01-01

    A national team has developed a baseline concept for a Compact Ignition Tokamak (CIT). The CIT mission is to achieve ignition and provide experimental capability to study the behavior of burning plasma. The design uses large magnetic fields on axis (about 10 T) and large plasma currents (about 9-10 MA). The magnet structure derives high strength from the use of a copper-Inconel composite plate design in the nose of region of the toroidal field (TF) coil and in the ohmic heating solenoid. Inertial cooling is used;liquid nitrogen temperatures are established at the beginning of each pulse. Capability is provided to operate either with a divertor or limiter based plasma. The design is very compact (1.32-m major radius, 0.43-m plasma radius), has 16 TF coils, and has 16 major horizontal access ports, about 30 cm by 80 cm, located between TF coils. The schedule is for a construction project to be authorized for the period FY 1988-93

  14. Maintainability features of the compact ignition tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Bushnell, C.W.

    1986-01-01

    The Compact Ignition Tokamak (CIT) is a deuterium-tritium (D-T) device envisaged to be the next experimental reactor in the US Fusion Program. The reactor will initially operate in a nonactivated hydrogen phase for approximately two years. This will permit verification of the integrity of the total system and allow hands-on repair to equipment which has experienced shakedown and early operation failures. Once D-T operations commence, reactor maintenance will require remote handling techniques. An evaluation has been completed to determine what maintenance operations must be performed on the CIT. A maintenance philosophy has been developed which is based upon the use of manipulator systems and robotics in the test cell. Replacement of life-limited equipment will be accomplished using a modular design approach for components, with simple remotely operable interfaces. Examples of operations to be done remotely include: (1) replacing of rf antennae and Faraday shields, (2) uncoupling diagnostic and fueling penetrations, (3) removing of all port covers, and (4) replacing first wall armor tiles, optical mirrors, and vacuum windows

  15. Current drive in high density tokamak plasma

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  16. Alfven wave experiments on the TORTUS tokamak

    International Nuclear Information System (INIS)

    Ballico, M.J.; Bowden, M.; Brand, G.F.; Brennan, M.H.; Cross, R.C.; Fekete, P.; James, B.W.

    1989-01-01

    Results are presented on the first observations of the Discrete Alfven Wave (DAW) and the first measurements of laser scattering off the kinetic Alfven wave in the TORTUS tokamak. TORTUS is a relatively small device, with major radius R=0.44m, minor radius 0.1m and has previously been operated routinely with B Φ =0.7T, I p =20 kA and n e ∼ 1x10 19 m -3 . Under these conditions, and over a wide frequency range (1-14 MHz), there has been no evidence of the DAW modes observed on TCA. Recently, a minor upgrade of TORTUS has permitted routine operation at B Φ =1.0 T, I p =39 kA, q(a)∼5 and n e ∼1-4 x 10 19 m -3 . At the operating frequency, 3.2 MHz, chosen for this study, DAW modes are observed clearly at both low and high densities. The appearance of DAW modes appears to be due to a steeper current profile at the higher plasma currents now generated in TORTUS. The general behaviour of DAW modes is in fact quite sensitive to the density and current profiles, indicating that DAW modes should provide a useful current profile diagnostic. (author) 6 refs., 2 figs

  17. Multiple time scale methods in tokamak magnetohydrodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Jardin, S.C.

    1984-01-01

    Several methods are discussed for integrating the magnetohydrodynamic (MHD) equations in tokamak systems on other than the fastest time scale. The dynamical grid method for simulating ideal MHD instabilities utilizes a natural nonorthogonal time-dependent coordinate transformation based on the magnetic field lines. The coordinate transformation is chosen to be free of the fast time scale motion itself, and to yield a relatively simple scalar equation for the total pressure, P = p + B/sup 2//2..mu../sub 0/, which can be integrated implicitly to average over the fast time scale oscillations. Two methods are described for the resistive time scale. The zero-mass method uses a reduced set of two-fluid transport equations obtained by expanding in the inverse magnetic Reynolds number, and in the small ratio of perpendicular to parallel mobilities and thermal conductivities. The momentum equation becomes a constraint equation that forces the pressure and magnetic fields and currents to remain in force balance equilibrium as they evolve. The large mass method artificially scales up the ion mass and viscosity, thereby reducing the severe time scale disparity between wavelike and diffusionlike phenomena, but not changing the resistive time scale behavior. Other methods addressing the intermediate time scales are discussed.

  18. What's happening at the edge of tokamaks

    International Nuclear Information System (INIS)

    Crandall, D.H.

    1987-01-01

    Handling the power deposition at the walls of a plasma fusion device and controlling the particle fueling of the plasma originated the interest in the edge of the plasma by magnetic fusion scientists. Recently this interest has intensified because of clear evidence that the quality of the central plasma confinement depends in unexpected ways on details of how the edge plasma is managed. Significant efforts are being pursued to understand and exploit the improved plasma confinement observed in the 'H-mode' obtained with divertors and in the 'super-shots' obtained with low neutral particle flux from the edge of TFTR limiter plasmas. The controls, that determine whether or not these well-confined plasmas are obtained, are applied in the edge plasma where a wealth of atomic and molecular processes occur. A qualitative overview of current research related to plasma edge and desirable features is presented to guide thoughts about atomic processes to be included in modeling and interpreting the plasma edge of tokamaks. (orig.)

  19. The Tokamak Fusion Core Experiment studies

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  20. Current drive in spherical tokamak plasmas

    International Nuclear Information System (INIS)

    Storer, R.

    1999-01-01

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

  1. How much does a tokamak reactor cost?

    Science.gov (United States)

    Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

    2017-10-01

    The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

  2. Tokamak-FED plasma-engineering assessments

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  3. Tokamak power system studies at ANL

    International Nuclear Information System (INIS)

    Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

    1986-06-01

    The following features, in particular, have been examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 6 to 7 T; (c) low toroidal current (I ≅ 4MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields they produce in the plasma. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors

  4. Compression experiments on the TOSKA tokamak

    International Nuclear Information System (INIS)

    Cima, G.; McGuire, K.M.; Robinson, D.C.; Wootton, A.J.

    1980-10-01

    Results from minor radius compression experiments on a tokamak plasma in TOSCA are reported. The compression is achieved by increasing the toroidal field up to twice its initial value in 200μs. Measurements show that particles and magnetic flux are conserved. When the initial energy confinement time is comparable with the compression time, energy gains are greater than for an adiabatic change of state. The total beta value increases. Central beta values approximately 3% are measured when a small major radius compression is superimposed on a minor radius compression. Magnetic field fluctuations are affected: both the amplitude and period decrease. Starting from low energy confinement times, approximately 200μs, increases in confinement times up to approximately 1 ms are measured. The increase in plasma energy results from a large reduction in the power losses during the compression. When the initial energy confinement time is much longer than the compression time, the parameter changes are those expected for an adiabatic change of state. (author)

  5. A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

    Science.gov (United States)

    Windsor, C. G.; Pautasso, G.; Tichmann, C.; Buttery, R. J.; Hender, T. C.; EFDA Contributors, JET; ASDEX Upgrade Team

    2005-05-01

    First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems.

  6. A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

    International Nuclear Information System (INIS)

    Windsor, C.G.; Buttery, R.J.; Hender, T.C.; Pautasso, G.; Tichmann, C.

    2005-01-01

    First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems

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

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S.

    1995-10-18

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

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

    International Nuclear Information System (INIS)

    Gillen, T.J.

    1980-11-01

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

  9. Deposit of thin films for Tokamaks conditioning; Deposito de peliculas delgadas para acondicionar Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Valencia A, R

    2006-07-01

    As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (<10 eV), both in noble and reactive gases, as well as the conditioning by thin film deposits of hydrogen rich amorphous carbon (carbonization) leading to a reduction in the plasma resistivity from 8.99 x 10{sup -6} to 4.5 x 10{sup -6} {omega}-m, thus taking the Z{sub ef} value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission

  10. Key features and progress of the KSTAR tokamak engineering

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  11. Time evolution of mass flows in a collisional tokamak

    International Nuclear Information System (INIS)

    Hassam, A.B.; Kulsrud, R.M.

    1978-04-01

    The time evolution, due to dissipative processes, of an initial pattern of poloidal and toroidal mass flows in a tokamak is considered. The calculation is applicable to a collisional, low β, axisymmetric tokamak of arbitrary minor cross section. Time rates of change of poloidal flows which are subsonic but larger than the diamagnetic speed are given according to the magnitude of the flow and the collisionality of the plasma. Over most of parameter space for typical tokamaks, the poloidal rotation is strongly damped by magnetic pumping at the rate (l/qR) 2 ν/sub ii/, where l is the mean free path, qR the ''connection length,'' and ν/sub ii/ the ion-ion collision frequency. At higher speeds, even stronger damping is effected by electron thermal conduction. The toroidal rotation is determined largely by the conservation of toroidal angular momentum. A heuristic explanation of the damping due to magnetic pumping is given

  12. Dynamic analysis of the tokamak support system in ITER

    International Nuclear Information System (INIS)

    Tado, S.; Kitamura, K.; Itou, Y.; Koizumi, K.; Tada, E.; Tsunematsu, T.

    1998-01-01

    The ITER tokamak machine has a relatively flexible support structure against horizontal loads as from an earthquake. Modal analyses and modal transient analyses have been carried out by using the finite element analysis code NASTRAN to assess the response of the tokamak driven by a horizontal seismic load with the ITER design response spectra. Two types of three-dimensional models and a simplified mass/spring model have been used as structural models. The three-dimensional models have been used for modal analyses and modal transient analyses. The simplified mass/spring model has been used for parameter surveying of the transient response. It has been shown in the analyses that the lowest natural frequency of the tokamak is about 2 cycles s -1 and that the maximum relative response displacement between the magnet and the vacuum vessel is about 14 mm. (orig.)

  13. A control approach for plasma density in tokamak machines

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Scharer, J.E.

    1992-01-01

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

  15. SUPERCOIL: A model for the computational design of tokamaks

    International Nuclear Information System (INIS)

    Borrass, K.

    1986-01-01

    SUPERCOIL is a code for the computational design of tokamaks, including in particular ignited next-generation devices. It takes into account all physical, technical and geometrical constraints relevant to the basic design of a tokamak. Among the solutions of the model equations that meet all constraints the one optimized with respect to a prescribed figure of merit (typically capital cost) is determined. The main components modelled are the plasma, blanket and shield, the TF system and the PF system. The main characteristics of the component models are summarized and a detailed description of the solution method is given. A typical NET study point is given as an illustrative example. The validity of the code is assessed by applying it to the ASDEX Upgrade tokamak. (orig.)

  16. Nonlinear MHD simulations of spherical tokamak and helical plasmas

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-11-01

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

  17. Profile-turbulence interactions, magnetohydrodynamic relaxations, and transport in tokamaks

    Science.gov (United States)

    Thyagaraja, A.; Knight, P. J.; de Baar, M. R.; Hogeweij, G. M. D.; Min, E.

    2005-09-01

    The dynamical behavior of the global, two-fluid, electromagnetic model of a tokamak plasma is explored under conditions corresponding to the Rijnhuizen tokamak project [A. J. H. Donné, Plasma Phys. Rep. 20, 192 (1994)] using the CUTIE code [A. Thyagaraja, Plasma Phys. Controlled Fusion 42, B255 (2000)]. Simulations of an off-axis electron-cyclotron-heated (350kW) hydrogen discharge and a purely Ohmic one over several resistive evolution times (τres≃15-20ms) are described. The results illustrate profile-turbulence interactions and the spectral transfer processes implicated in the spontaneous generation and maintenance of mesoscale zonal flows and dynamo currents. Relaxation phenomena, including off- and on-axis sawteeth and periodically repeating edge ballooning instabilities mediated by these mechanisms, are presented. The CUTIE model reproduces many observed features of the experiment qualitatively and suggests that global electromagnetic simulations may play an essential role in understanding tokamak turbulence and transport.

  18. Axisymmetric stability of vertically asymmetric tokamaks at large beta poloidal

    International Nuclear Information System (INIS)

    Yamazaki, K.; Fishman, H.; Okabayashi, M.; Todd, A.M.M.

    1981-09-01

    The stability of high-β vertically asymmetric tokamak equilibria to rigid displacements is investigated analytically. It is found that vertical stability at large beta poloidal is mainly determined by a coupling between the shape of the plasma surface and the Shafranov shift of the magnetic axis. To the lowest order, symmetric components of the plasma surface shape are found to be the critical destabilizing elements. Asymmetric components have little effect. The inclusion of higher order terms in the high β tokamak expansion leads to further destabilization. Qualitative agreement between these analytic results and numerical stability calculations using the PEST code is demonstrated

  19. Impurity control in near-term tokamak reactors

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Smith, D.L.; Brooks, J.N.

    1976-10-01

    Several methods for reducing impurity contamination in near-term tokamak reactors by modifying the first-wall surface with a low-Z or low-sputter material are examined. A review of the sputtering data and an assessment of the technological feasibility of various wall modification schemes are presented. The power performance of a near-term tokamak reactor is simulated for various first-wall surface materials, with and without a divertor, in order to evaluate the likely effect of plasma contamination associated with these surface materials

  20. Technology issues for decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  2. Conceptual analysis of a tokamak reactor with lithium dust jet

    International Nuclear Information System (INIS)

    Kuteev, B.V.; Krylov, S.V.; Sergeev, V.Yu.; Skokov, V.G.; Timokhin, V.M.

    2010-01-01

    The steady-state operation of tokamak reactors requires radiating a substantial part of the fusion energy dissipated in plasma to make more uniform the heat loads onto the first wall and to reduce the erosion of the divertor plates. One of the approaches to realize this goal uses injection of lithium dust jet into the scrape-off layer (SOL). A quantitative conceptual analysis of the reactor parameters with lithium dust jet injection is presented here. The effects of the lithium on the core and SOL plasma are considered. The first results of developing the lithium jet injection technology and its application to the T-10 tokamak are also presented.

  3. Geodesic acoustic modes in noncircular cross section tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sorokina, E. A., E-mail: sorokina.ekaterina@gmail.com; Lakhin, V. P. [National Research Center “Kurchatov Institute,” (Russian Federation); Konovaltseva, L. V. [People’s Friendship University of Russia (Russian Federation); Ilgisonis, V. I. [National Research Center “Kurchatov Institute,” (Russian Federation)

    2017-03-15

    The influence of the shape of the plasma cross section on the continuous spectrum of geodesic acoustic modes (GAMs) in a tokamak is analyzed in the framework of the MHD model. An expression for the frequency of a local GAM for a model noncircular cross section plasma equilibrium is derived. Amendments to the oscillation frequency due to the plasma elongation and triangularity and finite tokamak aspect ratio are calculated. It is shown that the main factor affecting the GAM spectrum is the plasma elongation, resulting in a significant decrease in the mode frequency.

  4. Rippling and drift instabilities in the straight cylinder tokamak

    International Nuclear Information System (INIS)

    Rogister, A.

    1984-01-01

    It is shown that the electron and ion diamagnetic drifts stabilize the rippling mode in the straigth cylindrical tokamak model. Parallel electron heat conduction is further stabilizing if the parameter etasub(e) = dlnTsub(e)/dlnN is positive. This has a consequence that the mode does not survive at temperatures exceeding, typically, 50 eV for standard values of magnetic field and density. The collisional drift wave is found to be always stable even when the effect of the tokamak current is included in the calculation. (orig.)

  5. 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

  6. Design and Analysis of the Thermal Shield of EAST Tokamak

    International Nuclear Information System (INIS)

    Xie Han; Liao Ziying

    2008-01-01

    EAST (Experimental Advanced Superconducting Tokamak) is a tokamak with superconducting toroidal and poloidal magnets operated at 4.5 K. In order to reduce the thermal load applied on the surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system at a minimum, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This study employed NASTRAN and ANSYS finite element codes to analyze the stress under a spectrum of loading conditions and combinations, providing a theoretical basis for an optimization design of the structure.

  7. Design and analysis of the thermal shield of EAST tokamak

    International Nuclear Information System (INIS)

    Xie Han; Liao Ziying

    2007-01-01

    The EAST (Experimental Advanced Superconducting Tokamak) is a Tokamak with superconducting toroidal and poloidal magnets operating at 4.5 K. In order to reduce the thermal load applied to surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system in minimal, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This paper employed NASTRAN and ANSYS finite element software to analyze the stress under a spectrum of loading conditions and combinations, providing a basis in theory for optimization design of the structure. (authors)

  8. Linear and nonlinear kinetic-stability studies in tokamaks

    International Nuclear Information System (INIS)

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

    1982-09-01

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

  9. Theoretical investigations of lower hybrid resonance heating in a tokamak

    International Nuclear Information System (INIS)

    Momota, Hiromu; Azumi, Masafumi; Okamoto, Masao; Takizuka, Tomonori; Fukuyama, Atsushi.

    1977-03-01

    Theoretical investigations of the lower hybrid wave heating in a tokamak have been performed systematically. The requisite power for the additional heating of a large scale tokamak, JT-60, is estimated to be about 10 MW. Linear behaviour of an incident wave is extensively examined and its physical picture is clarified. The effect of toroidicity on propagation and dispersion of a lower hybrid wave is also studied. Parametric phenomena are treated without dipole approximation and the threshold in JT-60 is also examined. The formation of the high energy ion tail attributed to a high power wave is shown by theory and a good agreement with numerical calculation is obtained. (auth.)

  10. The 'FINTOR 1' design. A minimum size Tokamak Experimental Reactor

    International Nuclear Information System (INIS)

    Bertolini, E.

    1978-01-01

    The work carried on by the FINTOR (Frascati, Ispra, Napoli, Tokamak Reactor) group has now reached a stage where the effects of the main physics and engineering constraints, for a minimum size Tokamak Experimental Reactor have been clearly identified. This phase, now completed, has allowed to produce a self-consistent design of each basic component of the reactor, FINTOR 1, and to identify the more relevant interface problems toward a further optimisation of the reactor dimensions and characteristics to be performed in the future (FINTOR 2)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hively, L.M.

    1980-01-01

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

  12. Scaling for scrape-off layer plasma in tokamak

    International Nuclear Information System (INIS)

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

    1977-12-01

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

  13. Internal Magnetic Configuration Measured by ECE Imaging on EAST Tokamak

    International Nuclear Information System (INIS)

    Xu Ming; Wen Yizhi; Xie Jinlin; Yu Changxuan; Gao Bingxi; Xu Xiaoyuan; Liu Wandong; Hu Liqun; Sun Youwen; Qian Jinping; Wan Baonian

    2013-01-01

    ECE imaging (electron cyclotron emission imaging) is an important diagnostic which can give 2D imaging of temperature fluctuation in the core of tokamak. A method based on ECE imaging is introduced which can give the information of the position of magnetic axis and the structure of internal magnetic surface for EAST tokamak. The EFIT equilibrium reconstruction is not reliable due to the absence of important core diagnostic at the initial phase for EAST, so the information given by ECE imaging could help to improve the accuracy of EFIT equilibrium reconstruction. (magnetically confined plasma)

  14. Tokamak start-up with electron-cyclotron heating

    International Nuclear Information System (INIS)

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

    1981-01-01

    Experiments are described in which the start-up voltage in a tokamak is reduced by about a factor of two by the use of a modest amount of electron cyclotron resonance heating power for pre-ionization. The solution of the zero-dimensional start-up equations indicates that the effect is due to the high initial density which increases the rate at which the conductivity increases in the neutral-dominated initial plasma. The effect extrapolates favourably to larger tokamaks. A 50% reduction in the start-up volt-second requirement and impurity reflux is also observed. (author)

  15. Control of plasma position in the CASTOR tokamak

    International Nuclear Information System (INIS)

    Valovic, M.

    1988-11-01

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

  16. 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

  17. Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Abiteboul, J.

    2012-10-01

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

  18. High and low frequency Alfven modes in tokamaks

    International Nuclear Information System (INIS)

    Briguglio, S.; Fogaccia, G.; Vlad, G.; Zonca, F.; Chen, L.; Dong, J.Q.; Santoro, R.A.

    2001-01-01

    We present an analysis of the typical features of shear Alfven waves in tokamak plasmas in a frequency domain ranging from the ''high'' frequencies (ω ≅ ν A /2qR 0 ; ν A being the Alfven speed and qR 0 the tokamak connection length) of the toroidal gap to the ''low'' frequencies, comparable with the thermal ion diamagnetic frequency, ω *pi and/or the thermal ion transit frequency ω ti = ν ti /qR 0 (ν ti being the ion thermal speed). (author)

  19. ELECTRON CYCLOTRON CURRENT DRIVE EFFICIENCY IN GENERAL TOKAMAK GEOMETRY

    International Nuclear Information System (INIS)

    LIN-LUI, Y.R; CHAN, V.S; PRATER, R.

    2003-01-01

    Green's-function techniques are used to calculate electron cyclotron current drive (ECCD) efficiency in general tokamak geometry in the low-collisionality regime. Fully relativistic electron dynamics is employed in the theoretical formulation. The high-velocity collision model is used to model Coulomb collisions and a simplified quasi-linear rf diffusion operator describes wave-particle interactions. The approximate analytic solutions which are benchmarked with a widely used ECCD model, facilitate time-dependent simulations of tokamak operational scenarios using the non-inductive current drive of electron cyclotron waves

  20. Data processing system for spectroscopy at Novillo Tokamak

    International Nuclear Information System (INIS)

    Ortega C, G.; Gaytan G, E.

    1998-01-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

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

    International Nuclear Information System (INIS)

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

    1983-04-01

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

  2. Integrated modeling of tokamak experiments with OMFIT

    International Nuclear Information System (INIS)

    Meneghini, Orso; Lao, Lang

    2013-01-01

    One Modeling Framework for Integrated Tasks (OMFIT) is a framework that allows data to be easily exchanged among different codes by providing a unifying data structure. The main idea at the base of OMFIT is to treat files, data and scripts as a uniform collection of objects organized into a tree structure, which provides a consistent way to access and manipulate such collection of heterogeneous objects, independent of their origin. Within the OMFIT tree, data can be copied/referred from one node to another and tasks can call each other allowing for complex compound task to be built. A top-level Graphical User Interface (GUI) allowing users to manage tree objects, carry out simulations and analyze the data either interactively or in batch. OMFIT supports many scientific data formats and when a file is loaded into the framework, its data populates the tree structure, automatically endowing it with many potential uses. Furthermore, seamless integration with experimental management systems allows direct manipulation of their data. In OMFIT modeling tasks are organized into modules, which can be easily combined to create arbitrarily-large multi-physics simulations. Modules inter-dependencies are seamlessly defined by variables referencing tree locations among them. Creation of new modules and customization of existing ones is encouraged by graphical tools for their management and an online repository. High level Application Programmer Interfaces (APIs) enable users to execute their codes on remote servers and creation application-specific GUIs. Finally, within OMFIT it is possible to visualize experimental and modeling data for both quick analysis and publication purposes. Examples of application to the DIII-D tokamak are presented. (author)

  3. Tokamak nonmaxwellian plasma dynamics in thermonuclear regime

    International Nuclear Information System (INIS)

    Cotsaftis, M.

    1987-01-01

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

  4. Fusion neutron diagnostics on ITER tokamak

    Science.gov (United States)

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

    2012-04-01

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

  5. Progress in diagnostics of the COMPASS tokamak

    Science.gov (United States)

    Weinzettl, V.; Adamek, J.; Berta, M.; Bilkova, P.; Bogar, O.; Bohm, P.; Cavalier, J.; Dejarnac, R.; Dimitrova, M.; Ficker, O.; Fridrich, D.; Grover, O.; Hacek, P.; Havlicek, J.; Havranek, A.; Horacek, J.; Hron, M.; Imrisek, M.; Komm, M.; Kovarik, K.; Krbec, J.; Markovic, T.; Matveeva, E.; Mitosinkova, K.; Mlynar, J.; Naydenkova, D.; Panek, R.; Paprok, R.; Peterka, M.; Podolnik, A.; Seidl, J.; Sos, M.; Stockel, J.; Tomes, M.; Varavin, M.; Varju, J.; Vlainic, M.; Vondracek, P.; Zajac, J.; Zacek, F.; Stano, M.; Anda, G.; Dunai, D.; Krizsanoczi, T.; Refy, D.; Zoletnik, S.; Silva, A.; Gomes, R.; Pereira, T.; Popov, Tsv.; Sarychev, D.; Ermak, G. P.; Zebrowski, J.; Jakubowski, M.; Rabinski, M.; Malinowski, K.; Nanobashvili, S.; Spolaore, M.; Vianello, N.; Gauthier, E.; Gunn, J. P.; Devitre, A.

    2017-12-01

    The COMPASS tokamak at IPP Prague is a small-size device with an ITER-relevant plasma geometry and operating in both the Ohmic as well as neutral beam assisted H-modes since 2012. A basic set of diagnostics installed at the beginning of the COMPASS operation has been gradually broadened in type of diagnostics, extended in number of detectors and collected channels and improved by an increased data acquisition speed. In recent years, a significant progress in diagnostic development has been motivated by the improved COMPASS plasma performance and broadening of its scientific programme (L-H transition and pedestal scaling studies, magnetic perturbations, runaway electron control and mitigation, plasma-surface interaction and corresponding heat fluxes, Alfvenic and edge localized mode observations, disruptions, etc.). In this contribution, we describe major upgrades of a broad spectrum of the COMPASS diagnostics and discuss their potential for physical studies. In particular, scrape-off layer plasma diagnostics will be represented by a new concept for microsecond electron temperature and heat flux measurements - we introduce a new set of divertor Langmuir and ball-pen probe arrays, newly constructed probe heads for reciprocating manipulators as well as several types of standalone probes. Among optical tools, an upgraded high-resolution edge Thomson scattering diagnostic for pedestal studies and a set of new visible light and infrared (plasma-surface interaction investigations) cameras will be described. Particle and beam diagnostics will be covered by a neutral particle analyzer, diagnostics on a lithium beam, Cherenkov detectors (for a direct detection of runaway electrons) and neutron detectors. We also present new modifications of the microwave reflectometer for fast edge density profile measurements.

  6. First results from the T-10 tokamak

    International Nuclear Information System (INIS)

    Berlizov, A.B.; Bobrovskij, G.A.; Bagdasarov, A.A.

    1977-01-01

    Measurements of plasma parameters are made on the T-10 tokamak with a toroidal magnetic field of 3.5 T and a plasma current of 0.4 MA. In macroscopically stable discharges of 1 s duration, the central electron temperature is Tsub(e)(0)=1.2 keV, the mean electron density reaches a value of nsub(e)=5X10 13 cm -3 , and the central ion temperature is measured to be 0.6-0.8 keV. The thermonuclear neutron yield reaches a value of 4X10 9 neutrons per shot, in agreement with the Tsub(i)(0) value measured by charge exchange. Sawtooth X-ray oscillations are observed. The effective ionic charge is found to be less than 2 for the inner region of the plasma column. The energy confinement time tausub(E) is calculated from the experimental profiles of the plasma parameters. The value of tausub(E) is 40 ms and increases up to 60 ms, while nsub(e) is increased up to 8.5X10 13 cm -3 as a result of cold-gas injection by a pulse valve. Violent disruption is observed in several regimes. Hard-X-ray and neutron radiation bursts take place during the disruption, both in hydrogen and in deuterium. More intensive and prolonged radiation fluxes of hard X-rays and non-thermonuclear neutrons are observed in some discharges where very intensive beams of relativistic runaway electrons seem to exist. (author)

  7. Parasitic momentum flux in the tokamak core

    Science.gov (United States)

    Stoltzfus-Dueck, T.

    2017-10-01

    Tokamak plasmas rotate spontaneously without applied torque. This intrinsic rotation is important for future low-torque devices such as ITER, since rotation stabilizes certain instabilities. In the mid-radius `gradient region,' which reaches from the sawtooth inversion radius out to the pedestal top, intrinsic rotation profiles may be either flat or hollow, and can transition suddenly between these two states, an unexplained phenomenon referred to as rotation reversal. Theoretical efforts to explain the mid-radius rotation shear have largely focused on quasilinear models, in which the phase relationships of some selected instability result in a nondiffusive momentum flux (``residual stress''). In contrast, the present work demonstrates the existence of a robust, fully nonlinear symmetry-breaking momentum flux that follows from the free-energy flow in phase space and does not depend on any assumed linear eigenmode structure. The physical origin is an often-neglected portion of the radial ExB drift, which is shown to drive a symmetry-breaking outward flux of co-current momentum whenever free energy is transferred from the electrostatic potential to ion parallel flows. The fully nonlinear derivation relies only on conservation properties and symmetry, thus retaining the important contribution of damped modes. The resulting rotation peaking is counter-current and scales as temperature over plasma current. As first demonstrated by Landau, this free-energy transfer (thus also the corresponding residual stress) becomes inactive when frequencies are much higher than the ion transit frequency, which allows sudden transitions between hollow and flat profiles. Simple estimates suggest that this mechanism may be consistent with experimental observations. This work was funded in part by the Max-Planck/Princeton Center for Plasma Physics and in part by the U.S. Dept. of Energy, Office of Science, Contract No. DE-AC02-09CH11466.

  8. Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

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

    Indian Academy of Sciences (India)

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

  10. Revised design for the Tokamak experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-03-01

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

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

    Indian Academy of Sciences (India)

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

  12. 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

  13. Conceptual Design of Alborz Tokamak Poloidal Coils System

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.

    2013-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.

  14. Investigations of low qa discharges in the SINP tokamak

    Indian Academy of Sciences (India)

    Low edge safety factor discharges including very low (1 < < 2) and ultra low (0 < < 1) have been obtained in the SINP tokamak. It has been observed that accessibility of these discharges depends crucially on the fast rate of plasma current rise. Several interesting results in terms of different time scales like ...

  15. Measurements of plasma position in TJ-I Tokamak

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Quasilinear kinetic modeling of RMP penetration into a tokamak plasma

    International Nuclear Information System (INIS)

    Heyn, M.F.; Kernbichler, W.; Leitner, P.; Ivanov, I.B.; Kasilov, S.V.

    2013-01-01

    The linear as well as the quasilinear problem of RMP penetration in tokamaks is solved consistently with a particle and energy conserving collision operator. The new collision operator ensures the Onsager symmetry of the quasilinear transport coefficient matrix and avoids artifacts such as fake heat convection connected with simplified collision models.

  17. 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

  18. A flexible software architecture for tokamak discharge control systems

    International Nuclear Information System (INIS)

    Ferron, J.R.; Penaflor, B.; Walker, M.L.; Moller, J.; Butner, D.

    1995-01-01

    The software structure of the plasma control system in use on the DIII-D tokamak experiment is described. This system implements control functions through software executing in real time on one or more digital computers. The software is organized into a hierarchy that allows new control functions needed to support the DIII-D experimental program to be added easily without affecting previously implemented functions. This also allows the software to be portable in order to create control systems for other applications. The tokamak operator uses an X-windows based interface to specify the time evolution of a tokamak discharge. The interface provides a high level view for the operator that reduces the need for detailed knowledge of the control system operation. There is provision for an asynchronous change to an alternate discharge time evolution in response to an event that is detected in real time. Quality control is enhanced through off-line testing that can make use of software-based tokamak simulators

  19. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    Amiel, Stephane

    2014-01-01

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

  20. Methods for the design and optimization of shaped tokamaks

    International Nuclear Information System (INIS)

    Haney, S.W.

    1988-05-01

    Two major questions associated with the design and optimization of shaped tokamaks are considered. How do physics and engineering constraints affect the design of shaped tokamaks? How can the process of designing shaped tokamaks be improved? The first question is addressed with the aid of a completely analytical procedure for optimizing the design of a resistive-magnet tokamak reactor. It is shown that physics constraints---particularly the MHD beta limits and the Murakami density limit---have an enormous, and sometimes, unexpected effect on the final design. The second question is addressed through the development of a series of computer models for calculating plasma equilibria, estimating poloidal field coil currents, and analyzing axisymmetric MHD stability in the presence of resistive conductors and feedback. The models offer potential advantages over conventional methods since they are characterized by extremely fast computer execution times, simplicity, and robustness. Furthermore, evidence is presented that suggests that very little loss of accuracy is required to achieve these desirable features. 94 refs., 66 figs., 14 tabs

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

    International Nuclear Information System (INIS)

    Ehst, D.A.

    1995-08-01

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

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

    International Nuclear Information System (INIS)

    Pedrotti, L.R.

    1977-01-01

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

  3. Evidence of Inward Toroidal Momentum Convection in the JET Tokamak

    DEFF Research Database (Denmark)

    Tala, T.; Zastrow, K.-D.; Ferreira, J.

    2009-01-01

    Experiments have been carried out on the Joint European Torus tokamak to determine the diffusive and convective momentum transport. Torque, injected by neutral beams, was modulated to create a periodic perturbation in the toroidal rotation velocity. Novel transport analysis shows the magnitude an...

  4. 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

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

    Czech Academy of Sciences Publication Activity Database

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

    2007-01-01

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

  6. Conceptual design of a radio-frequency driven compact tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Ludwig, G.O.; Montes, A.; Ueda, M.; Goes, L.C.S.

    1987-09-01

    Preliminary results of the design of a small compact tokamak are presented. The design incorporates advanced concepts as start-up and current drive by electron-cyclotron and lower-hybrid waves; plasma heating by intense ion beams; and achievement of high-β by decreasing the aspect ratio. (author) [pt

  7. On the choice of toroidal magnetic field for thermonuclear tokamaks

    International Nuclear Information System (INIS)

    Segre, S.E.

    1981-01-01

    The value of the magnetic field chosen for tokamak experiments is the result of a compromise between physics requirements, technological limits and financial constraints. The consequences of some physics requirements and limitations, in the light of recent results on the scaling of energy confinement and on limits of density are examined. (author)

  8. High beta tokamaks. [MHD equilibrium, stability, and transport calculations

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-01-01

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

  9. Recording non-local temperature rise in the RTP tokamak

    NARCIS (Netherlands)

    Hogeweij, G. M. D.; Mantica, P.; Gorini, G.; de Kloe, J.; Cardozo, N. J. L.; R. T. P. Team,

    2000-01-01

    In the Rijnhuizen Tokamak Project (RTP) plasmas with electron cyclotron heating (ECH), a transient rise of the core T-e is observed when hydrogen pellets are injected tangentially to induce fast cooling of the peripheral region. The core T-e rise is a sharp function of the normalized power

  10. Experiment of laser thomson scattering at HL-1 tokamak device

    International Nuclear Information System (INIS)

    Zuo Henian; Chen Jiafu; Yan Derong; Liu Aiping; Shi Peilan; Wang Wei; Liu Xiaomei

    1989-05-01

    The structure and performance of the Ruby Laser Thomson Scattering apparatus for HL-1 tokamak device is described. The method of acquisition and calibration of multichannel scattered signals are presented. Examples of measured electron temperature T. with experimental error are given

  11. Stochasticity and the m = 1 mode in tokamaks

    International Nuclear Information System (INIS)

    Izzo, R.; Monticello, D.A.; Stodiek, W.; Park, W.

    1986-05-01

    It has recently been proposed that stochasticity resulting from toroidal coupling could lead to a saturation of the m = 1 internal mode in tokamaks. We present results from the nonlinear evolution of the m = 1 mode with toroidal coupling that show that stochasticity is not enough to cause saturation of the m = 1 mode

  12. High-pressure, flux-conserving tokamak equilibria

    Energy Technology Data Exchange (ETDEWEB)

    Dory, R.A.; Peng, Y.K.M.

    1976-08-01

    Magnetohydrodynamic (MHD) tokamak equilibria are found with values of ..beta.. up to 20 percent and prescribed MHD safety factor values (e.g., q(axis) = 1 and q(edge) = 4.8) for tokamaks with aspect ratio A = 4 and D-shaped cross section. If such equilibria could be attained experimentally, they would be very attractive for decreasing the projected costs of tokamak power reactors substantially. In the flux-conserving tokamak (FCT) model, where rapid heating is applied to an already relatively hot plasma, these high ..beta.. equilibria are achievable. We study the quasi-static evolution of FCT equilibria as ..beta.. increases. An operating window is found in the pressure profile width w/sub p/: for high ..beta.. the values of w/sub p/ must lie between 0.40 and 0.55 of the plasma minor width. Within this window, plasma current and poloidal ..beta.. increase monotonically with ..beta... For fixed plasma boundary, significant poloidal surface currents are induced, but these can be eliminated by small increases in the plasma minor radius, the pressure profile width, and the vacuum toroidal field.

  13. 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)

  14. Plasma recycling and marfe phenomenon in a tokamak

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Hu Jiansheng; Li Chengfu; He Yexi

    1998-10-01

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

  16. Post-disruptive runaway electron beams in the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Vlainić, M.; Mlynář, Jan; Cavalier, Jordan; Weinzettl, Vladimír; Papřok, Richard; Imríšek, Martin; Ficker, Ondřej; Varavin, Mykyta; Noterdaeme, J.M.

    2015-01-01

    Roč. 81, č. 5 (2015), s. 475810506-475810506 ISSN 0022-3778 R&D Projects: GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * runaway electron beam * COMPASS Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.981, year: 2015

  17. 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

  18. Investigations of low qa discharges in the SINP tokamak

    Indian Academy of Sciences (India)

    Abstract. Low edge safety factor discharges including very low qa (1 qa. 2) and ultra low qa. (0 qa. 1) have been obtained in the SINP tokamak. It has been observed that accessibility of these discharges depends crucially on the fast rate of plasma current rise. Several interesting results in terms of different time scales like ...

  19. 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

  20. Current drive by Alfven waves in elongated cross section tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Inst. de Fisica; Assis, A.S. de [Universidade Federal Fluminense, Niteroi, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves (GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory