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

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

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

A computer code for Tokamak reactor concepts evaluation

International Nuclear Information System (INIS)

Rosatelli, F.; Raia, G.

1985-01-01

A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

Design study of toroidal magnets for tokamak experimental power reactors

International Nuclear Information System (INIS)

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

1976-12-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Oak Ridge Tokamak experimental power reactor study scoping report

International Nuclear Information System (INIS)

Roberts, M.

1977-03-01

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

Tokamak experimental power reactor conceptual design. Volume I

International Nuclear Information System (INIS)

1976-08-01

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

Neutronics calculations for the Oak Ridge National Laboratory Tokamak Reactor Studies

International Nuclear Information System (INIS)

Santoro, R.T.; Baker, V.C.; Barnes, J.M.

1976-01-01

Neutronics calculations have been carried out to analyze the nuclear performance of conceptual blanket and shield designs for the Tokamak Experimental Power Reactor (EPR) and the Tokamak Demonstration Reactor Plant (DRP) being considered at the Oak Ridge National Laboratory. These reactor designs represent a sequence in the commercialization of fusion-generated electrical power. All of the calculations were carried out using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV coupled neutron-gamma-ray transport cross-section data, fluence-to-kerma conversion factors, and radiation damage cross-section data. The calculations include spatial and integral heating-rate estimates in the reactor with emphasis on the recovery of fusion neutron energy in the blanket and limiting the heat-deposition rate in the superconducting toroidal field coils. Radiation damage due to atomic displacements and gas production produced in the reactor structural material and in the toroidal field coil windings were also estimated. The tritium-breeding ratio when natural lithium is used as the fertile material in the DRP blanket and in the experimental breeding modules in the EPR is also given

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

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

1988-01-01

The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

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

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

Design studies of Tokamak power reactor in JAERI

International Nuclear Information System (INIS)

Tone, T.; Nishikawa, M.; Tanaka, Y.

1985-01-01

Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

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

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

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

Decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

The Tokamak Fusion Test Reactor D-T modifications and operations

International Nuclear Information System (INIS)

1992-01-01

This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

Magnet design considerations for Tokamak fusion reactors

International Nuclear Information System (INIS)

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

1976-01-01

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

Energy storage for tokamak reactor cycles

International Nuclear Information System (INIS)

Buchanan, C.H.

1979-01-01

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

Cross-section sensitivity analyses for a Tokamak Experimental Power Reactor

International Nuclear Information System (INIS)

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

1977-09-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Physics design of an ultra-long pulsed tokamak reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Conceptual design of the steady state tokamak reactor (SSTR)

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Review of tokamak power reactor and blanket designs in the United States

International Nuclear Information System (INIS)

Baker, C.; Brooks, J.; Ehst, D.; Gohar, Y.; Smith, D.; Sze, D.

1986-01-01

The last major conceptual design study of a tokamak power reactor in the United States was STARFIRE which was carried out in 1979-1980. Since that time US studies have concentrated on engineering test reactors, demonstration reactors, parametric systems studies, scoping studies, and studies of selected critical issues such as pulsed vs. steady-state operation and blanket requirements. During this period, there have been many advancements in tokamak physics and reactor technology, and there has also been a recognition that it is desirable to improve the tokamak concept as a commercial power reactor candidate. During 1984-1985 several organizations participated in the Tokamak Power Systems Study (TPSS) with the objective of developing ideas for improving the tokamak as a power reactor. Also, the US completed a comprehensive Blanket Comparison and Selection Study which formed the basis for further studies on improved blankets for fusion reactors

New directions in tokamak reactors

International Nuclear Information System (INIS)

Baker, C.C.

1985-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Important aspects of radiation shielding for fusion reactor tokamaks

International Nuclear Information System (INIS)

Abdou, M.A.

1977-01-01

Radiation shielding is a key subsystem in tokamak reactors. Design of this shield must evolve from economic and technological trade-off studies that account for the strong interrelations among the various components of the reactor system. These trade-offs are examined for the bulk shield on the inner side of the torus and for the special shields of major penetrations. Results derived are applicable for a large class of tokamak-type reactors

Contour analysis of steady state tokamak reactor performance

International Nuclear Information System (INIS)

Devoto, R.S.; Fenstermacher, M.E.

1990-01-01

A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

Compact tokamak reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

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

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

International Nuclear Information System (INIS)

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

2016-01-01

Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Tokamak control simulator

International Nuclear Information System (INIS)

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

1976-01-01

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

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

DEFF Research Database (Denmark)

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

2012-01-01

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

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

International Nuclear Information System (INIS)

Xu Qiang; Wang Shaojie

1997-06-01

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

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-01-10

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

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

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

A conceptual design of a negative-ion-grounded advanced tokamak reactor

International Nuclear Information System (INIS)

Yamamoto, Shin; Ohara, Yoshihiro; Tani, Keiji

1988-05-01

The NAVIGATOR concept is based on the negative-ion-grounded 500 keV 20 MW neutral beam injection system (NBI system), which has been proposed and studied at JAERI. The NAVIGATOR concept contains two categories; one is the NAVIGATOR machine as a tokamak reactor, and the other is the NAVIGATOR philosophy as a guiding principle in fusion research. The NAVIGATOR machine implies an NBI heated and full inductive ramped-up reactor. The NAVIGATOR concept should be applied in a phased approach to and beyond the operating goal for the FER (Fusion Experimental Reactor, the next generation tokamak machine in Japan). The mission of the FER is to realize self-ignition and a long controlled burn of about 800 seconds and to develop and test fusion technologies, including the tritium fuel cycle, superconducting magnet, remote maintenance and breeding blanket test modules. The NAVIGATOR concept is composed of three major elements, that is, reliable operation scenarios, reliable maintenability and sufficient flexibility of the reactor. The NAVIGATOR concept well supports the ideas of phased operation and phased construction of the FER, which will result in the reduction of technological risk. The NAVIGATOR concept is expected to bring forth the fruits growing up in the present large tokamak machines in the form of next generation machines. In addition, the NAVIGATOR concept will supply many required databases for the DEMO reactor. The details of the NAVIGATOR concept is described in this paper, and the concept may indicate a feasible strategy for developing fusion research. (author)

Tokamak reactor startup power

International Nuclear Information System (INIS)

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

1983-01-01

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

Application of mineral insulated cable (MIC) in Tokamak fusion reactor

International Nuclear Information System (INIS)

Luo Tianyong; Jiang Jiaming; Cen Yishun

2014-01-01

To avoid the instability of plasma and achieve some experimental tasks in Tokamak fusion reactor, many in-vessel coils are designed such as the coils to mitigate the effect of Edge Localized Modes (ELMs coils) and the coils to provide vertical stabilization (VS coils). The in-vessel location presents special challenges in terms of nuclear radiation and temperature, and requires the use of mineral-insulated conductors. The in-vessel coils in ITER are designed to be Mineral-insulated Cable (MIC) with three-layer structures. The inner is hollow-core tube made by OFHC or CuCrZr, the middle is the insulation layer made by Mgo and the outer is the jacket by SS316L or Inconel 718. To control the effect of Edge Localized Modes and vertical instability of plasma, the MIC in-vessel coils shall be used in HL-2M. More details about the application of MIC in Tokamak fusion reactor will be shown in this report. (authors)

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

Beam heating requirements for a tokamak experimental power reactor

International Nuclear Information System (INIS)

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

1976-01-01

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

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-02-01

This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

Fast wave current drive in reactor scale tokamaks

International Nuclear Information System (INIS)

Moreau, D.

1992-01-01

The IAEA Technical Committee Meeting on Fast Wave Current Drive in Reactor Scale Tokamaks, hosted by the Commissariat a l'Energie Atomique (CEA), Departement de Recherches sur la Fusion Controlee (Centres d'Etudes de Cadarache, under the Euratom-CEA Association for fusion) aimed at discussing the physics and the efficiency of non-inductive current drive by fast waves. Relevance to reactor size tokamaks and comparison between theory and experiment were emphasized. The following topics are described in the summary report: (i) theory and modelling of radiofrequency current drive (theory, full wave modelling, ray tracing and Fokker-Planck calculations, helicity injection and ponderomotive effects, and alternative radio-frequency current drive effects), (ii) present experiments, (iii) reactor applications (reactor scenarios including fast wave current drive; and fast wave current drive antennas); (iv) discussion and summary. 32 refs

Neutral-beam-injected tokamak fusion reactors: a review

International Nuclear Information System (INIS)

Jassby, D.L.

1976-08-01

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

Public acceptance of fusion energy and scientific feasibility of a fusion reactor. DREAM (DRastically EAsy Maintenance) tokamak

International Nuclear Information System (INIS)

Nishio, Satoshi

1998-01-01

If the major part of the electric power demand will be supplied by tokamak fusion power plants, a suitable tokamak reactor must be an ultimate goal, i.e., the reactor must be excellent both in terms of construction cost and safety aspects including operation availability (maintainability and reliability). In attaining this goal, an approach focusing on both safety and availability (including reliability and maintainability) issues is the most promising strategy. The tokamak reactor concept with a very high aspect ratio configuration and SiC/SiC composite structural materials is compatible with this approach, which is called the DREAM (DRastically EAsy Maintenance) approach. The SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to good accessibility for the maintenance of machines. As an intermediate steps between an experimental reactor such as ITER and the ultimate goal, the development of prototype reactor which demonstrates electric power generation and an initial-phase commercial reactor which demonstrates for COE (cost of electricity) competitiveness has been investigated. Especially for the prototype reactor, material and technological immaturity must be considered. (J.P.N.)

Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors

International Nuclear Information System (INIS)

Picologlou, B.F.; Reed, C.B.; Hua, T.Q.

1989-01-01

The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab

Steady-state operation requirements of tokamak fusion reactor concepts

International Nuclear Information System (INIS)

Knobloch, A.F.

1991-06-01

In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

International Nuclear Information System (INIS)

Baker, C.C.

1980-01-01

STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described

Compact tokamak reactors. Part 1 (analytic results)

International Nuclear Information System (INIS)

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

1996-01-01

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

Design study of superconducting inductive energy storages for tokamak fusion reactor

International Nuclear Information System (INIS)

1977-08-01

Design of the superconducting inductive energy storages (SC-IES) has been studied. One SC-IES is for the power supply system in a experimental tokamak fusion reactor, and the other in a future practical reactor. Study started with definition of the requirements of SC-IES, followed by optimization of the coil shape and determination of major parameters. Then, the coil and the vessel were designed, including the following: for SC-IES of the experimental reactor, stored energy 10 GJ, B max 8 T, conductor NbTi and size 18 m diameter x 10 m height; for SC-IES of the practical reactor, stored energy 56 GJ, B max 10.5 T, conductor Nb 3 Sn and size 26 m diameter x 15 m height. Design of the coil protection system and an outline of the auxiliary systems (for refrigeration and evacuation) are also given, and further, problems and usefullness of SC-IES. (auth.)

Plasma driving system requirements for commercial tokamak fusion reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

1978-01-01

The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results presented in this paper can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

Plasma driving system requirements for commercial tokamak fusion reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

1977-01-01

The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

A tokamak reactor with servicing capability

International Nuclear Information System (INIS)

Mitchell, J.T.D.; Hollis, A.

1976-01-01

A conceptual design for a Tokamak reactor with practical facilities for the regular replacement of blanket components after the inevitable damage from neutron irradiation, and fatigue is described. This essential facility has been largely ignored in published fusion reactor designs. One exception is the inertially-confined Saturn proposal. Tokamak and other toroidal closed-line systems have very complex geometries and sub-system requirements, which result in blanket servicing being a very difficult problem. In the concept described the magnet shield is divided into two structures - an outer permanent one with access doors and an inner shield, part of and supporting the blanket inside. Servicing access is horizontally between the toroidal magnet coils, after moving some outer poloidal magnet coils. The reactor, reactor hall, workshops and remote-handling facilities are described, and the servicing requirements discussed. The important servicing operation is the remote replacement of radiation damaged blanket and shield - divided in this design into 20 sectors, each weighing 75-100 tons and 11-12 metres high. Analysis of the operation indicates that if one sector can be replaced during a single weekend - i.e. a period of low power demand - then the annual reactor-generator availability allowing as well for the general plant servicing should be >0.9. This level of availability should meet the requirements of generating authorities but the facilities, equipment and workshops necessary may be complex and expensive

Comparative study of cost models for tokamak DEMO fusion reactors

International Nuclear Information System (INIS)

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

2012-01-01

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

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.)

Resistive demountable toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Jassby, D.L.; Jacobsen, R.A.; Kalnavarns, J.; Masson, L.S.; Sekot, J.P.

1981-07-01

Readily demountable TF (toroidal-field) coils allow complete access to the internal components of a tokamak reactor for maintenance of replacement. The requirement of readily demountable joints dictates the use of water-cooled resistive coils, which have a host of decisive advantages over superconducting coils. Previous papers have shown that resistive TF coils for tokamak reactors can operate in the steady state with acceptable power dissipation (typically, 175 to 300 MW). This paper summarizes results of parametric studies of size optimization of rectangular TF coils and of a finite-element stress analysis, and examines several candidate methods of implementing demountable joints for rectangular coils constructed of plate segments

Burnup calculation for a tokamak commercial hybrid reactor

International Nuclear Information System (INIS)

Feng Kaiming; Xie Zhongyou

1990-08-01

A computer code ISOGEN-III and its associated data library BULIB have been developed for fusion-fission hybrid reactor burnup calculations. These are used to calcuate burnup of a tokamak commercial hybrid reactor. The code and library are introduced briefly, and burnup calculation results are given

Maintenance features of the Compact Ignition Tokamak fusion reactor

International Nuclear Information System (INIS)

Spampinato, P.T.; Hager, E.R.

1987-01-01

The Compact Ignition Tokamak (CIT) is envisaged to be the next experimental machine in the US Fusion Program. Its use of deuterium/tritium fuel requires the implementation of remote handling technology for maintenance and disassembly operations. The reactor is surrounded by a close-proximity nuclear shield which is designed to permit personnel access within the test cell, one day after shutdown. With the shield in place, certain maintenance activities in the cell may be done hands-on. Maintenance on the reactor is accomplished remotely using a boom-mounted manipulator after disassembling the shield. Maintenance within the plasma chamber is accomplished with two articulated boom manipulators that are capable of operating in a vacuum environment. They are stored in a vacuum enclosure behind movable shield plugs

Energy balance and efficiency of power stations with a pulsed Tokamak reactor

International Nuclear Information System (INIS)

Davenport, P.A.; Mitchell, J.T.D.; Darvas, J.; Foerster, S.; Sack, B.

1976-06-01

The energy balance of a fusion power station based on the TOKAMAK concept is examined with the aid of a model comprising three distinct elements: the reactor, the energy converter and the reactor operation equipment. The efficiency of each element is expressed in terms of the various energy flows and the product of these efficiencies gives the net station efficiency. The analysis takes account of pulsed operation and has general applicability. Numerical values for the net station efficiency are derived from detailed estimates of the energy flows for a TOKAMAK reactor and its auxiliary equipment operating with advanced energy converters. The derivation of these estimates is given in eleven appendices. The calculated station efficiencies span ranges similar to those quoted for the current generation of fission reactors, though lower than those predicted for HTGR and LMFBR stations. Credible parameter domains for pulsed TOKAMAK operation are firmly delineated and factors inimical to improved performance are indicated. It is concluded that the net thermal efficiency of a TOKAMAK reactor power station based on present designs and using advanced thermal converters will be approximately 0.3 and is unlikely to exceed 0.33. (orig.) [de

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

International Nuclear Information System (INIS)

Ivers, T.H.

1991-01-01

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

Safety and deterministic failure analyses in high-beta D-D tokamak reactors

International Nuclear Information System (INIS)

Selcow, E.C.

1984-01-01

Safety and deterministic failure analyses were performed to compare major component failure characteristics for different high-beta D-D tokamak reactors. The primary focus was on evaluating damage to the reactor facility. The analyses also considered potential hazards to the general public and operational personnel. Parametric designs of high-beta D-D tokamak reactors were developed, using WILDCAT as the reference. The size, and toroidal field strength were reduced, and the fusion power increased in an independent manner. These changes were expected to improve the economics of D-D tokamaks. Issues examined using these designs were radiation induced failurs, radiation safety, first wall failure from plasma disruptions, and toroidal field magnet coil failure

Apollo-L2, an advanced fuel tokamak reactor utilizing direct conversion

International Nuclear Information System (INIS)

Emmert, G.A.; Kulcinski, G.L.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

1989-01-01

A scoping study of a tokamak reactor fueled by a D- 3 He plasma is presented. The Apollo D- 3 He tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The low neutron wall loading (0.1 MW/m 2 ) permits a first wall lasting the life of the plant and enables the reactor to be classified as inherently safe. The cost of electricity is less than that from a similar power level DT reactor. 10 refs., 1 fig., 4 tabs

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Second preliminary design of JAERI experimental fusion reactor (JXFR)

International Nuclear Information System (INIS)

Sako, Kiyoshi; Tone, Tatsuzo; Seki, Yasushi; Iida, Hiromasa; Yamato, Harumi

1979-06-01

Second preliminary design of a tokamak experimental fusion reactor to be built in the near future has been performed. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics radiation shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel recirculating system, reactor cooling and tritium recovery systems and maintenance scheme. Safety analyses of the reactor system have been also performed. This paper gives a brief description of the design as of January, 1979. The feasibility study of raising the power density has been also studied and is shown as appendix. (author)

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

Oak Ridge Tokamak experimental power reactor study reference design

International Nuclear Information System (INIS)

Roberts, M.; Bettis, E.S.

1975-11-01

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

Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

International Nuclear Information System (INIS)

Hancox, R.; Mitchell, J.T.D.

1977-01-01

Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

Reactor costs and maintenance, with reference to the Culham Mark II conceptual Tokamak reactor design

International Nuclear Information System (INIS)

Hancox, R.; Mitchell, J.T.D.

1976-01-01

Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, due mainly to the low power density of the fusion reactor which affects both the reactor and building costs. In order to reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (orig.) [de

Tokamak engineering test reactor

International Nuclear Information System (INIS)

Conn, R.W.; Jassby, D.L.

1975-07-01

The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

Prototype tokamak fusion reactor based on SiC/SiC composite material focusing on easy maintenance

International Nuclear Information System (INIS)

Nishio, S.; Ueda, S.; Kurihara, R.; Kuroda, T.; Miura, H.; Sako, K.; Takase, H.; Seki, Y.; Adachi, J.; Yamazaki, S.; Hashimoto, T.; Mori, S.; Shinya, K.; Murakami, Y.; Senda, I.; Okano, K.; Asaoka, Y.; Yoshida, T.

2000-01-01

If the major part of the electric power demand is to be supplied by tokamak fusion power plants, the tokamak reactor must have an ultimate goal, i.e. must be excellent in construction cost, safety aspect and operational availability (maintainability and reliability), simultaneously. On way to the ultimate goal, the approach focusing on the safety and the availability (including reliability and maintainability) issues must be the more promising strategy. The tokamak reactor concept with the very high aspect ratio configuration and the structural material of SiC/SiC composite is compatible with this approach, which is called the DRastically Easy Maintenance (DREAM) approach. This is because SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to a good accessibility for the maintenance machines. As the intermediate steps along this strategy between the experimental reactor such as international thermonuclear experimental reactor (ITER) and the ultimate goal, a prototype reactor and an initial phase commercial reactor have been investigated. Especially for the prototype reactor, the material and technological immaturities are considered. The major features of the prototype and commercial type reactors are as follows. The fusion powers of the prototype and the commercial type are 1.5 and 5.5 GW, respectively. The major/minor radii for the prototype and the commercial type are of 12/1.5 m and 16/2 m, respectively. The plasma currents for the prototype and the commercial type are 6 and 9.2 MA, respectively. The coolant is helium gas, and the inlet/outlet temperatures of 500/800 and 600/900 deg. C for the prototype and the commercial type, respectively. The thermal efficiencies of 42 and 50% are obtainable in the prototype and the commercial type, respectively. The maximum toroidal field strengths of 18 and 20 tesla are assumed in the prototype and the commercial type, respectively. The thermal

Engineering aspects of a D-D commercial tokamak reactor

International Nuclear Information System (INIS)

Evans, K. Jr.; Baker, C.C.; Brooks, J.N.

1981-01-01

This paper presents some of the engineering aspects of WILDCAT, a conceptual design of a D-D tokamak, fusion reactor. This conceptual design has evolved from initial studies of D-D tokamak reactors, and is intended to be a study of a later-model, commerical fusion reactor in the same sense that STARFIRE was such a study for D-T fuel cycle. The major guidelines of the study have been to utilize as fully as possible the advantages of the D-D fuel cycle but to avoid unnecessary extrapolations of parameters from existing D-T designs, in particular STARFIRE. The paper consists of an overview of the reference design, a description of each of the major engineering systems (rf current drive, burn cycle, impurity control, first wall, blanket/shield, TF magnets, and tritium system, and a summary of conclusions)

Design features of HTMR-hybrid toroidal magnet tokamak reactor

International Nuclear Information System (INIS)

Rosatelli, F.; Avanzini, P.G.; Derchi, D.; Magnasco, M.; Grattarola, M.; Peluffo, M.; Raia, G.; Brunelli, B.; Zampaglione, V.

1984-01-01

The HTMR (Hybrid Toroidal Magnet Tokamak Reactor) conceptual design is aimed to demonstrate the feasibility of a Tokamak reactor which could fulfil the scientific and technological objectives expected from next generation devices with size and costs as small as possible. A hybrid toroidal field magnet, made up by copper and superconducting coils, seems to be a promising solution, allowing a considerable flexibility in machine performances, so as to gain useful margins in front of the uncertainties in confinement time scaling laws and beta and plasma density limits. The optimization procedure for the hybrid magnet, configuration, the main design features of HTMR and the preliminary mechanical calculations of the superconducting toroidal coils are described. (author)

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

International Nuclear Information System (INIS)

Merola, M.; Biggio, M.

1989-01-01

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

Wave-driver options for low-aspect-ratio steady-state tokamak reactors

International Nuclear Information System (INIS)

Ehst, D.A.

1981-02-01

Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A = 2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value

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

International Nuclear Information System (INIS)

Minato, Akio; Tone, Tatsuzo

1985-12-01

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

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

Conceptual design of a Tokamak hybrid power reactor (THPR)

International Nuclear Information System (INIS)

Matsuoka, F.; Imamura, Y.; Inoue, M.; Asami, N.; Kasai, M.; Yanagisawa, I.; Ida, T.; Takuma, T.; Yamaji, K.; Akita, S.

1987-01-01

A conceptual design of a fusion-fission hybrid tokamak reactor has been carried out to investigate the engineering feasibility and promising scale of a commercial hybrid reactor power plant. A tokamak fusion driver based on the recent plasma scaling law is introduced in this design study. The major parameters and features of the reactor are R=6.06 m, a=1.66 m, Ip=11.8 MA, Pf=668 MW, double null divertor plasma and steady state burning with RF current drive. The fusion power has been determined with medium energy multiplication in the blanket so as to relieve thermal design problems and produce electric power around 1000 MW. Uranium silicide is used for the fast fission blanket material to promise good nuclear performance. The coolant of the blanket is FLIBE and the tritium breeding blanket material is Li 2 O ceramics providing breeding ratio above unity

Commercial tokamak reactors with resistive toroidal field magnets

International Nuclear Information System (INIS)

Bombery, L.; Cohn, D.R.; Jassby, D.L.

1984-01-01

Scaling relations and design concepts are developed for commercial tokamak reactors that use watercooled copper toroidal field (TF) magnets. Illustrative parameters are developed for reactors that are scaled up in size from LITE test reactor designs, which use quasi-continuous copper plate magnets. Acceptably low magnet power requirements may be attainable in a moderate beta (β = 0.065) commercial reactor with a major radius of 6.2 m. The shielding thickness and magnet size are substantially reduced relative to values in commercial reactors with superconducting magnets. Operation at high beta (β = 0.14) leads to a reduction in reactor size, magnet-stored energy, and recirculating power. Reactors using resistive TF magnets could provide advantages of physically smaller devices, improved maintenance features, and increased ruggedness and reliability

Analysis of toroidal vacuum vessels for use in demonstration sized tokamak reactors

International Nuclear Information System (INIS)

Culbert, M.E.

1978-07-01

The vacuum vessel component of the tokamak fusion reactor is the subject of this study. The main objective of this paper was to provide guidance for the structural design of a thin wall externally pressurized toroidal vacuum vessel. The analyses are based on the available state-of-the-art analytical methods. The shortcomings of these analytical methods necessitated approximations and assumptions to be made throughout the study. A principal result of the study has been the identification of a viable vacuum vessel design for the Demonstration Tokamak Hybrid Reactor (DTHR) and The Next Step (TNS) Reactor

Design features of HTMR-Hybrid Toroidal Magnet Tokamak Reactor

International Nuclear Information System (INIS)

Rosatelli, F.; Avanzini, P.G.; Brunelli, B.; Derchi, D.; Magnasco, M.; Grattarola, M.; Peluffo, M.; Raia, G.; Zampaglione, V.

1985-01-01

The HTMR (Hybrid Toroidal Magnet Tokamak Reactor) conceptual design is aimed to demonstrate the feasibility of a Tokamak reactor which could fulfill the scientific and technological objectives expected from next generation devices (e.g. INTOR-NET) with size and costs as small as possible. An hybrid toroidal field magnet, made up by copper and superconducting coils, seems to be a promising solution, allowing a considerable flexibility in machine performances, so as to gain useful margins in front of the uncertainties in confinement time scaling laws and beta and plasma density limits. In this paper the authors describe the optimization procedure for the hybrid magnet configuration, the main design features of HTMR and the preliminary mechanical calculations of the superconducting toroidal coils

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.

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

International Nuclear Information System (INIS)

1976-01-01

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

Transients and burn dynamics in advanced tokamak fusion reactors

International Nuclear Information System (INIS)

Mantsinen, M.J.; Salomaa, R.R.E.

1994-01-01

Transient behavior of D 3 He-tokamak reactors is investigated numerically using a zero-dimensional code with prescribed profiles. Pure D 3 He start-up is compared to DT-assisted and DT-ignited start-ups. We have considered two categories of transients which could extinguish steady fusion burn: fuelling interruptions and sudden confinement changes similar to the L → H transients occurring in present-day tokamaks. Shutdown with various current and density ramp-down scenarios are studied, too. (author)

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

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

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

In-vessel maintenance concepts for tokamak fusion reactors

International Nuclear Information System (INIS)

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

1983-01-01

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

Physics design and experimental study of tokamak divertor

International Nuclear Information System (INIS)

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

2007-06-01

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

First preliminary design of an experimental fusion reactor

International Nuclear Information System (INIS)

1977-09-01

A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

High-energy tritium beams as current drivers in tokamak reactors

International Nuclear Information System (INIS)

Mikkelsen, D.R.; Grisham, L.R.

1983-04-01

The effect on neutral-beam design and reactor performance of using high-energy (approx. 3-10 MeV) tritium neutral beams to drive steady-state tokamak reactors is considered. The lower current of such beams leads to several advantages over lower-energy neutral beams. The major disadvantage is the reduction of the reactor output caused by the lower current-drive efficiency of the high-energy beams

Alfven wave heating in a tokamak reactor

International Nuclear Information System (INIS)

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

1990-01-01

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

Diamond Wire Cutting of the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Keith Rule; Erik Perry; Robert Parsells

2003-01-01

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

Wildcat: A commercial deuterium-deuterium tokamak reactor

International Nuclear Information System (INIS)

Evans, K.; Baker, C.C.; Barry, K.M.

1983-01-01

WILDCAT is a conceptual design of a catalyzed deuterium-deuterium tokamak commercial fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing deuterium-tritium (D-T) designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete conceptual design

Tritium experience in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

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

1998-01-01

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

Conceptual designs of power tokamak-type thermonuclear reactors

International Nuclear Information System (INIS)

Shejndlin, A.E.; Nedospasov, A.V.

1978-01-01

Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

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

International Nuclear Information System (INIS)

Abdou, M.A.

1976-01-01

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

Fractional power operation of tokamak reactors

International Nuclear Information System (INIS)

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

1986-01-01

Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

Tokamak fusion reactors with less than full tritium breeding

International Nuclear Information System (INIS)

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

1983-05-01

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

Source-to-incident flux relation for a tokamak fusion test reactor blanket module

International Nuclear Information System (INIS)

Imel, G.R.

1982-01-01

The source-to-incident 14-MeV flux relation for a blanket module on the Tokamak Fusion Test Reactor is derived. It is shown that assumptions can be made that allow an analytical expression to be derived, using point kernel methods. In addition, the effect of a nonuniform source distribution is derived, again by relatively simple point kernel methods. It is thought that the methodology developed is valid for a variety of blanket modules on tokamak reactors

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

International Nuclear Information System (INIS)

Tone, Tatsuzo

1978-07-01

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

Saturation of alpha particle driven instability in Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Gorelenkov, N.N.; Chen, Y.; White, R.B.; Berk, H.L.

1999-01-01

A nonlinear theory of kinetic instabilities near threshold [Berk et al., Plasma Phys. Rep. 23, 842 (1997)] is applied to calculate the saturation level of toroidicity-induced Alfven eigenmodes (TAE), and to be compared with the predictions of δf method calculations (Y. Chen, Ph.D. thesis, Princeton University, 1998). Good agreement is observed between the predictions of both methods and the predicted saturation levels are comparable to experimentally measured amplitudes of the TAE oscillations in Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)]. copyright 1999 American Institute of Physics

Review of the general atomic experimental fusion power reactor initial conceptual design

International Nuclear Information System (INIS)

Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

1976-01-01

The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

Resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1980-11-01

This paper analyzes the optimization of the geometry of resistive TF coils of rectangular bore for tokamak fusion test reactors and practical neutron generators. In examining the trade-offs between geometric parameters and magnetic field for reactors giving a specified neutron wall loading, either the resistive power loss or the lifetime coil cost can be minimized. Aspects of cooling, magnetic stress, and construction are addressed for several reference designs. Bending moment distributions in closed form have been derived for rectangular coils on the basis of the theory of rigid frames. Candidate methods of fabrication and of implementing demountable joints are summarized

Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

International Nuclear Information System (INIS)

Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.; Attenberger, S.E.

1988-01-01

The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor plasma (Tokamak Ignition/Burn Experimental Reactor) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-dimensional transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alpha concentration significantly influence the ignition and steady-state burn capability

START: the creation of a spherical tokamak

International Nuclear Information System (INIS)

Sykes, Alan

1992-01-01

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

ASDEX upgrade - definition of a tokamak experiment with a reactor compatible polaoidal divertor

International Nuclear Information System (INIS)

1982-03-01

ASDEX Upgrade is intended as the next experimental step after ASDEX. It is designed to investigate the physics of a divertor tokamak as closely as possible to fusion reactor requirements, without thermonuclear heating. It is characterized by a poloidal divertor configuration with divertor coils located outside the toroidal field coils, by machine parameters which allow a line density within the plasma boundary sufficient to screen fast CX particles from the plasma core, by a scrape-off layer essentially opaque to neutrals produced at the target plates, and, finally, by an auxiliary heating power high enough for producing a reactor-like power flux density through the plasma boundary. Design considerations on the basis of physical and technical constraints yielded the tokamak system optimized with respect to effort and costs as described in the following. It uses normal-conducting coil systems, is the size of ASDEX, and has a field of 3.9 T, a plasma current of up to 1.5 MA, and a pulse duration of 10 s. To provide the required power flux density, an ICRH power of 10 MW is needed. For comparison, a superconducting version is under investigation. (orig.)

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

Science.gov (United States)

Vdovin, V.

2014-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-02-12

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

Preconceptual design and assessment of a Tokamak Hybrid Reactor

Energy Technology Data Exchange (ETDEWEB)

Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

1980-09-01

The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

Preconceptual design and assessment of a Tokamak Hybrid Reactor

International Nuclear Information System (INIS)

Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

1980-09-01

The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb 3 Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered

Burn cycle requirements comparison of pulsed and steady-state tokamak reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Ehst, D.A.

1983-12-01

Burn cycle parameters and energy transfer system requirements were analyzed for an 8-m commercial tokamak reactor using four types of cycles: conventional, hybrid, internal transformer, and steady state. Not surprisingly, steady state is the best burn mode if it can be achieved. The hybrid cycle is a promising alternative to the conventional. In contrast, the internal transformer cycle does not appear attractive for the size tokamak in question

Tritium production and processing in a Tokamak reactor

International Nuclear Information System (INIS)

Leger, D.

1986-09-01

Important aspects of the tritium system in Tokamak reactors that have to be controlled are overviewed in this paper. The doubling time is one of them, that is to say the time required to produce, in addition to the tritium burned enough tritium to be able to supply the initial tritium inventory. Another one is the tritium permeation through walls. In addition to the permeation phenomena, large tritium inventories are trapped in the reactor structural material. Finally, the different atmospheres of halls, etc.., that can be contaminated with tritium, have to be reprocessed

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.)

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

Concept study of the Steady State Tokamak Reactor (SSTR)

International Nuclear Information System (INIS)

1991-06-01

The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)

Development of large insulator rings for the TOKAMAK Fusion Test Reactor

International Nuclear Information System (INIS)

Brown, T.; Tobin, A.

1977-01-01

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

Tokamak fusion test reactor. Final design report

International Nuclear Information System (INIS)

1978-08-01

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

Minerals resource implications of a tokamak fusion reactor economy

Energy Technology Data Exchange (ETDEWEB)

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

1979-09-01

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

Minerals resource implications of a tokamak fusion reactor economy

International Nuclear Information System (INIS)

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

1979-09-01

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

Design study of a fusion-driven tokamak hybrid reactor for fissile fuel production. Final report

International Nuclear Information System (INIS)

Rose, R.P.

1979-05-01

This study evaluated conceptual approaches for a tokamak fusion-driven fuel producing reactor. The conceptual design of this hybrid reactor was based on using projected state-of-the-art technology for the late 1980s. This reactor would be a demonstration plant and, therefore, first-of-a-kind considerations have been included. The conceptual definitions of two alternatives for the fusion driver were evaluated. A Two-Component Tokamak (TCT) concept, based on the TFTR plasma physics parameters, was compared to a Beam-Driven Thermonuclear (BDTN) concept, based on the USSR T-20 plasma physics parameters

Tokamaks. 2. ed.

International Nuclear Information System (INIS)

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

1997-01-01

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

Engineering feasibility of tight aspect ratio Tokamak (spherical torus) reactors

International Nuclear Information System (INIS)

Peng, Y-K.M.; Hicks, J.B.

1990-01-01

Engineering solutions are identified and analyzed for key high-power-density components of tight aspect ratio tokamak reactors (spherical torus reactors). The potentially extreme divertor heat loads can be reduced to about 3 MW/m 2 in expanded divertors using coils inside the demountable toroidal field coils. Given the long and narrow divertor channels, gaseous divertor targets become possible, which eliminate sputtering and increase the divertor life. The unshielded centre conductor post (CCP) of the toroidal field coil can be made of a single dispersion strengthened copper conductor cooled by high-velocity pressurized water to maintain acceptable copper temperature and strength. Damage and activation of the CCP at a neutron fluence of 10 MW-a/m 2 are also tolerable. Annual replacement of the centre post, the divertor assemblies and the blanket can be accomplished with vertical access for all torus components, which are modularized to reduce size and weight. The technical requirements of these solutions are shown to be comparable with, if not less demanding than, those estimated for conventional tokamak reactors. (author)

Factors affecting the minimum capital cost of a tokamak reactor

International Nuclear Information System (INIS)

Hancox, R.

1981-01-01

The Mk IIA Culham conceptual tokamak reactor design is a 2500 MWe steady-state reactor developed on the basis of a cost optimisation. A revised 1200 MWe conceptual design, the Mk IIB, used a lower wall loading and lower thermodynamic efficiency. A detailed costing of the Mk IIB design, however, showed it to have an unacceptably high capital cost. Since this high cost is a common characteristic of many fusion reactor designs, the cost optimisation of the Mk II design has been reconsidered. (author)

Evaluation of potential blanket concepts for a Demonstration Tokamak Hybrid Reactor

International Nuclear Information System (INIS)

Chapin, D.L.; Chi, J.W.H.; Kelly, J.L.

1978-01-01

An evaluation has been made of several different blanket concepts for use in a near-term Demonstration Tokamak Hybrid Reactor (DTHR), whose main objective would be to produce a significant amount of fissile fuel while demonstrating the feasibility of the tokamak hybrid reactor concept. The desirability of a simple design using proven technology plus a proliferation resistant fuel cycle led to the selection of a low temperature and pressure water-cooled, zircaloy clad ThO 2 blanket concept to breed 233 U. The nuclear performance and thermal-hydraulics characteristics of the blanket were evaluated to arrive at a consistent design. The blanket was found to be feasible for producing a significant amount of fissile fuel even with the limited operating conditions and blanket coverage in the DTHR

Development of large insulator rings for the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Brown, T.; Tobin, A.

1978-01-01

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

FRESCO: fusion reactor simulation code for tokamaks

International Nuclear Information System (INIS)

Mantsinen, M.J.

1995-03-01

The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)

Tokamak Fusion Test Reactor D-T results

International Nuclear Information System (INIS)

Meade, D.M.

1995-01-01

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, α confinement, α heating and possible α-driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of about 20MW of tritium and 14MW of deuterium neutral beams into the TFTR produced a plasma with a T-to-D density ratio of about 1 and yielding a maximum fusion power of about 9.2MW. The fusion power density in the core of the plasma was about 1.8MWm -3 , approximating that expected in a D-T fusion reactor. A TFTR plasma with a T-to-D density ratio of about 1 was found to have about 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass A of τ E ∝A 0.6 . The core ion temperature increased from 30 to 37keV owing to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 to 10.6keV can be attributed to electron heating by the α particles. The approximately 5% loss of α particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined high energy α particles and the resultant α ash density. At fusion power levels of 7.5MW, fluctuations at the toroidal Alfven eigen-mode frequency were observed by the fluctuation diagnostics. However, no additional α loss due to the fluctuations was observed. (orig.)

Lessons learned from the tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)

International Nuclear Information System (INIS)

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

1994-01-01

Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety, and health (ES ampersand H) characteristics of projected tokamak power plants. Summarized herein are the composite conclusions and lessons developed in the course of four conceptual tokamak power-plant designs. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances in both physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advances in materials are also needed for the exploitation of environmental advantages otherwise inherent in fusion power

Lessons learned from the Tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)

International Nuclear Information System (INIS)

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

1994-01-01

Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety and health (ES ampersand H) characteristics of projected tokamak power plants. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances relative to present understanding in physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advanced tokamak plasmas configured in the second-stability regime that achieve both high β and bootstrap fractions near unity through strong profile control offer high promise in this regard

Decontamination and decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Walton, G.R.; Perry, E.D.; Commander, J.C.; Spampinato, P.T.

1994-01-01

The Tokamak Fusion Test Reactor (TFTR) is scheduled to complete its end-of-life deuterium-tritium (D-T) experiments in September 1994. The D-T operation will result in the TFTR machine structure becoming activated, and plasma facing and vacuum components will be contaminated with tritium. The resulting machine activation levels after a two year cooldown period will allow hands on dismantling for external structures, but require remote dismantling for the vacuum vessel. The primary objective of the Decontamination and Decommissioning (D ampersand D) Project is to provide a facility for construction of a new Department of Energy (DOE) experimental fusion reactor by March 1998. The project schedule calls for a two year shutdown period when tritium decontamination of the vacuum vessel, neutral beam injectors and other components will occur. Shutdown will be followed by an 18 month period of D ampersand D operations. The technical objectives of the project are to: safely dismantle and remove components from the test cell complex; package disassembled components in accordance with applicable regulations; ship packages to a DOE approved disposal or material recycling site; and develop expertise using remote disassembly techniques on a large scale fusion facility. This paper discusses the D ampersand D objectives, the facility to be decommissioned, and the technical plan that will be implemented

Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

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

1999-01-01

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

Controlled thermonuclear fusion in TOKAMAK type reactors, the European example: Joint European Torus (JET)

International Nuclear Information System (INIS)

Paris, P.J.; Yassen, F.; Assis, A.S. de; Raposo, C.

1988-07-01

The development of controlled thermonuclear reaction in TOKAMAK type reactors, and the main projects in the world are presented. The main characteristics of the JET (Joint European Torus) program, the perspectives for energy production, and the international cooperation for viable use of the TOKAMAK are analysed. (M.C.K.) [pt

Internal transport barrier formation and pellet injection simulation in helical and tokamak reactors

International Nuclear Information System (INIS)

Higashiyama, You; Yamazaki, Kozo; Arimoto, Hideki; Garcia, Jeronimo

2008-01-01

In the future fusion reactor, plasma density peaking is important for increase in the fusion power gain and for achievement of confinement improvement mode. Density control and internal transport barrier (ITB) formation due to pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. First, pellet injection simulation is carried out, including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of high-field side (HFS) pellet injection is clarified. Second, ITB simulation with pellet injection is carried out with the confinement improvement model based on the E x B shear effects, and it is found that deep pellet penetration is helpful for ITB formation as well as plasma core fuelling in the reversed-shear tokamak and helical reactors. (author)

Blanket concepts for the ARIES commercial tokamak reactor study

International Nuclear Information System (INIS)

Grotz, S.P.; Ghoniem, N.M.; Hasan, M.Z.; Martin, R.C.; Najmabadi, F.; Sharafat, S.; Hua, T.; Sze, D.K.; Cheng, E.T.; Creedon, R.L.; Wong, C.P.C.; Herring, J.S.; Klein, A.; Snead, L.; Steiner, D.

1989-01-01

The ARIES study is a 3-year effort, started in 1988, exploring the potential of the tokamak to be an attractive and competitive commercial power reactor. Several different versions of the tokamak are being considered, combining different levels of extrapolations in physics and engineering databases. The first version studied in detail, ARIES-I, combines present-day physics (with minimal extrapolation) with aggressive engineering technology such as very high-field, superconducting magnets and low-activation silicon carbide composite materials. The ARIES-I version is designed to meet acceptable safety and environmental criteria. In particular, achieving a passively safe concept that meets Class-C waste disposal is one of the high leverage items in the design. This paper summarizes the scoping analysis and engineering design of the ARIES-I fusion-power-core subsystems. The ARIES-I design is a 1000 MW e power reactor, operating at steady state in the 1 st stability regime and uses a high magnetic field. Typical operating parameters of the ARIES-I strawman design are listed

Ripple induced trapped particle loss in tokamaks

International Nuclear Information System (INIS)

White, R.B.

1996-05-01

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

Computational model for superconducting toroidal-field magnets for a tokamak reactor

International Nuclear Information System (INIS)

Turner, L.R.; Abdou, M.A.

1978-01-01

A computational model for predicting the performance characteristics and cost of superconducting toroidal-field (TF) magnets in tokamak reactors is presented. The model can be used to compare the technical and economic merits of different approaches to the design of TF magnets for a reactor system. The model has been integrated into the ANL Systems Analysis Program. Samples of results obtainable with the model are presented

Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

International Nuclear Information System (INIS)

Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

1980-09-01

A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

Tokamak hybrid thermonuclear reactor for the production of fissionable fuel and electric power

International Nuclear Information System (INIS)

Velikhov, E.P.; Glukhikh, V.A.; Gur'ev, V.V.

1978-01-01

The results of feasibility studies of a tokamak- based hybrid reactor concept are presented. The system selected has a D-T plasma volume of 575 m 3 with additional plasma heating by injection of fast neutral particles. The method of heating makes it possible to achieve an economical two-component tokamak regime at ntau=(4-6)x10 13 sxcm -3 , i e. far below the Lawson criterion. Plasma and vacuum chamber are surrounded by a blanket where fissionable plutonium is produced and heat transformed into electric power is generated. Major plasma-neutron-physical characteristics of the 6905 MWth (2500 MWe) reactor and its electromagnetic system are presented. Evaluations show that the hybrid reactor can produce about 800 kg of Pu per 1GWth/yr as compared to 70-150 kg of Pu for fast breeder reactors. The increased Pu production rate is the major merit of the concept promising for both power generation and fuelling thermal fission reactions

Operation and control of high density tokamak reactors

International Nuclear Information System (INIS)

Attenberger, S.E.; McAlees, D.G.

1976-01-01

The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

Startup and shutdown of the PULSAR Tokamak Reactor

International Nuclear Information System (INIS)

Werley, K.A.; Bathke, C.G.

1994-01-01

Start-up conditions are examined for a pulsed tokamak reactor that uses only inductive plasma current drive for startup, burn and shutdown. A zero-dimensional (profile-averaged) model that describes plasma power and particle balance equations is used to study several aspects of plasma startup and shutdown, including optimization of the startup pathway tradeoff of auxiliary startup heating power versus startup time, volt-second consumtion, thermal stability and partial-power operations

Tokamak power systems studies, FY 1986: A second stability power reactor

International Nuclear Information System (INIS)

Ehst, D.; Baker, C.; Billone, M.

1987-03-01

This report presents the results of the work at Argonne National Laboratory (ANL) during FY-1986 on the Tokamak Power Systems Study (TPSS). The purpose of the TPSS is to explore and develop ideas that would lead to improvements in the tokamak as a power reactor concept. The work at ANL concentrated on plasma engineering, impurity control, and the blanket/first wall/shield system. The work in FY-1986 extended these studies and focused them on a reference design point. The key features of the design point include: second stability regime with higher β and larger aspect ratio, steady-state operation with fast wave current drive, impurity control via a self-pumped slot limiter, a self-cooled liquid lithium, vanadium alloy blanket with simplified poloidal flow, and reduced reactor building volume with vertical lift maintenance. Sufficient work was carried out to report a preliminary cost estimate. In addition, reactor implications of steady-state operation in the first stability regime were also studied. 174 refs., 124 figs., 65 tabs

Tokamak concept innovations

International Nuclear Information System (INIS)

1986-04-01

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

Study of an optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak

Energy Technology Data Exchange (ETDEWEB)

Hong, Bong Guen, E-mail: bghong@jbnu.ac.kr [Department of Quantum System Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of); Kim, Hoseok [Department of Applied Plasma Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of)

2016-11-15

Highlights: • Optimum configuration of a transmutation reactor based on a low aspect ratio tokamak was found. • Inboard and outboard radial build are determined by plasma physics, engineering and neutronics constraints. • Radial build and equilibrium fuel cycle play a major role in determining the transmutation characteristics. - Abstract: We determine the optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak. For self-consistent determination of the radial build of the reactor components, we couple a tokamak systems analysis with a radiation transport calculation. The inboard radial build of the reactor components is obtained from plasma physics and engineering constraints, while outboard radial builds are mainly determined by constraints on neutron multiplication, the tritium-breeding ratio, and the power density. We show that the breeding blanket model has an effect on the radial build of a transmutation blanket. A burn cycle has to be determined to keep the fast neutron fluence plasma-facing material below its radiation damage limit. We show that the radial build of the transmutation reactor components and the equilibrium fuel cycle play a major role in determining the transmutation characteristics.

The role of the neutral beam fueling profile in the performance of the Tokamak Fusion Test Reactor and other tokamak plasmas

International Nuclear Information System (INIS)

Park, H.K.; Batha, S.

1997-02-01

Scalings for the stored energy and neutron yield, determined from experimental data are applied to both deuterium-only and deuterium-tritium plasmas in different neutral beam heated operational domains in Tokamak Fusion Test Reactor. The domain of the data considered includes the Supershot, High poloidal beta, Low-mode, and limiter High-mode operational regimes, as well as discharges with a reversed magnetic shear configuration. The new important parameter in the present scaling is the peakedness of the heating beam fueling profile shape. Ion energy confinement and neutron production are relatively insensitive to other plasma parameters compared to the beam fueling peakedness parameter and the heating beam power when considering plasmas that are stable to magnetohydrodynamic modes. However, the stored energy of the electrons is independent of the beam fueling peakedness. The implication of the scalings based on this parameter is related to theoretical transport models such as radial electric field shear and Ion Temperature Gradient marginality models. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

Systems study of tokamak fusion--fission reactors

International Nuclear Information System (INIS)

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

1978-11-01

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

Conceptual design studies of experimental and demonstration fusion reactors

International Nuclear Information System (INIS)

1978-01-01

Since 1973 the FINTOR Group has been involved in conceptual design studies of TOKAMAK-type fusion reactors to precede the construction of a prototype power reactor plant. FINTOR-1 was the first conceptual design aimed at investigating the main physics and engineering constraints on a minimum-size (both dimensions and thermal power) tokamak experimental reactor. The required plasma energy confinement time as evaluated by various power balance models was compared with the values resulting from different transport models. For the reference design, an energy confinement time ten times smaller than neoclassical was assumed. This also implied a rather high (thermally stable) working temperature (above 20 keV) for the reactor. Other relevant points of the design were: circular plasma cross section, single-null axisymmetric divertor; lithium breeder, stainless steel structures, helium coolant; modular blanket and shield structure; copper-stabilized, superconducting Nb-Ti toroidal field and divertor coils; vertical field and transformer coils inside the toroidal coils; vacuum-tight containment vessel. Solutions involving air and iron transformer cores were compared. These assumptions led to a minimum size reactor with a thermal power of about 100MW and rather large dimensions (major radius of about 9m) similar to those of full-scale power reactors considered in other conceptual studies. The FINTOR-1 analysis was completed by the end of 1976. In 1977 a conceptual design of a Demonstration Power Reactor Plant (FINTOR-D) was started. In this study the main working assumptions differing from those of FINTOR-1 are: non-circular plasma cross section; plasma confinement compatible with trapped ion instabilities; cold (gas) blanket sufficient for wall protection (no divertor); wall loading between 1-3MW/m 2 and thermal power of a few GW. (author)

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

International Nuclear Information System (INIS)

Miller, R.L.

1989-01-01

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

UWMAK-II: a conceptual tokamak reactor design

International Nuclear Information System (INIS)

1975-10-01

This report describes the conceptual design of a Tokamak fusion power reactor, UWMAK-II. The aim of this study is to perform a self consistent and thorough analysis of a probable future fusion power reactor in order to assess the technological problems posed by such a system and to examine feasible solutions. UWMAK-II is a conceptual Tokamak fusion reactor designed to deliver 1716 MWe continuously and to generate 5000 MW(th) during the plasma burn. The structural material is 316 stainless steel and the primary coolant is helium. UWMAK-II is a low aspect ratio, low field design and includes a double null, axisymmetric poloidal field divertor for impurity control. In addition, a carbon curtain, made of two dimensional woven carbon fiber, is mounted on the first vacuum chamber wall to protect the plasma from high Z impurities and to protect the first wall from erosion by charged particle bombardment. The blanket is designed to minimize the inventory of both tritium and lithium while achieving a breeding ratio greater than one. This has led to a blanket design based on the use of a solid breeding material (LiAlO 2 ) with beryllium as a neutron multiplier. The lithium is enriched to 90 percent 6 Li and the blanket coolant is helium at a maximum pressure of 750 psia (5.2 x 10 6 N/m 2 ). A cell of the UWMAK-II blanket design is shown. The breeding ratio is between 1.11 and 1.19 based on one-dimensional discrete ordinates transport calculations, depending on the method of homogenization. Detailed Monte Carlo calculations, which take into account the more complicated geometry, give a breeding ratio of 1.06. The total energy per fusion is 21.56 MeV, which is fairly high

Vulcan: A steady-state tokamak for reactor-relevant plasma–material interaction science

International Nuclear Information System (INIS)

Olynyk, G.M.; Hartwig, Z.S.; Whyte, D.G.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Mumgaard, R.T.; Podpaly, Y.A.

2012-01-01

Highlights: ► A new scaling for obtaining reactor similarity in the divertor of scaled tokamaks. ► Conceptual design for a tokamak (“Vulcan”) to implement this new scaling. ► Demountable superconducting coils and compact neutron shielding. ► Helium-cooled high-temperature vacuum vessel and first wall. ► High-field-side lower hybrid current drive for non-inductive operation. - Abstract: An economically viable magnetic-confinement fusion reactor will require steady-state operation and high areal power density for sufficient energy output, and elevated wall/blanket temperatures for efficient energy conversion. These three requirements frame, and couple to, the challenge of plasma–material interaction (PMI) for fusion energy sciences. Present and planned tokamaks are not designed to simultaneously meet these criteria. A new and expanded set of dimensionless figures of merit for PMI have been developed. The key feature of the scaling is that the power flux across the last closed flux surface P/S ≃ 1 MW m −2 is to be held constant, while scaling the core volume-averaged density weakly with major radius, n ∼ R −2/7 . While complete similarity is not possible, this new “P/S” or “PMI” scaling provides similarity for the most critical reactor PMI issues, compatible with sufficient current drive efficiency for non-inductive steady-state core scenarios. A conceptual design is developed for Vulcan, a compact steady-state deuterium main-ion tokamak which implements the P/S scaling rules. A zero-dimensional core analysis is used to determine R = 1.2 m, with a conventional reactor aspect ratio R/a = 4.0, as the minimum feasible size for Vulcan. Scoping studies of innovative fusion technologies to support the Vulcan PMI mission were carried out for three critical areas: a high-temperature, helium-cooled vacuum vessel and divertor design; a demountable superconducting toroidal field magnet system; and a steady-state lower hybrid current drive system

Local Physics Basis of Confinement Degradation in JET ELMy H-Mode Plasmas and Implications for Tokamak Reactors

International Nuclear Information System (INIS)

Budny, R.V.; Alper, B.; Borba, D.; Cordey, J.G.; Ernst, D.R.; Gowers, C.

2001-01-01

First results of gyrokinetic analysis of JET [Joint European Torus] ELMy [Edge Localized Modes] H-mode [high-confinement modes] plasmas are presented. ELMy H-mode plasmas form the basis of conservative performance predictions for tokamak reactors of the size of ITER [International Thermonuclear Experimental Reactor]. Relatively high performance for long duration has been achieved and the scaling appears to be favorable. It will be necessary to sustain low Z(subscript eff) and high density for high fusion yield. This paper studies the degradation in confinement and increase in the anomalous heat transport observed in two JET plasmas: one with an intense gas puff and the other with a spontaneous transition between Type I to III ELMs at the heating power threshold. Linear gyrokinetic analysis gives the growth rate, gamma(subscript lin) of the fastest growing modes. The flow-shearing rate omega(subscript ExB) and gamma(subscript lin) are large near the top of the pedestal. Their ratio decreases approximately when the confinement degrades and the transport increases. This suggests that tokamak reactors may require intense toroidal or poloidal torque input to maintain sufficiently high |gamma(subscript ExB)|/gamma(subscript lin) near the top of the pedestal for high confinement

Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

International Nuclear Information System (INIS)

Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

1986-06-01

Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost

Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

International Nuclear Information System (INIS)

Sink, D.A.; Gibson, G.

1979-03-01

The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based on two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Design study of blanket structure for tokamak experimental fusion reactor

International Nuclear Information System (INIS)

1979-11-01

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

Resistive requirements for the vacuum wall of a tokamak fusion reactor

International Nuclear Information System (INIS)

Brooks, J.N.; Ehat, D.; Harkness, S.D.; Norem, J.; Stevens, H.; Turner, L.

1978-01-01

Most conceptual designs of tokamak power reactors have incorporated a ceramic insulator in the vacuum wall to make the wall electrically non-conducting. Such a material will have to be highly resistant to radiation damage at doses up to at least 10 MW-yr/m 2 while being compatible with a coolant and a first wall whose dimensions change due to thermal cycling and radiation damage. Thus there is considerable incentive to assess the consequences of eliminating the flux breaker from the design and having a conducting boundary instead. In this initial study the question of having a finite wall resistance has been examined in terms of its major implications on both the normal and abnormal operation of a tokamak reactor. This study has been conducted within the framework of the ANL-EPR-77 design although the results should provide some guidance for future reactors as well. The EPR design referred to is a 5 m major radius tokamak with an aspect ratio of 3.5, and with an equilibrium plasma current of 7.3 MA. The vacuum chamber is designed to accommodate a non-circular plasma with a height to width ratio of up to 1.65. The basic vacuum wall design is shown in Fig. 1. It is located about 0.4 M from the plasma boundary and has an irregular polygon shape made of sixteen sections, one per TF coil interval. Variations of this design having a range of resistance values have been used in the analysis

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

International Nuclear Information System (INIS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-01-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

A design study of superconducting energy storage system for a tokamak fusion reactor

International Nuclear Information System (INIS)

Ueda, Kazuo

1979-01-01

A design study of a superconducting inductive energy storage system (SC-IES) has been carried out in commission with JAERI. The SC-IES is to be applied to the power supply system for a tokamak experimental fusion reactor. The study was initiated with the definition of the requirement for the SC-IES and selection of the coil shape. The design of the coil and the cryostat has been followed. The design parameters are: stored energy 10 GJ, B max 8 T, conductor Nb-Ti, overall size 18 m (diameter) x 10 m (height). Technical problems and usefullness of SC-IES are discussed also. (author)

Deuterium-tritium experiments on the Tokamak Fusion Test reactor

International Nuclear Information System (INIS)

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

1994-09-01

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

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

International Nuclear Information System (INIS)

Ford, G.W.K.

1987-01-01

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

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

Directions for attractive tokamak reactors: The ARIES-II and ARIES-IV second-stability designs

International Nuclear Information System (INIS)

Najmabadi, F.; Conn, R.W.

1993-01-01

ARIES is a research program to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The ARIES study has developed four visions for tokamaks. All four designs are steady-state, 1000-MWe (net) power reactors. The ARIES-II and ARIES-IV designs assume potential advances in plasma physics (such as second-stability operation) predicted by theory but not yet established experimentally. The two designs have the same fusion plasma but different fusion-power-core. There are only minor differences between the ARIES-II and ARIES-IV plasma parameters. ARIES-IV is a 1000-MWe reactor with an average neutron wall loading of 3 MW/m 2 , and a mass power density of about 120 kWe/tonne of fusion power core. The reactor major radius is 6.1 m, the plasma minor radius is 1.5 m and the plasma elongation is 2, and the plasma triangularity is 0.67. The plasma current is low (6.8 MA), B on-axis is 7.7 T (corresponding to a maximum field at the coil of 16T), and the toroidal beta is 3.4% (Troyon coefficient = 6). The operating regime is optimized such that most of the plasma current (∼ 90%) is provided by the bootstrap current. ARIES-II uses liquid lithium as the coolant and tritium breeder. V-5Cr-5Ti is used as the structural material so that the potential of low-activation metallic blankets can be studied. ARIES-IV uses helium as the coolant, a solid tritium-breeding material (Li 2 O), and silicon carbide composite as structural material. The waste produced by neutron activation in both designs is found to meet the criteria allowing shallow-land burial under U.S. regulations. The cost of electricity for the ARIES-II-IV class of reactors is estimated to be about 20% lower than comparable, steady-state first-stability reactors (e.g. ARIES-I). 25 refs, 2 figs, 1 tab

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

Ignition and time-dependent fractional power operation of tokamak reactors

International Nuclear Information System (INIS)

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

1986-01-01

The eventual utilization of a tokamak fusion reactor for commercial power necessitates a thorough understanding of the operational requirements at full and fractional power levels and during transitions from one operating level to another. In this study we examine the role of burn control in maintaining the reactor plasma at equilibrium to avoid thermal runaway during fractional power operation. Because these requirements rely so heavily on the assumptions that govern the plasma transport, this study focuses on time-dependent analyses and a comparison of ignition requirements using a range of energy confinement

Radiation transport effects in divertor plasmas generated during a tokamak reactor disruption

International Nuclear Information System (INIS)

Peterson, R.R.; MacFarlane, J.J.; Wang, P.

1994-01-01

Vaporization of material from tokamak divertors during disruptions is a critical issue for tokamak reactors from ITER to commercial power plants. Radiation transport from the vaporized material onto the remaining divertor surface plays an important role in the total mass loss to the divertor. Radiation transport in such a vapor is very difficult to calculate in full detail, and this paper quantifies the sensitivity of the divertor mass loss to uncertainties in the radiation transport. Specifically, the paper presents the results of computer simulations of the vaporization of a graphite coated divertor during a tokamak disruption with ITER CDA parameters. The results show that a factor of 100 change in the radiation conductivity changes the mass loss by more than a factor of two

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.

Energy transport requirements for tokamak reactors in the second ballooning stability regime

International Nuclear Information System (INIS)

Potok, R.E.; Bromberg, L.; Cohn, D.R.

1986-01-01

The authors present an analysis of ignition confinement constraints on a tokamak reactor operating in the second regime of ballooning stability. This regime is characterized by flat plasma pressure profiles, with a sharp pressure gradient near a conducting first wall at the plasma edge. The energy confinement time is determined by transport processes across the pressure gradient region. The authors have found that the required transport needed for ignition in the edge region is very close to the value predicted by neoclassical ion conductivity scaling. Only by carefully tailoring the conductivity scaling across the flux coordinate were the authors able to match both the kink stability and ignition requirements. With optimistic assumptions, R/sub o/ ≅ 7 m appears to be the minimum major radius for an economical tokamak reactor in the second ballooning stability regime. This paper presents a base design case at R/sub o/ = 7 m, and shows how the reactor design varies with changes in major radius, ion transport scaling, and electron transport scaling

Application of high temperature ceramic superconductors (CSC) to commercial tokamak reactors

International Nuclear Information System (INIS)

Ehst, D.A.; Kim, S.; Gohar, Y.; Turner, L.; Smith, D.L.; Mattas, R.

1988-08-01

Ceramic superconductors operating near liquid nitrogen temperature may experience higher heating rates without losing stability, compared conventional superconductors. This will permit cable design with less stabilizer, reducing fabrication costs for large fusion magnets. Magnet performance is studied for different operating current densities in the superconductor, and cost benefits to commercial tokamak reactors are estimated. It appears that 10 kA /center dot/ cm/sup /minus/2/ (at 77 K and /approximately/10 T) is a target current density which must be achieved in order for the ceramic superconductors to compete with conventional materials. At current densities around 50 kA /center dot/ cm/sup /minus/2/ most potential benefits have already been gained, as magnet structural steel begins to dominate the cost at this point. For a steady state reactor reductions of /approximately/7% are forecast for the overall capital cost of the power plant in the best case. An additional /approximately/3% cost saving is possible for pulsed tokamaks. 9 refs., 4 figs., 8 tabs

Application of high temperature ceramic superconductors (CSC) to commercial tokamak reactors

International Nuclear Information System (INIS)

Ehst, D.A.; Kim, S.; Gohar, Y.; Turner, L.; Smith, D.L.; Mattas, R.

1987-10-01

Ceramic superconductors operating near liquid nitrogen temperature may experience higher heating rates without losing stability, compared to conventional superconductors. This will permit cable design with less stabilizer, reducing fabrication costs for large fusion magnets. Magnet performance is studied for different operating current densities in the superconductor, and cost benefits to commercial tokamak reactors are estimated. It appears that 10 kA . cm -2 (at 77 K and ∼10 T) is a target current density which must be achieved in order for the ceramic superconductors to compete with conventional materials. At current densities around 50 kA . cm -2 most potential benefits have already been gained, as magnet structural steel begins to dominate the cost at this point. For a steady state reactor reductions of ∼7% are forecast for the overall capital cost of the power plant in the best case. An additional ∼3% cost saving is possible for pulsed tokamaks. 9 refs., 4 figs., 8 tabs

Neutronics design for a spherical tokamak fusion-transmutation reactor

International Nuclear Information System (INIS)

Deng Meigen; Feng Kaiming; Yang Bangchao

2002-01-01

Based on studies of the spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. By using the one-dimension transport and burn-up code BISON3.0 to process optimized design, a set of plasma parameters and blanket configuration suitable for the transmutation of MA (Minor Actinides) nuclear waste is selected. Based on the one-dimension calculation, two-dimension calculation has been carried out by using two-dimension neutronics code TWODANT. Combined with the neutron flux given by TWODANT calculation, burn-up calculation has been processed by using the one-dimension radioactivity calculation code FDKR and some useful and reasonable results are obtained

TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

International Nuclear Information System (INIS)

Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

1986-09-01

Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m 2 and a surface heat flux of 1 MW/m 2 . The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO 2 rods. The helium coolant pressure is 5 MPa, entering the module at 297 0 C and exiting at 550 0 C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-09-01

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

Summary report on tokamak confinement experiments

International Nuclear Information System (INIS)

1982-03-01

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

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1985-01-01

The Fusion Experimental Reactor (FER) being developed at JAERI as a next generation tokamak to JT-60 has a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. During FY82 and FY83 a comprehensive and intensive conceptual design study has been conducted for a pulsed operation FER as a reference option which employs a conventional inductive current drive and a double-null divertor. In parallel with the reference design, studies have been carried out to evaluate advanced reactor concepts such as quasi-steady state operation and steady state operation based on RF current drive and pumped limiter, and comparative studies for single-null divertor/pumped limiter. This report presents major results obtained primarily from FY83 design studies, while the results of FY82 design studies are described in previous references (JAERI-M 83-213--216). (author)

Conceptual integrated approach for the magnet system of a tokamak reactor

International Nuclear Information System (INIS)

Duchateau, J.-L.; Hertout, P.; Saoutic, B.; Artaud, J.-F.; Zani, L.; Reux, C.

2014-01-01

Highlights: • We give a conceptual approach of a fusion reactor magnet system based on analytical formula. • We give design criteria for the CS and TF cable in conduit conductors and for the magnet system structural description. • We apply this conceptual approach to ITER and we crosscheck with actual characteristics. • We apply this conceptual approach to a possible version of DEMO. - Abstract: In the framework of the reflexion about DEMO, a conceptual integrated approach for the magnet system of a tokamak reactor is presented. This objective is reached using analytical formulas which are presented in this paper, coupled to a Fortran code ESCORT (Electromagnetic Superconducting System for the Computation of Research Tokamaks), to be integrated into SYCOMORE, a code for reactor modelling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modelling task force. The analytical formulas deal with all aspects of the magnet system, starting from the derivation of the TF system general geometry, from the plasma main characteristics. The design criteria for the cable current density and the structural design of the toroidal field and central solenoid systems are presented, enabling to deliver the radial thicknesses of the magnets and enabling also to estimate the plasma duration of the plateau. As a matter of fact, a pulsed version DEMO is presently actively considered in the European programmes. Considerations regarding the cryogenics and the protection are given, affecting the general design. An application of the conceptual approach is presented, allowing a comparison between ESCORT output data and actual ITER parameters and giving the main characteristics of a possible version for DEMO

Some considerations on a plasma in the JAERI experimental fusion reactor

International Nuclear Information System (INIS)

Tone, T.; Yamato, H.; Maki, K.

1976-01-01

The preliminary analysis of the plasma characteristics for the JAERI tokamak experimental fusion reactor is reported. In order to make the reactor compact, the self-sustaining condition has been removed. Stationary heating by 200 keV neutral deuteron beam to maintain the power balance is applied expecting the power amplification by the TCT effect. The main parameters determined are power output of 100 MW, toroidal field on axis of 6 T, aspect ratio of 4.5 and major radius of 6.75 m. The results of the plasma power balance, fueling by means of the gas blanket scheme, power stabilization with feedback and the start-up are presented

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

International Nuclear Information System (INIS)

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

1991-12-01

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

Repair/maintenance design for tokamak experimental fusion reactor

International Nuclear Information System (INIS)

1978-10-01

Repair and maintenance design for JXFR has been studied. The reactor is in eight modules so that a damaged module alone can be separated from the other modules and transferred from the reactor room to a repair shop. Design work covers overhaul procedure, dismounting equipments (overhead cranes, auto welder/cutter and remote handling equipments), transport system of a module (module mounting carriages and rotating carriage), repair equipment for blanket, earthquake-proof analysis of the reactor, reactor room structure, repair shop layout, management of radioactive wastes, time and the number of persons required for overhaul etc. Though the repair and maintenance system is almost complete, there still remain problems for further study in joints of blanket cooling piping, auto welder/cutter and earthquake-proof strength in reactor disassemblage. More detailed studies and R and D are necessary for engineering perfection. (author)

Prospects for toroidal fusion reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.D.

1994-01-01

Work on the International Thermonuclear Experimental Reactor (ITER) tokamak has refined understanding of the realities of a deuterium-tritium (D-T) burning magnetic fusion reactor. An ITER-like tokamak reactor using ITER costs and performance would lead to a cost of electricity (COE) of about 130 mills/kWh. Advanced tokamak physics to be tested in the Toroidal Physics Experiment (TPX), coupled with moderate components in engineering, technology, and unit costs, should lead to a COE comparable with best existing fission systems around 60 mills/kWh. However, a larger unit size, ∼2000 MW(e), is favored for the fusion system. Alternative toroidal configurations to the conventional tokamak, such as the stellarator, reversed-field pinch, and field-reversed configuration, offer some potential advantage, but are less well developed, and have their own challenges

Physics analysis of the Apollo D-3He tokamak reactor

International Nuclear Information System (INIS)

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

1990-01-01

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

Concept design on RH maintenance of CFETR Tokamak reactor

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Songtao; Wan, Yuanxi; Li, Jiangang; Ye, Minyou; Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua

2014-01-01

Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed

Analytic description of tokamak equilibrium sustained by high fraction bootstrap current

International Nuclear Information System (INIS)

Shi Bingren

2002-01-01

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

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

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

1992-01-01

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

Power supply requirements for a tokamak fusion reactor

International Nuclear Information System (INIS)

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

1979-02-01

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

Tokamak reactor designs as a function of aspect ratio

International Nuclear Information System (INIS)

Wong, C.P.C.; Stambaugh, R.D.

2000-01-01

This paper assesses the technical and economic potential of tokamak power plants which utilize superconducting coil (SC) or normal conducting coil (NC) designs as a function of aspect ratio (A). Based on the results from plasma equilibrium calculations, the key physics design parameters of β N , β p , β T , and κ were fitted to parametric equations covering A in the range of 1.2-6. By using ARIES-RS and ARIES-ST as reference design points, a fusion reactor system code was used to project the performance and cost of electricity (COE) of SC and NC reactor designs over the same range of A. The principle difference between the SC and the NC designs are the inboard standoff distance between the coil and the inboard first wall, and the maximum central column current density used for respective coil types. Results show that at an output power of 2 GWe both NC and SC designs can project COE in the respectable range of 62-65 mill/kW h at gross thermal efficiency of 46%, with neutron wall loading (Γ n ) ∼7 MW/m 2 . More importantly, we have learned that based on the present knowledge of equilibrium physics and fusion power core components and system design we can project the performance and COE of reactor designs at least for the purpose of comparative assessment. Tokamak design points can then be selected and optimized for testing or commercial devices as a function of output power, A and Γ n for both SC and NC design options

TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

International Nuclear Information System (INIS)

Hatch, A.J.

1975-01-01

1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

Superconducting coil design for a tokamak experimental power reactor

International Nuclear Information System (INIS)

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

1977-01-01

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

Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

1993-01-01

The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

Decommissioning of the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Perry, E.; Chrzanowski, J.; Gentile, C.; Parsells, R.; Rule, K.; Strykowsky, R.; Viola, M.

2003-01-01

The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

Plasma engineering analyses of tokamak reactor operating space

International Nuclear Information System (INIS)

Houlberg, W.; Attenberger, S.E.

1981-01-01

A comprehensive method is presented for analyzing the potential physics operating regime of fusion reactor plasmas with detailed transport codes. Application is made to the tokamak Fusion Engineering Device (FED). The relationships between driven and ignited operation and supplementary heating requirements are examined. The reference physics models give a finite range of density and temperature over which physics objectives can be reached. Uncertainties in the confinement scaling and differences in supplementary heating methods can expand or contract this operating regime even to the point of allowing ignition with the more optimistic models

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Scoping studies of tritium handling in a tokamak experimental power reactor

International Nuclear Information System (INIS)

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

1975-01-01

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

On some interesting properties of the working temperature in a Tokamak reactor

International Nuclear Information System (INIS)

Brunelli, B.

1987-01-01

A D,T burning plasma has two equilibrium temperatures T/sub 1/ and T/sub 2/ wherein power-in equals power-out. At marginal ignition: T/sub 2/ = T/sub 1/ = T/sub 0/. It is shown that, under hypothesis usually satisfied in a Tokamak reactor, the temperature T/sub 0/ has a peculiar behaviour with respect to the reactor parameters. Simple expressions are given for T/sub 0/ T/sub 1/ and T/sub 2/ which have been found quite straightforward for a well-grounded discussion of the thermal reactor dynamics. Typical cases of interest are discussed

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

Advanced commercial tokamak study

International Nuclear Information System (INIS)

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

1985-12-01

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

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.

Tokamak Engineering Technology Facility scoping study

International Nuclear Information System (INIS)

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

1976-03-01

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

Dielectronic satellite spectra of hydrogenlike iron from TFTR [Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Decaux, V.; Bitter, M.; Hsuan, H.; von Goeler, S.; Hill, K.W.; Hulse, R.A.; Taylor, G.; Park, H.; Bhalla, C.P.

1990-08-01

Spectra of hydrogenlike iron, Fe26, have been observed from Tokamak Fusion Test Reactor (TFTR) plasmas with a high-resolution crystal spectrometer. The experimental arrangement permits simultaneous observation of the Fe26 Ly-α 1 and Ly-α 2 lines and the associated dielectronic satellites, which are due to transitions 1snl-2pnl' with n ≥ 2, as well as the heliumlike 1s 2 ( 1 S 0 )-1s4p( 1 P 1 )and both hydrogenlike Ly-β 1 and Ly-β 2 lines from chromium. Relative wavelengths and line intensities can be determined very accurately. The spectral data are in very good agreement with theoretical calculations. The observed spectra have also been used to estimate the total dielectronic recombination rate coefficient of Fe26. 30 refs., 4 figs., 3 tabs

Feedback stabilization of the resistive shell mode in a tokamak fusion reactor

International Nuclear Information System (INIS)

Fitzpatrick, R.

1997-01-01

Stabilization of the 'resistive shell mode' is vital to the success of the 'advanced tokamak' concept. The most promising reactor relevant approach is to apply external feedback using, for instance, the previously proposed 'fake rotating shell' scheme [R. Fitzpatrick and T. H. Jensen, Phys. Plasmas 3, 2641 (1996)]. This scheme, like other simple feedback schemes, only works if the feedback controlled conductors are located inside the 'critical radius' at which a perfectly conducting shell is just able to stabilize the ideal external kink mode. In general, this is not possible in a reactor, since engineering constraints demand that any feedback controlled conductors be placed outside the neutron shielding blanket (i.e., relatively far from the edge of the plasma). It is demonstrated that the fake rotating shell feedback scheme can be modified so that it works even when the feedback controlled conductors are located well beyond the critical radius. The gain, bandwidth, current, and total power requirements of such a feedback system for a reactor sized plasma are estimated to be less than 100, a few Hz, a fews tens of kA, and a few MW, respectively. These requirements could easily be met using existing technology. It is concluded that feedback stabilization of the resistive shell mode is possible in a tokamak fusion reactor. copyright 1997 American Institute of Physics

Power supply requirements for a tokamak fusion reactor

International Nuclear Information System (INIS)

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

1979-01-01

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

DEALS: a maintainable superconducting magnet system for tokamak fusion reactors

International Nuclear Information System (INIS)

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

1979-01-01

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

Tight aspect ratio tokamak power reactor with superconducting TF coils

International Nuclear Information System (INIS)

Nishio, S.; Tobita, K.; Konishi, S.; Ando, T.; Hiroki, S.; Kuroda, T.; Yamauchi, M.; Azumi, M.; Nagata, M.

2003-01-01

Tight aspect ratio tokamak power reactor with super-conducting toroidal field (TF) coils has been proposed. A center solenoid coil system and an inboard blanket were discarded. The key point was how to find the engineering design solution of the TF coil system with the high field and high current density. The coil system with the center post radius of less than 1 m can generate the maximum field of ∼ 20 T. This coil system causes a compact reactor concept, where the plasma major and minor radii of 3.75 m and 1.9 m, respectively and the fusion power of 1.8 GW. (author)

Dynamic stabilization of D—T burn in Tokamak reactors

Institute of Scientific and Technical Information of China (English)

ShiBing－Ren; LongYong－Xing

1997-01-01

A simple,engineeringly feasible dynamic method is supposed to control the deuterium-tritium burn process in Tokamak reactors operated in an advanced scenario.The thermal transport of the D-T plasma is described by an anomalous thermal conduction which is a radially increasing function and the central conduction value is proportional to the central temperature of the plasma.The dynamic external heating power is selected to be inversely proportional to certain power function of this temperature,As a result,the D-T burn can undergo in controllable way in different temperature regimes with different power output.Anomalous alpha particle transport effect is taken into account.It can affect the resultant plasma equilibrium ,the reactor efficency,the operation mode and so on.

Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

Science.gov (United States)

Shimomura, Y.

1994-05-01

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

International Nuclear Information System (INIS)

Sager, G.T.; Wong, C.P.C.; Kapich, D.D.; McDonald, C.F.; Schleicher, R.W.

1993-11-01

The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankie and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

Computer modelling of radioactive source terms at a tokamak reactor

International Nuclear Information System (INIS)

Meide, A.

1984-12-01

The Monte Carlo code MCNP has been used to create a simple three-dimensional mathematical model representing 1/12 of a tokamak fusion reactor for studies of the exposure rate level from neutrons as well as gamma rays from the activated materials, and for later estimates of the consequences to the environment, public, and operating personnel. The model is based on the recommendations from the NET/INTOR workshops. (author)

Orbit effects on impurity transport in a rotating tokamak plasma

International Nuclear Information System (INIS)

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

1988-05-01

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

Optimization of OH coil recharging scenario of quasi-steady operation in tokamak fusion reactor by lower hybrid wave current drive

International Nuclear Information System (INIS)

Sugihara, M.; Fujisawa, N.; Nishio, S.; Iida, H.

1984-01-01

Using simple physical model equations optimum plasma and rf parameters for an OH coil recharging scenario of quasi-steady operation in tokamak fusion reactors by lower hybrid wave current drive are studied. In this operation scenario, the minimization of the recharge time of OH coils or stored energy for it will be essential and can be realized by driving sufficient current without increasing the plasma temperature too much. Low density and broad spectrum are shown to be favorable for the minimization. In the case of FER (Fusion Experimental Reactor under design study in JAERI) baseline parameters, the minimum recharge time is 3-5 s/V s. (orig.)

Remote servicing considerations for near term tokamak power reactors (TNS). Final summary

International Nuclear Information System (INIS)

Spampinato, P.T.

1977-01-01

Next generation Tokamaks require special consideration for remote servicing. Three major problems are highlighted: (1) movement of heavy components, (2) remote connection/disconnection of joints, and (3) remote cutting, welding, and leak detection. The first problem is assumed to be handled with existing expertise and is not considered. The remaining problems are thought to be minimized by considering two engineering departures from conventional tokamak design; locating the field shaping coils outside of the toroidal coils and enclosing the total device within an evacuated reactor cell. Five topics under this vacuum building concept are discussed: incremental cost, vacuum pumping, tritium containment, activation topology, and first year operations

Start-up of spherical tokamak without a center solenoid

International Nuclear Information System (INIS)

Maekawa, Takashi; Nagata, Masayoshi

2012-01-01

For low-aspect tokamak reactors, spherical tokamak reactors, ST-type FESF/CTFs, it is essential to remove or minimize a central solenoid (CS). Even with the minimized CS, non-inductive start up of the plasma current is required. Rapid increase in the spontaneous plasma current at the final stage of current start-up drives ignition. At the initial stage, formation of plasma and magnetic surfaces are required. As non-inductive plasma start-up scenarios, ECH/ECCD, LHCD, HHFW, DC HELICITY injection, plasma merging and NBI have been studied. In the present article, the present status and future prospect of experimental and theoretical works on these subjects. (author)

Safety analyses of the ARIES tokamak reactor designs

International Nuclear Information System (INIS)

Herring, J.S.; McCarthy, K.A.; Dolan, T.J.

1994-01-01

The ARIES design has sought to maximize environmental and safety advantages of fusion through careful selection of materials and design. The ARIES-I tokamak reactor design consists of an SiC composite structure for the first wall and blanket, cooled by 10MPa helium. The breeder is Li 2 ZrO 3 . The divertor consists of SiC composite tubes coated with 2mm tungsten. Loss-of-cooling accident (LOCA) calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. The ARIES-II design includes liquid lithium and vanadium, both of which have low activation, multiple barriers between the lithium and air and an inert cover gas to prevent lithium-air reactions. The ARIES-II reactor is passively safe with a total 1km early dose of about 88rem (0.88Sv). ARIES-III was an extensive examination of the viability of a D- 3 He fueled tokamak power reactor. Because neutrons are produced only through side reactions (D+D→ 3 He+n, and D+D→T+p followed by D+T→ 4 He+n), the reactor has a reduced activation of the first wall and shield, low afterheat and class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. We modeled a LOCA in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, below 600 C, release fractions are small. We analyzed the disposition of the 20g per day of tritium that is produced by D-D reactions and removed by vacuum pumps. The ARIES-IV coolant is helium and the breeder is lithium oxide. The structure is silicon carbide. Since the neutron multiplier, beryllium metal, is combustible, releasing about 60MJkg -1 , beryllium is the chief source of chemical energy. Less than 10% of the 24 Na inventory is likely to diffuse out of the SiC during a fire in which the beryllium is consumed. Therefore, the offsite dose would be less than 200rem. ((orig.))

Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie; Jian Hongbing

1992-01-01

The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

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

Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Design of inductively driven long pulse tokamak reactors: IDLT

International Nuclear Information System (INIS)

Ogawa, Yuichi

1998-01-01

Based on scientific data based adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R and D program, the scientific feasibility of inductively-driven tokamak fusion reactors is studied. A low wall-loading DEMO fusion reactor is designed, which utilizes an austenitic stainless steel in conjunction with significant data bases and operating experiences, since we have given high priority to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the DEMO reactor with the relatively large volume (i.e., major radius of 10 m) is employed, plasma ignition is achievable with a low fusion power of 0.8 GW, and an operation period of 4 - 5 hours is available only with inductive current drive. Disadvantages of pulsed operation in commercial fusion reactors include fatigue in structural materials and the necessity of an energy storage system to compensate the electric power during the dwell time. To overcome these disadvantages, a pulse length is prolonged up to about 10 hours, resulting in the remarkable reduction of the total cycle number to 10 4 during the life of the fusion plant. (author)

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)

General description of preliminary design of an experimental fusion reactor and the future problems

International Nuclear Information System (INIS)

Sako, Kiyoshi

1976-01-01

Recently, the studies on plasma physics has progressed rapidly, and promising experimental data emerged successively. Especially expectation mounts high that Tokamak will develop into power reactors. In Japan, the construction of large plasma devices such as JT-60 of JAERI is going to start, and after several years, the studies on plasma physics will come to the end of first stage, then the main research and development will be directed to power reactors. The studies on the design of practical fusion reactors have been in progress since 1973 in JAERI, and the preliminary design is being carried out. The purposes of the preliminary design are the clarification of the concept of the experimental reactor and the requirements for the studies on core plasma, the examination of the problems for developing main components and systems of the reactor, and the development of design technology. The experimental reactor is the quasi-steady reactor of 100 MW fusion reaction output, and the conditions set for the design and the basis of their setting are explained. The outline of the design, namely core plasma, blankets, superconductive magnets and the shielding with them, vacuum wall, neutral particle injection heating device, core fuel supply and exhaust system, and others, is described. In case of scale-up the reactor structural material which can withstand neutron damage must be developed. (Kako, I.)

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

An advanced conceptual Tokamak fusion power reactor utilizing closed cycle helium gas turbines

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

UWMAK-III is a conceptual Tokamak reactor designed to study the potential and the problems associated with an advanced version of Tokamaks as power reactors. Design choices have been made which represent reasonable extrapolations of present technology. The major features are the noncircular plasma cross section, the use of TZM, a molybdenum based alloy, as the primary structural material, and the incorporation of a closed-cycle helium gas turbine power conversion system. A conceptual design of the turbomachinery is given together with a preliminary heat exchanger analysis that results in relatively compact designs for the generator, precooler, and intercooler. This paper contains a general description of the UWMAK-III system and a discussion of those aspects of the reactor, such as the burn cycle, the blanket design and the heat transfer analysis, which are required to form the basis for discussing the power conversion system. The authors concentrate on the power conversion system and include a parametric performance analysis, an interface and trade-off study and a description of the reference conceptual design of the closed-cycle helium gas turbine power conversion system. (Auth.)

Conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR), September 1978

International Nuclear Information System (INIS)

Kelley, J.L.

1978-12-01

The flexibility of the fusion hybrid reactor to function as a fuel production facility, power plant, waste disposal burner or combinations of all of these, as well as the reactor's ability to use proliferation resistant fuel cycles, has provided the incentive to assess the feasibility of a near-term demonstration plant. The goals for a Demonstration Tokamak Hybrid Reactor (DTHR) were established and an initial conceptual design was selected. Reactor performance and economics were evaluated and key developmental issues were assessed. The study has shown that a DTHR is feasible in the late 1980's, a significant quantity of fissile fuel could be produced from fertile thorium using present day fission reactor blanket technology, and a large number of commercially prototypical components and systems could be developed and operationally verified. The DTHR concept would not only serve as proof-of-principle for hybrid technology, but could be operated in the ignited mode and provide major advancements for pure fusion technology

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

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

International Nuclear Information System (INIS)

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

1975-01-01

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

Engineering solutions for components facing the plasma in experimental power reactors

International Nuclear Information System (INIS)

Casini, G.; Farfaletti-Casali, F.

1985-01-01

A review of the engineering problems related to the structures in front of the plasma of experimental Tokamak-type reactors is made. Attention is focused on the so-named ''first wall'', i.e. the wall side of the blanket segments facing the plasma, and on the collector plates of the impurity control system, in particular for the case of the single-null poloidal divertor. Even if the uncertainties related to the plasma-wall interaction are stil relevant, some engineering solutions which look manageable are identified and described. (orig.)

Cryogenic analysis of forced-cooled, superconducting TF magnets for compact tokamak reactors

International Nuclear Information System (INIS)

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

1988-01-01

Current designs for compact tokamak reactors require the toroidal- field (TF) superconducting magnets to produce fields from 10 to 15 T at the winding pack, using high-current densities to high nuclear heat loads (greater than 1 kW/coil in some instances), which are significantly greater than the conduction and radiation heat loads for which cryogenic systems are usually designed. A cryogenic system for the TF winding pack for two such tokamak designs has been verified by performing a detailed, steady-state heat-removal analysis. Helium properties along the forced-cooled conductor flow path for a range of nuclear heat loads have been calculated. The results and implications of this analysis are presented. 12 refs., 6 figs

Experimental observations related to the thermodynamic properties of tokamak plasmas

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Some stress-related issues in tokamak fusion reactor first walls

International Nuclear Information System (INIS)

Majumdar, S.; Pai, B.; Ryder, R.H.

1987-01-01

Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m 2 , and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems

Ion cyclotron transmission spectroscopy in the Tokamak Fusion Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Greene, G.J.

1993-09-01

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

Ion cyclotron transmission spectroscopy in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Greene, G.J.

1993-09-01

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

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

Comparative study of pulsed and steady-state tokamak reactor burn cycles

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.

1984-05-01

Four distinct operating modes have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue in pulsed poloidal field coils; out-of-plant fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics and engineering which will help achieve these goals for different burn cycles

Conceptual design of Fusion Experimental Reactor (FER)

International Nuclear Information System (INIS)

Tone, T.; Fujisawa, N.

1983-01-01

Conceptual design studies of the Fusion Experimental Reactor (FER) have been performed. The FER has an objective of achieving selfignition and demonstrating engineering feasibility as a next generation tokamak to JT-60. Various concepts of the FER have been considered. The reference design is based on a double-null divertor. Optional design studies with some attractive features based on advanced concepts such as pumped limiter and RF current drive have been carried out. Key design parameters are; fusion power of 440 MW, average neutron wall loading of 1MW/m 2 , major radius of 5.5m, plasma minor radius of 1.1m, plasma elongation of 1.5, plasma current of 5.3MA, toroidal beta of 4%, toroidal field on plasma axis of 5.7T and tritium breeding ratio of above unity

Steady-state resistive toroidal-field coils for tokamak reactors

International Nuclear Information System (INIS)

Kalnavarns, J.; Jassby, D.L.

1979-12-01

If spatially-averaged values of the beta ratio can reach 5 to 10% in tokamaks, as now seems likely, resistive toroidal-field coils may be advantageous for use in reactors intended for fusion-neutron applications. The present investigation has parameterized the design of steady-state water-cooled copper coils of rectangular cross section in order to maximize figures of merit such as the ratio of fusion neutron wall loading to coil power dissipation. Four design variations distinguished by different ohmic-heating coil configurations have been examined. For a wall loading of 0.5 MW/m 2 , minimum TF-coil lifetime costs (including capital and electricity costs) are found to occur with coil masses in the range 2400 to 4400 tons, giving 200 to 250 MW of resistive dissipation, which is comparable with the total power drain of the other reactor subsystems

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

Generation of net electric power with a tokamak reactor under foreseeable physical and engineering conditions

International Nuclear Information System (INIS)

Hiwatari, R.; Asaoka, Y.; Okano, K.; Yoshida, T.; Tomabechi, K.

2004-01-01

This study reveals for the first time the plasma performance required for a tokamak reactor to generate net electric power under foreseeable engineering conditions. It was found that the reference plasma performance of the ITER inductive operation mode with β N = 1.8, HH = 1.0, andf nGW 0.85 had sufficient potential to achieve the electric break-even condition (net electric power P e net = 0MW) under the following engineering conditions: machine major radius 6.5m ≤ R p ≤ 8.5m, the maximum magnetic field on TF coils B tmax = 16 T, thermal efficiency η e 30%, and NBI system efficiency η NBI = 50%. The key parameters used in demonstrating net electric power generation in tokamak reactors are β N and fη GW . ≥ 3.0 is required for P e net ∼ 600MW with fusion power P f ∼ 3000MW. On the other hand, fη GW ≥ 1.0 is inevitable to demonstrate net electric power generation, if high temperatures, such as average temperatures of T ave > 16 keV, cannot be selected for the reactor design. To apply these results to the design of a tokamak reactor for demonstrating net electric power generation, the plasma performance diagrams on the Q vs P f (energy multiplication factor vs fusion power) space for several major radii (i.e. 6.5, 7.5, and 8.5 m) were depicted. From these figures, we see that a design with a major radius R p ∼ 7.5m seems preferable for demonstrating net electric power generation when one aims at early realization of fusion energy. (author)

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

International Nuclear Information System (INIS)

2005-01-01

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

Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

International Nuclear Information System (INIS)

Stacey, W.M.

2001-01-01

The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (P fus ∼ 10-100 MW, β N ∼ 2-3, Q p ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

Large aspect ratio tokamak study

International Nuclear Information System (INIS)

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

1979-01-01

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

Tests of vacuum interrupters for the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Warren, R.; Parsons, M.; Honig, E.; Lindsay, J.

1979-04-01

The Tokamak Fusion Test Reactor (TFTR) project at Princeton University requires the insertion of a resistor in an excited ohmic-heating coil circuit to produce a plasma initiation pulse (PIP). It is expected that the maximum duty for the switching system will be an interruption of 24 kA with an associated recovery voltage of 25 kV. Vacuum interrupters were selected as the most economical means to satisfy these requirements. However, it was felt that some testing of available systems should be performed to determine their reliability under these conditions. Two interrupter systems were tested for over 1000 interruptions each at 24 kA and 25 kV. One system employed special Westinghouse type WL-33552 interrupters in a circuit designed by LASL. This circuit used a commercially available actuator and a minimum size counterpulse bank and saturable reactor. The other used Toshiba type VGB2-D20 interrupters actuated by a Toshiba mechanism in a Toshiba circuit using a larger counterpulse bank and saturable reactor

Neutral-beam current drive in tokamaks

International Nuclear Information System (INIS)

Devoto, R.S.

1986-01-01

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

Neutral-beam current drive in tokamaks

International Nuclear Information System (INIS)

Devoto, R.S.

1987-01-01

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

Experimental fusion power reactor conceptual design study. Final report. Volume II

International Nuclear Information System (INIS)

Baker, C.C.

1976-12-01

This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following sections: (1) reactor components, (2) auxiliary systems, (3) operations, (4) facility design, (5) program considerations, and (6) conclusions and recommendations

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

International Nuclear Information System (INIS)

Bruhn, M.L.

1988-04-01

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

Tokamak reactor for treating fertile material or waste nuclear by-products

Science.gov (United States)

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

Potential minimum cost of electricity of superconducting coil tokamak power reactors

International Nuclear Information System (INIS)

Reid, R.L.; Peng, Y-K. M.

1989-01-01

The potential minimum cost of electricity (COE) for superconducting tokamak power reactors is estimated by increasing the physics (confinement, beta limit, bootstrap current fraction) and technology [neutral beam energy, toroidal field (TF) coil allowable stresses, divertor heat flux, superconducting coil critical field, critical temperature, and quench temperature rise] constraints far beyond those assumed for ITER until the point of diminishing returns is reached. A version of the TETRA systems code, calibrated with the ITER design and modified for power reactors, is used for this analysis, limiting this study to reactors with the same basic device configuration and costing algorithms as ITER. A minimum COE is reduced from >200 to about 80 mill/kWh when the allowable design constraints are raised to 2 times those of ITER. At 4 times the ITER allowables, a minimum COE of about 60 mill/kWh is obtained. The corresponding tokamak has a major radius of approximately 4 m, a plasma current close to 10 MA, an aspect ratio of 4, a confinement H- factor ≤3, a beta limit of approximately 2 times the first stability regime, a divertor heat flux of about 20 MW/m 2 , a Β max ≤ 18 T, and a TF coil average current density about 3 times that of ITER. The design constraints that bound the minimum COE are the allowable stresses in the TF coil, the neutral beam energy, and the 99% bootstrap current (essentially free current drive). 14 refs., 4 figs., 2 tabs

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-05-27

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

Monte Carlo analysis of the effects of penetrations on the performance of a tokamak fusion reactor

International Nuclear Information System (INIS)

Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

1977-01-01

Adjoint Monte Carlo calculations have been carried out to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections and the nuclear heating and radiation damage were obtained using the latest available response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 greater than in the fully shielded coils depending on the location. Substantial increases in the radation damage were also noted

Some technical constraints on possible Tokamak machines from next generation to reactor size

International Nuclear Information System (INIS)

Knobloch, A.

1975-11-01

A simplified consistent scaling of possible Tokamak reactors is set up in the power range of 0.1 - 10 GW. The influence of some important parameters on the scaling is shown and the role of some technical constraints is discussed. The scaling is evaluated for the two cases of a circular and a strongly elongated plasma section. (orig.) [de

Magnet design approach for pulsed tokamak reactors

International Nuclear Information System (INIS)

Kim, S.H.; Evans, K. Jr.; Ehst, D.A.

1983-12-01

A choice of various operating modes of a tokamak reactor will have considerable impact on the fatigue lives and cost of ohmic heating (OH), equilibrium field (EF), and toroidal field (TF) coils. OH AND EF coil requirements and their costs, as well as the effects of the fringing fields of the EF coils on the TF coils, have been studied under cyclic operation in the range of N = 10 2 to 10 6 cycles, spanning the range from a noninductively driven reactor (STARFIRE) to a conventional ohmically driven reactor. For a reference design of TF coils the design of the central OH solenoid has been studied as a function of its maximum field, B/sup OH/. Increasing requirements for structural support lead to only negligible increases in volt-seconds for B/sup OH/ greater than or equal to 10.0 T. Fatigue failure of the OH coil is not a concern for N less than or equal to 10 5 ; for N approx. 10 6 fatigue limits the strain to small values, resulting in small increases in structural requirements and modest decreases in volt-seconds. Should noninductive current drive be achievable we note that this not only eliminates the OH coil, but it also permits EF coil placement in the inboard region, which facilitates the creation of highly shaped plasma cross sections (large triangularity, or bean-shaped equilibria). We have computed the stored energy, coil configuration and fringing fields for a number of EF coil design options

Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

International Nuclear Information System (INIS)

Guthrie, J.A.S.

1980-04-01

In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

Studies of tokamak fusion reactor dynamics. Progress report, June 1, 1975--February 15, 1976

International Nuclear Information System (INIS)

Mills, R.G.; Gralnick, S.L.

1976-01-01

An investigation of the effect of plasma shape and position on the inductive coupling between the plasma and the external poloidal field coils is briefly described. Research on a multi-node time-dependent point kinetics code with which to study the operating dynamics of a tokamak reactor is also mentioned

A Midsize Tokamak As Fast Track To Burning Plasmas

International Nuclear Information System (INIS)

Mazzucato, E.

2010-01-01

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

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

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

Fusion reactors - types - problems

International Nuclear Information System (INIS)

Schmitter, K.H.

1979-07-01

A short account is given of the principles of fusion reactions and of the expected advantages of fusion reactors. Descriptions are presented of various Tokamak experimental devices being developed in a number of countries and of some mirror machines. The technical obstacles to be overcome before a fusion reactor could be self-supporting are discussed. (U.K.)

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

International Nuclear Information System (INIS)

Mantsinen, M.

1999-01-01

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

Preparations for deuterium tritium experiments on the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

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

1994-04-01

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

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

International Nuclear Information System (INIS)

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Jaeger, E.F.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Kwon, S.; Labik, G.; Lam, N.T.; LaMarche, P.H.; Laughlin, M.J.; Lawson, E.; LeBlanc, B.; Leonard, M.; Levine, J.; Levinton, F.M.; Loesser, D.; Long, D.; Machuzak, J.; Mansfield, D.E.; Marchlik, M.; Marmar, E.S.; Marsala, R.; Martin, A.; Martin, G.; Mastrocola, V.; Mazzucato, E.; McCarthy, M.P.; Majeski, R.; Mauel, M.; McCormack, B.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Milora, S.L.; Monticello, D.; Mueller, D.; Murakami, M.; Murphy, J.A.; Nagy, A.; Navratil, G.A.; Nazikian, R.; Newman, R.; Nishitani, T.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D.K.; Park, H.; Park, W.; Paul, S.F.; Pavlov, Y.I.; Pearson, G.; Perkins, F.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C.K.; Pitcher, S.; Popovichev, S.; Qualls, A.L.; Raftopoulos, S.; Ramakrishnan, R.; Ramsey, A.; Rasmussen, D.A.; Redi, M.H.

1994-01-01

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

Electromagnetic effects involving a tokamak reactor first wall and blanket

International Nuclear Information System (INIS)

Turner, L.R.; Evans, K. Jr.; Gelbard, E.; Prater, R.

1980-01-01

Four electromagnetic effects experienced by the first wall and blanket of a tokamak reactor are considered. First, the first wall provides reduction of the growth rate of vertical axisymmetric instability and stabilization of low mode number interval kink modes. Second, if a rapid plasma disruption occurs, a current will be induced on the first wall, tending to maintain the field formerly produced by the plasma. Third, correction of plasma movement can begin on a time scale much faster than the L/R time of the first wall and blanket. Fourth, field changes, especially those from plasma disruption or from rapid discharge of a toroidal field coil, can cause substantial eddy current forces on elements of the first wall and blanket. These effects are considered specifically for the first wall and blanket of the STARFIRE commercial reactor design study

Criteria of the efficiency for radiation protection of tokamak reactor superconducting magnet coils

International Nuclear Information System (INIS)

Zimin, S.A.

1988-01-01

Factors determining serviceability of the main elements (superconductor, stabilizing conductor, insulation) of superconducting magnet coils for tokamak reactors are discussed. It is suggested that the limiting values of total and specific energy release in the material of superconducting coils, increase in electric resistance of the stabilizing conductor, decrease in the superconductor critical current and damage of the superconducting magnet insulation should be used as criteria of the reactor internal radiation protection efficiency. The conclusion is made that neutron fluence in the magnet coil components considered can be used as a generalized criterion of the first approximation for the evaluation of the protection efficiency

Design constraints for rf-driven steady-state tokamak reactors

International Nuclear Information System (INIS)

Ehst, D.A.

1979-02-01

Plasma current density profiles are computed due to electron Landau damping of lower hybrid waves launched into model tokamak density and temperature profiles. The total current and current profile shape are chosen consistent with magnetohydrodynamic equilibrium for a variety of temperature and density distributions and plasma beta values. Surface current equilibria appear attractive and are accessible to waves with n/sub z/ as low as 1.2. By suitably choosing the spectrum location and width it is possible to drive the 9.8 MA current of a 7.0-m reactor with as little as 2.8% of the fusion power recirculated as rf input from the waveguides

Start-up simulations of the PULSAR pulsed tokamak reactor

International Nuclear Information System (INIS)

Werley, K.A.; Bathke, C.G.

1993-01-01

Start-up conditions are examined for a pulsed tokamak reactor that uses only inductively driven plasma current (and bootstrap current). A zero-dimensional (profile-averaged) model containing plasma power and particle balance equations is used to study several aspects of plasma start-up, including: (1) optimization of the start-up pathway; (2) tradeoffs of auxiliary start-up heating power versus start-up time; (3) volt-second consumption; (4) thermal stability of the operating point; (5) estimates of the diverter heat flux and temperature during the start-up transient; (6) the sensitivity of the available operating space to allowed values of the H confinement factor; and (7) partial-power operations

Probabilistic analysis of divertor plate lifetime in tokamak reactors

International Nuclear Information System (INIS)

Golinescu, R.P.; Kazimi, M.S.

1994-01-01

Defining a methodology for a reliability estimate of the International Tokamak Experimental Reactor (ITER) divertor is the objective of the study summarized in this paper. If ITER could be designed such that no transients of any type occurred, the divertor reliability would be controlled by erosion of material during normal operation. The occurrence of several transient events results in important contribution to the expected divertor failure rate. Some transients cause the temperature in the divertor plate (DP) to rise; if these temperatures get too high, the structural elements in the DP will weaken and subsequently suffer structural failure and possibly reach the melting temperature. Using the limited data available leads to the result that there is a high probability that the DP will reliably withstand a peak heat flux of 11 MW/m 2 . However, transient events will lead to a much shorter lifetime than desirable for DP's, mainly due to the expected severe effects of plasma disruptions. If transients occurred, but the shutdown mechanism succeeded to perform without inducing a disruption, divertor reliability could be significantly improved. Improved characterization of the disruption conditions, and enlarged scope of failure modes should be pursued to gain confidence in the present conclusions

A ceramic breeder in a poloidal tube blanket for a tokamak reactor

International Nuclear Information System (INIS)

Amici, A.; Anzidei, L.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.; Zampaglione, V.; Petrizzi, L.

1989-01-01

A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods (γLiAlO 2 ) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to γLiAlO 2 volume ratio is 4/1. The He inlet and outlet branches are cooling Be and γLiAlO 2 , respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load=2 MW/m 2 ), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature=570 0 C; inlet He temperature=250 0 ; total extracted power=2700 MW; He pumping power percentage=2%; minimum/maximum γLiAlO 2 temperature=400/900 0 C; maximum structural temperature=475 0 C; and maximum Be temperature=525 0 C. (orig.)

Ion cyclotron emission in tokamak plasmas; Emission cyclotronique ionique dans les plasmas de tokamak

Energy Technology Data Exchange (ETDEWEB)

Fraboulet, D.

1996-09-17

Detection of {alpha}(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, {alpha} 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. {alpha} 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 {alpha} density in a reactor. (author). 166 refs.

Direct energy conversion and neutral beam injection for catalyzed D and D-3He tokamak reactors

International Nuclear Information System (INIS)

Blum, A.S.; Moir, R.W.

1977-01-01

The calculated performance of single stage and Venetian blind direct energy converters for Catalyzed D and D- 3 He Tokamak reactors are discussed. Preliminary results on He pumping are outlined. The efficiency of D and T neutral beam injection is reviewed

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

Directory of Open Access Journals (Sweden)

Yutthapong Pinanroj

2014-04-01

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

Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

International Nuclear Information System (INIS)

Chen Yiping; Liu Songlin

2010-01-01

Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

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

Evaluation of a nonevaporable getter pump for tritium handling in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Singleton, M.F.; Griffith, C.M.

1978-01-01

Lawrence Livermore Laboratory has tested and evaluated a commercially available getter pump for use with tritium in the Tokamak Fusion Test Reactor (TFTR). The pump contains Zr(84%)--Al in cartridge form with a concentric heating unit. It performed well in all tests, except for frequent heater failures

Tokamak ARC damage

International Nuclear Information System (INIS)

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

1985-01-01

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

Tokamak ARC damage

Energy Technology Data Exchange (ETDEWEB)

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

1985-01-01

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

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

International Nuclear Information System (INIS)

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

1992-03-01

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

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

International Nuclear Information System (INIS)

Anon.

1994-01-01

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

Recent Activities on the Experimental Research Programme Using Small Tokamaks

International Nuclear Information System (INIS)

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

2006-01-01

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

Status report on the conceptual design of a commercial tokamak hybrid reactor (CTHR)

International Nuclear Information System (INIS)

1979-09-01

A preliminary conceptual design is presented 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. The study has been made in sufficient depth to indicate no insurmountable technical problems exist and has provided a basis for valid 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

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)

Culham Conceptual Tokamak Mark II. Design study of the layout of a twin-reactor fusion power station

International Nuclear Information System (INIS)

Guthrie, J.A.S.; Harding, N.H.

1981-07-01

This report describes the building layout and outline design for the nuclear complex of a fusion reactor power station incorporating two Culham Conceptual Tokamak Reactors Mk.II. The design incorporates equipment for steam generation, process services for the fusion reactors and all facilities for routine and non-routine servicing of the nuclear complex. The design includes provision of temporary facilities for on site construction of the major reactor components and shows that these facilities may be used for disassembly of the reactors either for major repair and/or decommissioning. Preliminary estimates are included, which indicate the cost benefits to be obtained from incorporating two reactors in one nuclear complex and from increased wall loading. (author)

Analysis of Confinement Strategies for a Tokamak Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Girard, Christian; Gaillard, Jean-Philippe; Marbach, Gabriel; Cambi, Gilio; Cook, Ian; Johansson, Lise-Lotte; Meyder, Rainer; Mustoe, Julian; Pinna, Tonio

2001-01-15

The Safety and Environmental Assessment of Fusion Power (SEAFP) was performed in the framework of the European fusion program, whose results have already been published. The European Commission decided to continue this program for some identified issues that required development. One of these issues was the analysis and specification of the containment concepts that minimize accidental releases to the environment.To perform such an assessment, a methodology was followed to identify the most challenging accidental sequences in terms of containment integrity.The results of the accident selection and analysis that were performed during the extension of the SEAFP-2 program are given. Preliminary recommendations for the definition of a confinement strategy for tokamak fusion reactors are established.

Analysis of Confinement Strategies for a Tokamak Fusion Reactor

International Nuclear Information System (INIS)

Girard, Christian; Gaillard, Jean-Philippe; Marbach, Gabriel; Cambi, Gilio; Cook, Ian; Johansson, Lise-Lotte; Meyder, Rainer; Mustoe, Julian; Pinna, Tonio

2001-01-01

The Safety and Environmental Assessment of Fusion Power (SEAFP) was performed in the framework of the European fusion program, whose results have already been published. The European Commission decided to continue this program for some identified issues that required development. One of these issues was the analysis and specification of the containment concepts that minimize accidental releases to the environment.To perform such an assessment, a methodology was followed to identify the most challenging accidental sequences in terms of containment integrity.The results of the accident selection and analysis that were performed during the extension of the SEAFP-2 program are given. Preliminary recommendations for the definition of a confinement strategy for tokamak fusion reactors are established

Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

International Nuclear Information System (INIS)

Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

2015-01-01

Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor

Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

International Nuclear Information System (INIS)

Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

1979-01-01

In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

Magnetic confinement experiment. I: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1995-08-01

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

New tritium monitor for the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Jalbert, R.A.

1985-01-01

At DT-fueled fusion reactors, there will be a need for tritium monitors that can simultaneously measure in real time the concentrations of HTO, HT and the activated air produced by fusion neutrons. Such a monitor has been developed, tested and delivered to the Princeton Plasma Physics Laboratory for use at the Tokamak Fusion Test Reactor (TFTR). It uses semipermeable membranes to achieve the removal of HTO from the sampled air for monitoring and a catalyst to convert the HT to HTO, also for removal and monitoring. The remaining air, devoid of tritium, is routed to a third detector for monitoring the activated air. The sensitivities are those that would be expected from tritium instruments employing conventional flow-through ionization chambers: 1 to 3 μCi/m 3 . Its discriminating ability is approximately 10 -3 for any of the three components (HTO, HT and activated air) in any of the other two channels. For instance, the concentration of HT in the HTO channel is 10 -3 times its original concentration in the sampled air. This will meet the needs of TFTR

Tritium Experience in Large Tokamaks: Application to ITER

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D; Gentile, C.; Federici, G.; Haanges, R.

1998-05-01

Recent experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports on an IEA workshop that brought together scientists and engineers to share experience and expertise on all fusion-related tritium issues. Extensive discussion periods were devoted to exploring outstanding issues and identifying potential R ampersand D avenues to address them. This paper summarizes the presentations, discussions, and recommendations

Alternate ohmic heating coil arrangements for compact tokamak

International Nuclear Information System (INIS)

Dawson, J.W.; Moretti, A.; Stevens, H.C.; Thompson, K.

1978-01-01

The results for a number of ohmic heating (OH) coil arrangements which will allow the reduction of the major radius of Experimental Power Reactor (EPR) tokamaks will be given. In each case the results are compared, at least indirectly, to the reference case, which has the OH solenoid inside the central core of the reactor. The goal for the alternate geometries studied was to stay within the requirements imposed by the EPR conditions on the plasma and to produce as much or more OH V-s as the reference case

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-06-01

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

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.

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.

Tokamak power systems studies, FY 1985

International Nuclear Information System (INIS)

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

1985-12-01

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

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.

Development of integrated insulation joint for cooling pipe in tokamak reactor

International Nuclear Information System (INIS)

Nishio, Satoshi; Abe, Tetsuya; Kawamura, Masashi; Yamazaki, Seiichiro.

1994-08-01

In a tokamak fusion reactor, an electrically insulated part is needed for an in-vessel piping system in order to break an electric circuit loop. When a closed loop is formed in the piping system, large induced electromagnetic forces during a plasma disruption (rapid plasma current quench) could give damages on the piping system. Ceramic brazing joint is a conventional method for the electric circuit break, but an application to the fusion reactor is not feasible due to its brittleness. Here, a stainless steel/ceramics/stainless steel functionally gradient material (FGM) has been proposed and developed as an integrated insulation joint of the piping system. Both sides of the joint can be welded to the main pipes, and expected to be reliable even in the fusion reactor environment. When the FGM joint is manufactured by way of a sintering process, a residual thermal stress is the key issue. Through detailed computations of the residual thermal stress and several trial productions, tubular elements of FGM joints have been successfully manufactured. (author)

RF-driven tokamak reactor with sub-ignited, thermally stable operation

International Nuclear Information System (INIS)

Harten, L.P.; Bers, A.; Fuchs, V.; Shoucri, M.M.

1981-02-01

A Radio-Frequency Driven Tokamak Reactor (RFDTR) can use RF-power, programmed by a delayed temperature measurement, to thermally stabilize a power equilibrium below ignition, and to drive a steady state current. We propose the parameters for such a device generating approx. = 1600 MW thermal power and operating with Q approx. = 40 (= power out/power in). A one temperature zero-dimensional model allows simple analytical formulation of the problem. The relevance of injected impurities for locating the equilibrium is discussed. We present the results of a one-dimensional (radial) code which includes the deposition of the supplementary power, and compare with our zero-dimensional model

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.

Diagnostic of the spatial and velocity distribution of alpha particles in tokamak fusion reactor using beat-wave generated lower hybrid wave. Progress report, 1994-1995

International Nuclear Information System (INIS)

Hwang, D.Q.; Horton, R.D.; Evans, R.

1995-01-01

The alpha particle population from fusion reactions in a DT tokamak reactor can have dramatic effects on the pressure profiles, energetic particle confinement, and the overall stability of the plasma; thus leading to important design consideration of a fusion reactor based on the tokamak concept. In order to fully understand the effects of the alpha population, a non-invasive diagnostic technique suitable for use in a reacting plasma environment needs to be developed to map out both the spatial and velocity distribution of the alphas. The proposed experimental goals for the eventual demonstration of LH wave interaction with a fast ion population is given in the reduced 3 year plan in table 1. At present time the authors are approaching the 8th month in their first year of this project. Up to now, their main effort has been concentrated in the operation of the two beat wave sources in burst mode. The second priority in the experimental project is the probe diagnostics and computer aided data acquisition system. The progress made so far is given, and they are ready to perform the beat-wave generated lower hybrid wave experiment. Some theoretical calculation had been reported at APS meetings. More refined theoretical models are being constructed in collaboration with Drs. J. Rogers and E. Valeo at PPPL

Plan for decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

Limiter/vacuum system for plasma impurity control and exhaust in tokamaks

International Nuclear Information System (INIS)

Abdou, M.; Brooks, J.; Mattas, R.

1980-01-01

A detailed design of a limiter/vacuum system for plasma impurity control and exhaust has been developed for the STARFIRE tokamak power plant. It is shown that the limiter/vacuum concept is a very attractive option for power reactors. It is relatively simple and inexpensive and deserves serious experimental verification

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1990-04-01

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

Tritium pellet injector design for tokamak fusion test reactor

International Nuclear Information System (INIS)

Fisher, P.W.; Baylor, L.R.; Bryan, W.E.

1985-01-01

A tritium pellet injector (TPI) system has been designed for the Tokamak Fusion Test Reactor (TFTR) Q approx. 1 phase of operation. The injector gun utilizes a radial design with eight independent barrels and a common extruder to minimize tritium inventory. The injection line contains guide tubes with intermediate vacuum pumping stations and fast valves to minimize propellant leakage to the torus. The vacuum system is designed for tritium compatibility. The entire injector system is contained in a glove box for secondary containment protection against tritium release. Failure modes and effects have been analyzed, and structural analysis has been performed for most intense predicted earthquake conditions. Details of the design and operation of this system are presented in this paper

A programmatic framework for the Tokamak Physics Experiment (TPX)

International Nuclear Information System (INIS)

Thomassen, K.I.; Goldston, R.J.; Neilson, G.H.

1993-01-01

Significant advances have been made in the confinement of reactor-grade plasmas, so that the authors are now preparing for experiments at the open-quotes power breakevenclose quotes level in the JET and TFTR experiments. In ITER the authors will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, the authors must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the US domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX the authors can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. The authors can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current US fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program

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.)

Technological and environmental prospects of low aspect ratio tokamak reactor VECTOR

International Nuclear Information System (INIS)

Nishio, S.; Tobita, K.; Isono, T.; Nakamura, Y.; Sato, M.; Sakurai, S.; Yamauchi, M.; Nishitani, T.; Tani, K.; Sengoku, S.; Kudo, Y.; Song, Y.; Tokimatsu, K.; Shinya, K.; Senda, I.; Konishi, S.

2005-01-01

Optimization of VECTOR design parameters has led to a fusion output of 2.5 GW with a small reactor weight of 8,800 tons. Current ramp-up simulation with TSC (Tokamak Simulation Code) demonstrated a stable current ramp-up with externally non-inductive current drive and bootstrap current when a central current is induced externally enough to avoid an extreme reversed shear such as current hole. It was numerically confirmed that such a low-A reactor would have an advantage of α-particle confinement. The divertor is designed to use a flux expansion of scrape-off-layer in the divertor region to maintain the heat flux on the divertor plate to be lower than 10 MW/m 2 without remote radiative cooling. Because of its compactness, CO 2 emission in a life cycle of a VECTOR power plant is estimated to be as low as 3.2 g-CO 2 /kWh, being lower than that of an ITER-sized DEMO reactor (4.9 g-CO 2 /kWh). As to the waste management of VECTOR, on the basis of reactor design and radiological considerations, we suggest reusing a liquid metal breeding material (PbLi) and neutron shield material (TiH 2 ) in successive reactors. Due to this waste management, its disposal waste would be reduced to as low as 3,000-4,000 tons, which is comparable with the radioactive waste of a light water reactor (4,000 tons in metal). (author)

A conceptual design of superconducting spherical tokamak reactor

International Nuclear Information System (INIS)

Nagayama, Yoshio; Shinya, Kichiro; Tanaka, Yasutoshi

2012-01-01

This paper presents a fusion reactor concept named 'JUST (Japanese Universities' Super Tokamak reactor)'. From the plasma confinement system to the power generation system is evaluated in this work. JUST design has features as follows: the superconducting magnet, the steady state operation with high bootstrap current fraction, the easy replacement of neutron damaged first wall, the high heat flux in the divertor, and the low cost (or high β). By winding the OH solenoid over the center stack of toroidal field coil, we have the low aspect ratio and the 80cm thick neutron shield to protect the superconducting center stack. JUST is designed by using the 0-D transport code under the assumption that the energy confinement time is 1.8 times of the IPB98(y,2) scaling. Main parameters are as follows: the major radius of 4.5m, the aspect ratio of 1.8, the elongation ratio of 2.5, the toroidal field of 2.36T, the plasma current of 18MA, the toroidal beta of 22%, the central electron and ion temperature of 15keV and the fusion thermal power of 2.4GW. By using the mercury heat exchanger and the steam turbine, the heat efficiency is 33% and the electric power is 0.74GW. (author)

Design study of electrical power supply system for tokamak fusion power reactor

International Nuclear Information System (INIS)

1977-01-01

Design study of the electrical power supply system for a 2000MWt Tokamak-type fusion reactor has been carried out. The purposes are to reveal and study problems in the system, leading to a plan of the research and development. Performed were study of the electrical power supply system and design of superconducting inductive energy storages and power switches. In study of the system, specification and capability of various power supplies for the fusion power reactor and design of the total system with its components were investigated. For the superconducting inductive energy storages, material choice, design calculation, and structural design were conducted, giving the size, weight and performance. For thyristor switches, circuit design in the parallel / series connection of element valves and cooling design were studied, providing the size and weight. (auth.)

Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

International Nuclear Information System (INIS)

Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

1981-10-01

Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production

Safety in the ARIES-III D-3He tokamak reactor design

International Nuclear Information System (INIS)

Herring, J.S.; Dolan, T.J.

1992-01-01

This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

Maximum attainable power density and wall load in tokamaks underlying reactor relevant constraints

International Nuclear Information System (INIS)

Borrass, K.; Buende, R.

1979-09-01

The characteristic data of tokamaks optimized with respect to their power density or wall load are determined. Reactor relevant constraints are imposed, such as a fixed plant net power output, a fixed blanket thickness and the dependence of the maximum toroidal field on the geometry and conductor material. The impact of finite burn times is considered. Various scaling laws of the toroidal beta with the aspect ratio are discussed. (orig.) 891 GG/orig. 892 RDG [de

Reactor similarity for plasma–material interactions in scaled-down tokamaks as the basis for the Vulcan conceptual design

International Nuclear Information System (INIS)

Whyte, D.G.; Olynyk, G.M.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Hartwig, Z.S.; Mumgaard, R.T.; Podpaly, Y.A.

2012-01-01

Highlights: ► Discussion of similarity scalings for reduced-size tokamaks. ► Proposal of a new set of scaling laws for divertor similarity. ► Discussion of how the new scaling provides fidelity to a reactor. ► The new scaling is used as the basis for the Vulcan conceptual design. - Abstract: Dimensionless parameter scaling techniques are a powerful tool in the study of complex physical systems, especially in tokamak fusion experiments where the cost of full-size devices is high. It is proposed that dimensionless similarity be used to study in a small-scale device the coupled issues of the scrape-off layer (SOL) plasma, plasma–material interactions (PMI), and the plasma-facing material (PFM) response expected in a tokamak fusion reactor. Complete similarity is not possible in a reduced-size device. In addition, “hard” technological limits on the achievable magnetic field and peak heat flux, as well as the necessity to produce non-inductive scenarios, must be taken into account. A practical approach is advocated, in which the most important dimensionless parameters are matched to a reactor in the reduced-size device, while relaxing those parameters which are far from a threshold in behavior. “Hard” technological limits are avoided, so that the reduced-size device is technologically feasible. A criticism on these grounds is offered of the “P/R” model, in which the ratio of power crossing the last closed flux surface (LCFS), P, to the device major radius, R, is held constant. A new set of scaling rules, referred to as the “P/S” scaling (where S is the LCFS area) or the “PMI” scaling, is proposed: (i) non-inductive, steady-state operation; (ii) P is scaled with R 2 so that LCFS areal power flux P/S is constant; (iii) magnetic field B constant; (iv) geometry (elongation, safety factor q * , etc.) constant; (v) volume-averaged core density scaled as n≈n ¯ e ∼R −2/7 ; and (vi) ambient wall material temperature T W,0 constant. It is

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.)

Comparisons of calculated and measured spectral distributions of neutrons from a 14-MeV neutron source inside the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr.; Emmett, M.B.; Drischler, J.D.

1985-12-01

A recent paper presented neutron spectral distributions (energy greater than or equal to0.91 MeV) measured at various locations around the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The neutron source for the series of measurements was a small D-T generator placed at various positions in the TFTR vacuum chamber. In the present paper the results of neutron transport calculations are presented and compared with these experimental data. The calculations were carried out using Monte Carlo methods and a very detailed model of the TFTR and the TFTR test cell. The calculated and experimental fluences per unit energy are compared in absolute units and are found to be in substantial agreement for five different combinations of source and detector positions

First wall and blanket design for the STARFIRE commercial tokamak power reactor

International Nuclear Information System (INIS)

Morgan, G.D.; Trachsel, C.A.; Cramer, B.A.; Bowers, D.A.; Smith, D.L.

1979-01-01

The first wall and blanket design concepts being evaluated for the STARFIRE commercial tokamak reactor study are presented. The two concepts represent different approaches to the mechanical design of a tritium breeding blanket using the reference materials options. Each concept has a separate ferritic steel first wall cooled by heavy water (D 2 O), and a ferritic steel blanket with solid lithium oxide breeder cooled by helium. A separate helium purge system is used in both concepts to extract tritium. The two concepts are compared and relative advantages and disadvantages for each are discussed

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

Requirements for US regulatory approval of the International Thermonuclear Experimental Reactor (ITER)

International Nuclear Information System (INIS)

Petti, D.A.; Haire, J.C.

1993-12-01

The International Thermonuclear Experimental Reactor (ITER) is the first fusion machine that will have sufficient decay heat and activation product inventory to pose potential nuclear safety concerns. As a result, nuclear safety and environmental issues will be much more important in the approval process for the design, siting, construction, and operation of ITER in the United States than previous fusion devices, such as the Tokamak Fusion Test Reactor. The purpose of this report is (a) to provide an overview of the regulatory approval process for a Department of Energy (DOE) nuclear facility; (b) to present the dose limits used by DOE to protect workers, the public, and the environment from the risks of exposure to radiation and hazardous materials; (c) to discuss some key nuclear safety-related issues that must be addressed early in the Engineering Design Activities (EDA) to obtain regulatory approval; and (d) to provide general guidelines to the ITER Joint Central Team (JCT) concerning the development of a regulatory framework for the ITER project

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

ASTRA - an automatic system for transport analysis in a tokamak

International Nuclear Information System (INIS)

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

1991-08-01

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

Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor

International Nuclear Information System (INIS)

Efthimion, P.C.; Goeler, S. von; Houlberg, W.A.

2001-01-01

Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

Observation of neoclassical transport in reverse shear plasmas on the tokamak fusion test reactor

International Nuclear Information System (INIS)

Efthimion, P.C.; Von Goeler, S.; Houlberg, W.A.

1999-01-01

Perturbative experiments on the Tokamak Fusion Test Reactor (TFTR) have investigated the transport of multiple ion species in reverse shear plasmas. The profile evolution of trace tritium and helium, and intrinsic carbon indicate the formation of core particle transport barriers in ERS plasmas. There is an order of magnitude reduction in the particle diffusivity inside the reverse shear region. The diffusivities for these species in ERS plasmas agree with neoclassical theory. (author)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

Challender, R.S.; Reynolds, P.

1976-01-01

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

Experimental operation of the KT-5C tokamak in USTC

International Nuclear Information System (INIS)

Wen Yizhi; Wan Shude; Zhai Kan; Liu Wandong

1995-01-01

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

Fusion reactor development: A review

International Nuclear Information System (INIS)

Anon.

1983-01-01

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

Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

International Nuclear Information System (INIS)

Seki, Y.; Mori, S.

1984-01-01

The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

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

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

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

Blanket design study for a Commercial Tokamak Hybrid Reactor (CTHR)

International Nuclear Information System (INIS)

Chapin, D.L.; Green, L.; Lee, A.Y.; Culbert, M.E.; Kelly, J.L.

1979-09-01

The results are presented of a study on two blanket design concepts for application in a Commercial Tokamak Hybrid Reactor (CTHR). Both blankets operate on the U-Pu cycle and are designed to achieve tritium self-sufficiency while maximizing the fissile fuel production within thermal and mechanical design constraints. The two blanket concepts that were evaluated were: (1) a UC fueled, stainless steel clad and structure, helium cooled blanket; and (2) a UO 2 fueled, zircaloy clad, stainless steel structure, boiling water cooled blanket. Two different tritium breeding media, Li 2 O and LiH, were evaluated for use in both blanket concepts. The use of lead as a neutron multiplier or reflector and graphite as a reflector was also considered for both blankets

Conceptual design of PF coil system and operation scenario on inductively-operated day-long pulsed tokamak reactor

International Nuclear Information System (INIS)

Wang, J.F.; Yamamoto, T.; Ogawa, Y.

1994-01-01

It is said that disadvantages of pulsed operation in tokamak fusion reactor are fatigue problem of structural materials and an introduction of energy storage System to compensate the power during the dwell time. To overcome theses disadvantages the authors have designed an inductively-operated ultralong pulsed tokamak called (IDLT) reactor where plasma with a major radius of 10 m are employed so as to provide a magnetic flux necessary to sustain a plasma current inductively during 10 hours or more. This makes it possible to reduce the total cycle number to be around 10 4 during the life of the fusion plant. In pulsed operation reactors the shorter dwell time with a quick start-up and shut down of plasma is very convenient to realize a high availability of the power plant, but it will induce more severe conditions for the hardware design. The authors assumed the dwell time of 5∼10 minutes and analyzed the feasibility of plasma operation scenario for IDLT reactor, especially paying much attention to PF coil system. The stored energy of PF coil system becomes ∼100 GJ, which is comparable with that of toroidal field coil system. When the plasma current of 14 MA is ramped up with a time of 100 seconds, it is found that the maximum capacity of 1 GW is necessary for PF coil power supply. Engineering issues related with AC/hysterisis loss should be carefully examined

A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

Directory of Open Access Journals (Sweden)

Mitsuru Kikuchi

2010-11-01

Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

Thermally excited proton spin-flip laser emission in tokamaks

International Nuclear Information System (INIS)

Arunasalam, V.; Greene, G.J.

1993-07-01

Based on statistical thermodynamic fluctuation arguments, it is shown here for the first time that thermally excited spin-flip laser emission from the fusion product protons can occur in large tokamak devices that are entering the reactor regime of operation. Existing experimental data from TFTR supports this conjecture, in the sense that these measurements are in complete agreement with the predictions of the quasilinear theory of the spin-flip laser

Some problems of a tokamak with fast-neutral injection

International Nuclear Information System (INIS)

Pistunovich, V.I.

1976-01-01

An attempt has been made to formulate the fundamental physical problems which should be solved for designing a reactor on the basis of a tokamak with injection. The problems enumerated should be considered particularly for the suggested system, though qualitative solutions of some of them have been already known. They are the following. 1) Ionization of the atomic beam and distribution of the generated ion density. 2) Capture of fast ions by the tokamak magnetic field. 3) Anisotropy of the velocity distribution function. 4) The effect of the neutral gas density on the shape of the one-dimensional distribution function. 5)Stability of the ion distribution function on deceleration of the ion beam in plasma. The paper permits evaluating the advantages and shortcomings of the reactor on the basis of a tokamak with injection from viewpoints of the requirements on confinement and heating of plasma, and of the effect of impurities and neutral working gas on the output parameters of the reactor. The author believes that the regime of a two-component tokamak with small values of quality Q (<=) 3 shows greatest advantages over the ignition reactor, though in this case a reactor may be economically profitable (Q approximately 10) only with an increase of the total quality of the reactor by additional energy release in a blanket on using fission reactions

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

Design of an advanced bundle divertor for the Demonstration Tokamak Hybrid Reactor

International Nuclear Information System (INIS)

Yang, T.F.; Lee, A.Y.; Ruck, G.W.; Prevenslik, T.V.; Smeltzer, G.

1979-01-01

The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm 2 ) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm 2 ) and for ISX-B 2 (11 kA/cm 2 ). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure

Burn Control Mechanisms in Tokamaks

Science.gov (United States)

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

2015-11-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-09-01

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

Remote maintenance considerations for swimming pool tokamak reactor

International Nuclear Information System (INIS)

Niikura, S.; Yamada, M.; Kasai, M.

1983-01-01

Swimming Pool Tokamak Reactor (SPTR) is one of the candidate devices which are expected to demonstrate physical and engineering feasibility for fusion power reactors. In SPTR, water shield is adopted instead of solid shield structures. Among the advantages of SPTR are, from viewpoint of remote maintenance, small handling weight and high space availability between TF coils and a vacuum vessel. On the other hand, high dose rate during reactor repair and adverse effects on remote maintenance equipment by the shielding water might be the disadvantage of SPTR, where it is assumed that the shielding water is drained during reactor repair. Since the design of SPTR is still at the preliminary stage, for remote maintenance, much effort has been directed to clarification of design conditions such as environment and handling weight. As for the remote maintenance system concepts, studies have been focussed on those for a vacuum vessel and its internal structure (blanket, divertor and protection walls) expected to be repaired more frequently. The vacuum vessel assembly is divided into 21 sectors and number of TF coils is 14. A pair of TF coils are connected with each other by antitorque beams on the whole side surface. Vacuum vessel cassettes and associated blanket, divertor and protection walls are replaced through seven windows between TF coils pairs. Therefore each vacuum vessel cassette is required moving mechanisms in toroidal and radial directions. Options for slide mechanisms are wheels, balls, rollers and water bearings. Options for driving the cassette are self-driving by hydraulic motors and external driving by rack-pinion, wires or specific vehicles. As a result of studies, the moving mechanism with wheels and hydraulic motors has been selected for the reference design, and the system with water bearings and rack-pinion as an alternative. Furthermore typical concepts have been obtained for remote maintenance equipment such as wall-mounted manipulators, tools for

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

Low temperature plasma near a tokamak reactor limiter

International Nuclear Information System (INIS)

Braams, B.J.; Singer, C.E.

1985-01-01

Analytic and two-dimensional computational solutions for the plasma parameters near a toroidally symmetric limiter are illustrated for the projected parameters of a Tokamak Fusion Core Experiment (TFCX). The temperature near the limiter plate is below 20 eV, except when the density 10 cm inside the limiter contact is 8 x 10 13 cm -3 or less and the thermal diffusivity in the edge region is 2 x 10 4 cm 2 /s or less. Extrapolation of recent experimental data suggests that neither of these conditions is likely to be met near ignition in TFCX, so a low plasma temperature near the limiter should be considered a likely possibility

Conceptual designs of tokamak reactor and R D

International Nuclear Information System (INIS)

Fukai, Yuzo; Yamato, Harumi; Sawada, Yoshio

1983-01-01

The conceptual design of both FER (Fusion Experimental Reactor) and R-project is now under way as the new step of JT-60. From the engineering viewpoint, these reactors, requiring D-T operation, have the challenge, such as the handling of tritium and components irradiated by neutron bombardment. Toshiba's design team is participating to these projects in order to realize the reactor and plant concept coping with the above objectives. This paper represents the conceptual design contributions of the FER and R-project as well as R D technology which are now under development, such as tritium handling app aratus, reactor materials, etc. (author)

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

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

International Nuclear Information System (INIS)

Li Jiangang; Yang Yu

2003-01-01

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

Advances in fusion reactor design

International Nuclear Information System (INIS)

Baker, C.C.

1987-01-01

The author addresses the tokamak as a power reactor. Contrary to popular opinion, there are still a few people that think a tokamak might make a good fusion power reactor. In thinking about advances in fusion reactor design, in the U.S., at least, that generally means advances relevant to the Starfire design. He reviews some of the features of Starfire. Starfire is the last major study done of the tokamak as a reactor in this country. It is now over eight years old in the sense that eight years ago was really the time in which major decisions were made as to its features. Starfire was a tokamak with a major radius of seven meters, about twice the linear dimensions of a machine like TIBER

Plasma control issues for an advanced steady state tokamak reactor

International Nuclear Information System (INIS)

Moreau, D.

2001-01-01

This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

Plasma control issues for an advanced steady state tokamak reactor

International Nuclear Information System (INIS)

Moreau, D.; Voitsekhovitch, I.

1999-01-01

This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

Overview on the progress of tokamak experimental research in China

International Nuclear Information System (INIS)

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

2001-01-01

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

Second regime tokamak operation at large aspect ratio

International Nuclear Information System (INIS)

Navratil, G.A.

1989-01-01

This paper reviews the need for high beta in economic tokamak reactors and summarizes recent results on the scaling of the second regime beta limit for high-n ballooning modes using optimized pressure profiles as well as results on low-n mode stability at the first regime beta limit from the Columbia HBT tokamak. While several experiments have studied ballooning limits using high εβ p plasmas, the most important question for the use of the second stability regime for tokamak reactor improvement is how to achieve these high values of εβ p while at the same time increasing the value of beta to several times the Troyon beta limit. An approach to the study of this key question on beta limits using modest sized, large aspect ratio tokamaks is described. (author). 28 refs, 7 figs, 1 tab

High-field, high-density tokamak power reactor

International Nuclear Information System (INIS)

Cohn, D.R.; Cook, D.L.; Hay, R.D.; Kaplan, D.; Kreischer, K.; Lidskii, L.M.; Stephany, W.; Williams, J.E.C.; Jassby, D.L.; Okabayashi, M.

1977-11-01

A conceptual design of a compact (R 0 = 6.0 m) high power density (average P/sub f/ = 7.7 MW/m 3 ) tokamak demonstration power reactor has been developed. High magnetic field (B/sub t/ = 7.4 T) and moderate elongation (b/a = 1.6) permit operation at the high density (n(0) approximately 5 x 10 14 cm -3 ) needed for ignition in a relatively small plasma, with a spatially-averaged toroidal beta of only 4%. A unique design for the Nb 3 Sn toroidal-field magnet system reduces the stress in the high-field trunk region, and allows modularization for simpler disassembly. The modest value of toroidal beta permits a simple, modularized plasma-shaping coil system, located inside the TF coil trunk. Heating of the dense central plasma is attained by the use of ripple-assisted injection of 120-keV D 0 beams. The ripple-coil system also affords dynamic control of the plasma temperature during the burn period. A FLIBE-lithium blanket is designed especially for high-power-density operation in a high-field environment, and gives an overall tritium breeding ratio of 1.05 in the slowly pumped lithium

Thermal and nonthermal electron cyclotron emission by high-temperature tokamak plasmas

International Nuclear Information System (INIS)

Airoldi, A.; Ramponi, G.

1997-01-01

An analysis of the electron cyclotron emission (ECE) spectra emitted by a high-temperature tokamak plasma in the frequency range of the second and third harmonic of the electron cyclotron frequency is made, both in purely Maxwellian and in non-Maxwellian cases (i.e., in the presence of a current-carrying superthermal tail). The work is motivated mainly by the experimental observations made in the supershot plasmas of the Tokamak Fusion Test Reactor (TFTR), where a systematic disagreement is found between the T e measurements by second-harmonic ECE and Thomson scattering. We show that, by properly taking into account the overlap of superthermals-emitted third harmonic with second-harmonic bulk emission, the radiation temperature observed about the central frequency of the second harmonic may be enhanced up to 30%endash 40% compared to the corresponding thermal value. Moreover we show that, for parameters relevant to the International Thermonuclear Experimental Reactor (ITER) with T e (0)>7 keV, the overlap between the second and the downshifted third harmonic seriously affects the central plasma region, so that the X-mode emission at the second harmonic becomes unsuitable for local T e measurements. copyright 1997 American Institute of Physics

Parameter study of high-β tokamak reactors with circular and strongly elongated cross section

International Nuclear Information System (INIS)

Herold, H.

1977-05-01

A simplified reactor model is used to study the influence of critical β-values on economy parameters and dimensions of possible long time pulsed tokamak reactors. Various betas deduced from stability and equilibrium MHD theory are introduced and put into the scaling in context with technological constraints, as maximum B-field, core constraint, maximum wall loading a.o. The plasma physical concepts treated comprise circular and strongly elongated cross section and approximated FCT equilibria. The computational results are presented as plots of possible economy parameter ranges (magnet energy, wall loading, volumina, investment costs per unit power) dependent on β for suitably chosen hierarchies of the constraints. - A burn time reduction by the build ups of α-pressure may be possible for the pressure profile sensitive high-β equilibria (FCT). Burn times in the 1O sec range, resulting from simple estimates, would about cancel the economic advantages of reactors with high-β equilibria compared to a β = 5% standardreactor (UWMAK I). (orig.) [de

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

International Nuclear Information System (INIS)

Perez-Navarro, A.

1977-01-01

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

Locking mechanism for in-vessel components of tokamak reactor

International Nuclear Information System (INIS)

Nishio, S.; Shimizu, K.; Koizumi, K.; Tada, E.

1992-01-01

The locking and unlocking mechanism for in-vessel replaceable components such as blanket modules, is one of the most critical issues of the tokamak fusion reactor, since the sufficient stiffness against the enormous electromagnetic loads and the easy replaceability are required. In this paper, the authors decide that a caulking cotter joint is worth initiating the R and D from veiwpoints of an effective use of space, a replaceability, a removability of nuclear heating, and a reliability. In this approach, the cotter driving (thrusting and plucking) mechanism is a critical technology. A flexible tube concept has been developed as the driving mechanism, where the stroke and driving force are obtained by a fat shape by the hydraulic pressure. The original normal tube is subjected to the working percentage of more than several hundreds percentage (from thickness of 1.2 mm to 0.2 mm) for plastically forming the flexible tube

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

Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

International Nuclear Information System (INIS)

Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

2002-01-01

these works is also a part of further Kazakhstan contribution in ITER project. JSC 'UMP' provides feasibility study of manufacturing possibility of complete details from beryllium together with superconductor components taking into consideration ITER project requirements. It is supposed that new grades of beryllium, such as SS65C or DSHG-200, will be developed or alternative grade of beryllium with its further verification will be used. As a support for project ITER participation as well as for development of new contemporary trends in science and technology development, and for training the high skilled personnel a unique and the only one in the world KTM experimental tokamak installation is being developed in the country. It will be used for testing structural materials and important elements of thermonuclear reactors. At present a draft of scientific program for KTM reactor researches has been worked out; also feasibility study has been conducted, preliminary design is completed and the work of design documentation development and modeling of the installation is got under way. KTM tokamak, which is constructed in NNC RK in partnership with Russian organizations, will allow to perform predicative tests of future ITER reactor assemblies being produced at JSC 'UMP'. It will promote international cooperation in the field of thermonuclear material science and new material production. It will also contribute greatly in new technologies and modern designs of thermonuclear fusion reactors debugging. ITER and KTM projects are the examples of Successful Eurasia cooperation in the field of peaceful atomic energy use

Tokamak fusion test reactor FELIX plate experiment

International Nuclear Information System (INIS)

Hua, T.O.; Nygren, R.E.; Turner, L.R.

1986-01-01

For a conducting material exposed to both a time-varying and a static magnetic field, such as a limiter blade in a tokamak, the induced eddy currents and the deflection arising from those eddy currents can be strongly coupled. The coupling effects reduce the currents and deflections markedly, sometimes an order of magnitude, from the values predicted if coupling is neglected. A series of experiments to study current-deflection coupling were performed using the Fusion Electromagnetic Inductance Experiment (FELIX) facility at Argonne National Laboratory. Magnetic damping and magnetic stiffness resulting from the coupling are discussed, and analytical expressions for induced eddy current and rigid body rotation in the FELIX plate experiment are compared with the experimental results. Predictions for the degree of coupling based on various parameters are made using the analytical model

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

International Nuclear Information System (INIS)

1988-05-01

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

Tokamak burn cycle study: a data base for comparing long pulse and steady-state power reactors

International Nuclear Information System (INIS)

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

1983-11-01

Several distinct operating modes (conventional ohmic, noninductive steady state, internal transformer, etc.) have been proposed for tokamaks. Our study focuses on capital costs and lifetime limitations of reactor subsystems in an attempt to quantify sensitivity to pulsed operation. Major problem areas considered include: thermal fatigue on first wall, limiter/divertor; thermal energy storage; fatigue and eddy current heating in toroidal field coils; electric power supply costs; and noninductive driver costs. We assume a high availability and low cost of energy will be mandatory for a commercial fusion reactor, and we characterize improvements in physics (current drive efficiency) and engineering (superior materials) which will help achieve these goals for different burn cycles

Integrated modeling of temperature profiles in L-mode tokamak discharges

Energy Technology Data Exchange (ETDEWEB)

Rafiq, T.; Kritz, A. H.; Tangri, V. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Voitsekhovitch, I. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Budny, R. V. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

2014-12-15

Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.

Mathematical modeling plasma transport in tokamaks

Energy Technology Data Exchange (ETDEWEB)

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

1997-01-01

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

Mathematical modeling plasma transport in tokamaks

International Nuclear Information System (INIS)

Quiang, Ji

1995-01-01

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

The ARIES-III D-3He tokamak reactor: Design-point determination and parametric studies

International Nuclear Information System (INIS)

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

1991-01-01

The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-3 design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. Results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-1 is included. 11 refs., 5 figs

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Energy conversion options for ARIES-III - A conceptual D-3He tokamak reactor

International Nuclear Information System (INIS)

Santarius, J.F.; Blanchard, J.P.; Emmert, G.A.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Ghoneim, N.M.; Hasan, M.Z.; Mau, T.K.; Greenspan, E.; Herring, J.S.; Kernbichler, W.; Klein, A.C.; Miley, G.H.; Miller, R.L.; Peng, Y.K.M.

1989-01-01

The potential for highly efficient conversion of fusion power to electricity provides one motivation for investigating D- 3 He fusion reactors. This stems from: (1) the large fraction of D- 3 He power produced in the forms of charged particles and synchrotron radiation which are amenable to direct conversion, and (2) the low neutron fluence and lack of tritium breeding constraints, which increase design flexibility. The design team for a conceptual D- 3 He tokamak reactor, ARIES-III, has investigated numerous energy conversion options at a scoping level in attempting to realize high efficiency. The energy conversion systems have been studied in the context of their use on one or more of three versions of a D- 3 He tokamak: a first stability regime device, a second stability regime device, and a spherical torus. The set of energy conversion options investigated includes bootstrap current conversion, compression-expansion cycles, direct electrodynamic conversion, electrostatic direct conversion, internal electric generator, liquid metal heat engine blanket, liquid metal MHD, plasma MHD, radiation boiler, scrape-off layer thermoelectric, synchrotron radiation conversion by rectennas, synchrotron radiation conversion by thermal cycles, thermionic/AMTEC/thermal systems, and traveling wave conversion. The original set of options is briefly discussed, and those selected for further study are described in more detail. The four selected are liquid metal MHD, plasma MHD, rectenna conversion, and direct electrodynamic conversion. Thermionic energy conversion is being considered, and some options may require a thermal cycle in parallel or series. 17 refs., 3 figs., 1 tab

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)

Operation and control of high density tokamak reactors

International Nuclear Information System (INIS)

Attenberger, S.E.; McAlees, D.G.

1976-01-01

The incentive for high density operation of a tokamak reactor was discussed. It is found that high density permits ignition in a relatively small, moderately elongated plasma with a moderate magnetic field strength. Under these conditions, neutron wall loadings approximately 4 MW/m 2 must be tolerated. The sensitivity analysis with respect to impurity effects shows that impurity control will most likely be necessary to achieve the desired plasma conditions. The charge exchange sputtered impurities are found to have an important effect so that maintaining a low neutral density in the plasma is critical. If it is assumed that neutral beams will be used to heat the plasma to ignition, high energy injection is required (approximately 250 keV) when heating is accompished at full density. A scenario is outlined where the ignition temperature is established at low density and then the fueling rate is increased to attain ignition. This approach may permit beams with energies being developed for use in TFTR to be successfully used to heat a high density device of the type described here to ignition

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Radiation environment of fusion experimental reactor

International Nuclear Information System (INIS)

Mori, Seiji; Seki, Yasushi

1988-01-01

Next step device (experimental reactor), which is planned to succeed the large plasma experimental devices such as JT-60, JET and TFTR, generates radiation (neutron + gamma ray) during its operation. Radiation (neutronic) properties of the material are basis for the study on neutron utilization (energy recovery and tritium breeding), material selection (irradiation damage and lifetime evaluation) and radiation safety (personnel exposure and radiation waste). It is necessary, therefore, to predict radiation behaviour in the reactor correctly for the engineering design of the reactor. This report describes the outline of the radiation environment of the reactor based on the information obtained by the neutronic and shielding design calculation of the fusion experimental reactor (FER). (author)

The impact of confinement scaling on ITER [International Thermonuclear Experimental Reactor] parameters

International Nuclear Information System (INIS)

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

1988-09-01

Energy confinement scaling is a major concern in the design of the International Thermonuclear Experimental Reactor (ITER). The existing database for tokamaks can be fitted with a number of different confinement scaling expressions that have similar degrees of approximation. These scaling laws predict confinement times for ITER that vary by over an order of magnitude. The uncertainties in the form and magnitude of these scaling laws must be substantially reduced before the plasma performance of ITER can be predicted with adequate reliability. The TETRA systems code is used to calculate the dependence of major ITER parameters on the scaling laws currently in use. Design constraints of interest in the present phase of ITER consideration are used, and the minimum-cost devices arising from these constraints are reviewed. 9 refs., 13 figs., 4 tabs

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

Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

1995-12-31

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

MHD equilibrium methods for ITER [International Thermonuclear Experimental Reactor] PF [poloidal field] coil design and systems analysis

International Nuclear Information System (INIS)

Strickler, D.J.; Galambos, J.D.; Peng, Y.K.M.

1989-03-01

Two versions of the Fusion Engineering Design Center (FEDC) free-boundary equilibrium code designed to computer the poloidal field (PF) coil current distribution of elongated, magnetically limited tokamak plasmas are demonstrated and applied to the systems analysis of the impact of plasma elongation on the design point of the International Thermonuclear Experimental Reactor (ITER). These notes were presented at the ITER Specialists' Meeting on the PF Coil System and Operational Scenario, held at the Max Planck Institute for Plasma Physics in Garching, Federal Republic of Germany, May 24--27, 1988. 8 refs., 6 figs., 4 tabs

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

DEFF Research Database (Denmark)

Ondac, Peter; Horacek, Jan; Seidl, Jakub

2015-01-01

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

Power balance in an Ohmically heated fusion reactor

International Nuclear Information System (INIS)

Christiansen, J.P.; Roberts, K.V.

1982-01-01

A simplified power-balance equation (zero-dimensional model) is used to study the performance of an Ohmically heated fusion reactor with emphasis on a pulsed reversed-field pinch concept (RFP). The energy confinement time tausub(E) is treated as an adjustable function, and empirical tokamak scaling laws are employed in the numerical estimates, which are supplemented by 1-D ATHENE code calculations. The known heating rates and energy losses are represented by the net energy replacement time tausub(W), which is exhibited as a surface in density (n) and temperature (T) space with a saddle point (nsub(*), Tsub(*)), the optimum ignition point. It is concluded that i) ignition by Ohmic heating is more practicable for the RFP reactor than for a tokamak reactor with the same tausub(E), (ii) if at fixed current the minor radius can be reduced or at fixed minor radius the current can be increased, then it is found that Ohmic ignition becomes more likely when present tokamak scaling laws are used. More definitive estimates require, however, a knowledge of tausub(E), which can only be obtained by establishing a reliable set of experimental RFP scaling laws and, in particular, by extending RFP experiments closer to the reactor regime. (author)

Enhancement of confinement in tokamaks

International Nuclear Information System (INIS)

Furth, H.P.

1986-01-01

The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

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

Status of fusion reactor concept development in Japan

International Nuclear Information System (INIS)

Tsuji-Iio, Shunji

1996-01-01

Fusion power reactor studies in Japan based on magnetic confinement schemes are reviewed. As D-T fusion reactors, a steady-state tokamak reactor (SSTR) was proposed and extensively studied at the Japan Atomic Energy Research Institute (JAERI) and an inductively operated day-long tokamak reactor (IDLT) was proposed by a group at the University of Tokyo. The concept of a drastically easy maintenance (DREAM) tokamak reactor is being developed at JAERI. A high-field tokamak reactor with force-balanced coils as a volumetric neutron source is being studied by our group at Tokyo Institute of Technology. The conceptual design of a force-free helical reactor (FFHR) is under way at the National Institute for Fusion Science. A design study of a D- 3 He field-reversed configuration (FRC) fusion reactor called ARTEMIS was conducted by the FRC fusion working group of research committee of lunar base an lunar resources. (author)

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

A Study of Performance in Low-Power Tokamak Reactor with Integrated Predictive Modeling Code

International Nuclear Information System (INIS)

Pianroj, Y.; Onjun, T.; Suwanna, S.; Picha, R.; Poolyarat, N.

2009-07-01

Full text: A fusion hybrid or a small fusion power output with low power tokamak reactor is presented as another useful application of nuclear fusion. Such tokamak can be used for fuel breeding, high-level waste transmutation, hydrogen production at high temperature, and testing of nuclear fusion technology components. In this work, an investigation of the plasma performance in a small fusion power output design is carried out using the BALDUR predictive integrated modeling code. The simulations of the plasma performance in this design are carried out using the empirical-based Mixed Bohm/gyro Bohm (B/gB) model, whereas the pedestal temperature model is based on magnetic and flow shear (δ α ρ ζ 2 ) stabilization pedestal width scaling. The preliminary results using this core transport model show that the central ion and electron temperatures are rather pessimistic. To improve the performance, the optimization approach are carried out by varying some parameters, such as plasma current and power auxiliary heating, which results in some improvement of plasma performance

Base isolation technique for tokamak type fusion reactor using adaptive control

International Nuclear Information System (INIS)

Koizumi, T.; Tsujiuchi, N.; Kishimoto, F.; Iida, H.; Fujita, T.

1991-01-01

In this paper relating to the isolation device of heavy structure such as nuclear fusion reactor, a control rule for reducing the response acceleration and relative displacement simultaneously was formulated, and the aseismic performance was improved by employing the adaptive control method of changing the damping factors of the system adaptively every moment. The control rule was studied by computer simulation, and the aseismic effect was evaluated in an experiment employing a scale model. As a results, the following conclusions were obtained. (1) By employing the control rule presented in this paper, both absolute acceleration and relative displacement can be reduced simultaneously without making the system unstable. (2) By introducing this control rule in a scale model assuming the Tokamak type fusion reactor, the response acceleration can be suppressed down to 78 % and also the relative displacement to 79 % as compared with the conventional aseismic method. (3) The sensitivities of absolute acceleration and relative displacement with respect to the control gain are not equal. However, by employing the relative weighting factor between the absolute acceleration and relative displacement, it is possible to increase the control capability for any kind of objective structures and appliances. (author)

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

International Nuclear Information System (INIS)

Cavalluzzo, S.

1983-01-01

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

Present status of Tokamak research

International Nuclear Information System (INIS)

Basu, Jayanta

1991-01-01

The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

MIRI: A multichannel far-infrared laser interferometer for electron density measurements on TFTR [Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Mansfield, D.K.; Park, H.K.; Johnson, L.C.; Anderson, H.M.; Chouinard, R.; Foote, V.S.; Ma, C.H.; Clifton, B.J.

1987-07-01

A ten-channel far-infrared laser interferometer which is routinely used to measure the spatial and temporal behavior of the electron density profile on the TFTR tokamak is described and representative results are presented. This system has been designed for remote operation in the very hostile environment of a fusion reactor. The possible expansion of the system to include polarimetric measurements is briefly outlined. 13 refs., 8 figs

ITER: the first experimental fusion reactor

International Nuclear Information System (INIS)

Rebut, P.H.

1995-01-01

The International Thermonuclear Experimental Reactor (ITER) project is a multiphased project, at present proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement between the European Atomic Energy Community, the Government of Japan, the Government of the USA and the Government of Russia (''the parties''). The project is based on the tokamak, a Russian invention which has been brought to a high level of development and progress in all major fusion programs throughout the world.The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for commercial energy production and to test technologies for a demonstration fusion power plant. During the extended performance phase of ITER, it will demonstrate the characteristics of a fusion power plant, producing more than 1500MW of fusion power.The objective of the engineering design activity (EDA) phase is to produce a detailed, complete and fully integrated engineering design of ITER and all technical data necessary for the future decision on the construction of ITER.The ITER device will be a major step from present fusion experiments and will encompass all the major elements required for a fusion reactor. It will also require the development and the implementation of major new components and technologies.The inside surface of the plasma containment chamber will be designed to withstand temperature of up to 500 C, although normal operating temperatures will be substantially lower. Materials will have to be carefully chosen to withstand these temperatures, and a high neutron flux. In addition, other components of the device will be composed of state-of-the-art metal alloys, ceramics and composites, many of which are now in the early stage of development of testing. (orig.)

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

International Nuclear Information System (INIS)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

Energy Technology Data Exchange (ETDEWEB)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

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

Science.gov (United States)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

Fusion reactor remote maintenance study. Final report

International Nuclear Information System (INIS)

Sniderman, M.

1979-04-01

An analysis of a major maintenance operation, the remote replacement of a modular sector of a tokamak reactor, was performed in substantial detail. Specific assumptions were developed which included concepts from various existing designs so that the operation which was studied includes some design features generic to any fusion reactor design. Based on the work performed in this study, the principal conclusions are: (1) It appears feasible to design a tokamak fusion reactor plant with availability comparable to existing fossil and fission plants, but this will require diligence and comprehensive planning during the complete design phase. (2) Since the total fusion program is paced by the success of each device, maintenance considerations must be incorporated into each device during design, even if the device is an experimental unit. (3) Innovative approaches, such as automatic computer controlled operations, should be developed so that large step reductions in planned maintenance times can be achieved

A probabilistic methodology for the design of radiological confinement of tokamak reactors

International Nuclear Information System (INIS)

Golinescu, Ruxandra P.; Morosan, Florinel; Kazimi, Mujid S.

1997-01-01

A methodology using probabilistic risk assessment techniques is proposed for evaluating the design of multiple confinement barriers for a fusion plant within the context of a limited allowable risk. The methodology was applied to the reference design of the International Thermonuclear Experimental Reactor (ITER). Accident sequence models were developed to determine the probability of radioactive releases from each confinement barrier. The current ITER design requirements, that set environmental radioactive release limits for individual event sequences grouped in categories by frequency, is extended to derive a limit on the plant overall risk. This avoids detailed accounting for event uncertainties in both frequency and consequence. Thus, an analytical form for a limit line is derived as a complementary cumulative frequency of permissible radioactive releases to the environment. The line can be derived using risk aversion of the designer's own choice. By comparing the releases from each confinement barrier against this limit line, a decision can be made about the number of barriers required to comply with the design requirements. A decision model using multi-attribute utility function theory was constructed to help the designer in choosing the type of the tokamak building while considering preferences for attributes such as construction cost, project completion time, technical feasibility and public attitude. Sensitivity analysis on some of the relevant parameters in the model was performed

Tokamak and RFP ignition requirements

International Nuclear Information System (INIS)

Werley, K.A.

1991-01-01

A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

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

Science.gov (United States)

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

2017-11-01

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

Extremely shaped plasmas to improve the Tokamak concept

Energy Technology Data Exchange (ETDEWEB)

Piras, F.

2011-04-15

experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is

Extremely shaped plasmas to improve the Tokamak concept

International Nuclear Information System (INIS)

Piras, F.

2011-04-01

experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is

Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

International Nuclear Information System (INIS)

2009-01-01

In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

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

Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

CERN Document Server

Lumia, M E

2002-01-01

A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

M.E. Lumia; C.A. Gentile

2002-01-01

A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed

Possible fusion reactor

International Nuclear Information System (INIS)

Yoshikawa, S.

1976-05-01

A scheme to improve performance characteristics of a tokamak-type fusion reactor is proposed. Basically, the tokamak-type plasma could be moved around so that the plasma could be heated by compression, brought to the region where the blanket surrounds the plasma, and moved so as to keep wall loading below the acceptable limit. This idea should be able to help to economize a fusion reactor

Present status of fusion reactor materials, 4

International Nuclear Information System (INIS)

Nagasaki, Ryukichi; Shiraishi, Kensuke; Watanabe, Hitoshi; Murakami, Yoshio; Takamura, Saburo

1982-01-01

Recently, the design of fusion reactors such as Intor has been carried out, and various properties that fusion reactor materials should have been clarified. In the Japan Atomic Energy Research Institute, the research and development of materials aiming at a tokamak type experimental fusion reactor are in progress. In this paper, the problems, the present status of research and development and the future plan about the surface materials and structural materials for the first wall, blanket materials and magnet materials are explained. The construction of the critical plasma testing facility JT-60 developed by JAERI has progressed smoothly, and the operation is expected in 1985. The research changes from that of plasma physics to that of reactor technology. In tokamak type fusion reactors, high temperature D-T plasma is contained with strong magnetic field in vacuum vessels, and the neutrons produced by nuclear reaction, charged particles diffusing from plasma and neutral particles by charge exchange strike the first wall. The PCA by improving 316 stainless steel is used as the structural material, and TiC coating techniques are developed. As the blanket material, Li 2 O is studied, and superconducting magnets are developed. (Koko, I.)

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

International Nuclear Information System (INIS)

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

1977-12-01

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

Experimental and calculating study on the stressed state of superconducting coils of toroidal field in the T-15 tokamak

International Nuclear Information System (INIS)

Vaulina, I.G.; Gusev, S.V.; Sivkova, G.N.

1987-01-01

Results of calculational and experimental atudy of stress-deformed state of superconducting coils of the T-15 tokamak toroidal field are presented. The calculations are made using the method of finite elements and refined theory of cores. Experimental studies were carried out using elastic tensometric model of polymer materials. Test results are compared with the calculational results. Divergence between calculational and experimental values of displacement of characteristic points in the unit does not exceed 20 %. Results of model studies confirm the expediency of the calculational model used for designing SOTP unit for the T-15 tokamak

Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

Science.gov (United States)

Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

2017-10-01

The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-11-15

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

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

International Nuclear Information System (INIS)

Fedorczak, N.

2010-01-01

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

Experimental fusion power reactor conceptual design study. Final report. Volume III

International Nuclear Information System (INIS)

Baker, C.C.

1976-12-01

This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following appendices: (1) tradeoff code analysis, (2) residual mode transport, (3) blanket/first wall design evaluations, (4) shielding design evaluation, (5) toroidal coil design evaluation, (6) E-coil design evaluation, (7) F-coil design evaluation, (8) plasma recycle system design evaluation, (9) primary coolant purification design evaluation, (10) power supply system design evaluation, (11) number of coolant loops, (12) power conversion system design evaluation, and (13) maintenance methods evaluation

Plasma density remote control system of experimental advanced superconductive tokamak

International Nuclear Information System (INIS)

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

2007-01-01

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

Fusion potential for spherical and compact tokamaks

International Nuclear Information System (INIS)

Sandzelius, Mikael

2003-02-01

The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

Fusion potential for spherical and compact tokamaks

Energy Technology Data Exchange (ETDEWEB)

Sandzelius, Mikael

2003-02-01

The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

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

Review of tearing mode stabilization by RF power in tokamaks

International Nuclear Information System (INIS)

Giruzzi, G.; Zabiego, M.; Zohm, H.

1999-01-01

Control of tearing modes by means of heating and current drive inside the magnetic islands is one of the most important applications of RF power in tokamak reactors. The theoretical basis of this concept is reviewed, focusing on aspects related to RF-plasma interaction. Applications to the stabilization of neoclassical tearing modes in ITER by Electron Cyclotron Current Drive are presented to illustrate the basic physical dependences. The most significant experimental results and prospects for future applications are also discussed

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

Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

International Nuclear Information System (INIS)

Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

1985-01-01

Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed