Tokamak-like confinement at high beta and low field in the reversed field pinch

International Nuclear Information System (INIS)

Sarff, J S; Anderson, J K; Biewer, T M; Brower, D L; Chapman, B E; Chattopadhyay, P K; Craig, D; Deng, B; Hartog, D J Den; Ding, W X; Fiksel, G; Forest, C B; Goetz, J A; O'Connell, R; Prager, S C; Thomas, M A

2003-01-01

For several reasons, improved-confinement achieved in the reversed field pinch (RFP) during the last few years can be characterized as 'tokamak-like'. Historically, RFP plasmas have had relatively poor confinement due to tearing instability which causes magnetic stochasticity and enhanced transport. Tearing reduction is achieved through modification of the inductive current drive, which dramatically improves confinement. The electron temperature increases to >1 keV and the electron heat diffusivity decreases to approx. 5 m 2 s -1 , comparable with the transport level expected in a tokamak plasma of the same size and current. This corresponds to a 10-fold increase in global energy confinement. Runaway electrons are confined, and Fokker-Planck modelling of the electron distribution reveals that the diffusion at high energy is independent of the parallel velocity, uncharacteristic of stochastic transport. Improved-confinement occurs simultaneously with increased beta approx. 15%, while maintaining a magnetic field strength ten times weaker than a comparable tokamak. Measurements of the current, magnetic, and electric field profiles show that a simple Ohm's Law applies to this RFP sustained without dynamo relaxation

Axisymmetric stability of vertically asymmetric tokamaks at large beta poloidal

International Nuclear Information System (INIS)

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

1981-09-01

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

Proposal for the construction of a High-Beta Tokamak at LASL

International Nuclear Information System (INIS)

Van der Laan, P.C.T.; Freidberg, J.P.; Thomas, K.S.

1976-06-01

The large heating rate inherent to implosion heating allows the rapid generation of high-beta tokamak plasmas. A study of these plasmas in the proposed HBT machine can give information on how MHD equilibrium and stability limit β and q. Both a wide current profile and a moderate elongation of the minor cross section should help to raise the permissible peak β values in HBT to at least 20 percent. The longer term loss processes occurring in MHD-stable plasmas are to be investigated. The main parameters of HBT are: R = 0.30 m, minor cross section a racetrack of width and height 0.24 m and 0.48 m, B/sub phi/ = 2 T, I/sub phi/ approximately 750 kA

High-beta characteristics of first and second-stable spherical tokamaks in reconnection heating experiments of TS-3

International Nuclear Information System (INIS)

Ono, Y.

2002-01-01

Novel formations of ultra-high-beta Spherical Tokamak (ST) have been developed in the TS-3 device using high power heating of merging/ reconnection. In Type-A merging, two STs were merged together to build up the plasma beta. In Type-B merging, an oblate FRC was initially formed by merging of two spheromaks with opposing toroidal field B t and was transformed into an ultra-high-beta ST by applying external B t . Ballooning stability analyses confirmed formations of the first-stable STs by Type- A merging and the second-stable STs by Type-B merging and also the unstable STs by both mergings, revealing the ballooning stability window consistent with measured high-n instabilities. We made (1) those model analyses of the produced STs for the first time using the BALLOO stability code, revealing that hollowness/ broadness of current/pressure profiles widen significantly the window to the second-stable regime. This paper also addresses (2) normalized betas of the second-stable STs as large as 6-17 for comparison with the Troyon scaling and (3) a promising scaling of the reconnection heating energy. (author)

Experimental observations of MHD instabilities in the high-beta tokamak Torus-II

International Nuclear Information System (INIS)

Machida, M.

1982-01-01

The CO 2 laser scattering and interferometry diagnostics have been used to study the MHD instabilities in the high-beta tokamak Torus-II. Detailed measurements of the density and density fluctuation profiles have been performed. In order to measure density fluctuations with wavelengths longer than 2 cm, an interferometric like, phase matching technique has been developed. The toroidal and poloidal mode numbers have been measured using a double-beam, two-position technique. Working at high-beta values, average β greater than or equal to 10%, we have found parameters where the growing instabilities are created or suppressed. The plasma lifetime for both cases is seen to be about the same and the loss of the plasma appears to be caused by the decay in the external fields. The growing instability parameters are within the MHD regime, and it only grows at the outer edge of the plasma. This is in agreement with the theoretical Ballooning mode instability. The frequency and mode number measurements also agree with the Kinetic theory description of Ballooning modes. The comparison with possible other modes, such as Tearing and Drift instabilities, is performed and the Ballooning growth rate is shown to be the best fit to the experimental values

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

Observations of plasma rotation in the high-beta tokamak Torus II

International Nuclear Information System (INIS)

Kostek, C.; Marshall, T.C.

1982-01-01

Toroidal and poloidal plasma rotation are measured in a high Beta tokamak device by studying the Doppler shift of the 4686 A He II line. The toroidal flow motion is in the same direction as the plasma current at an average velocity of 1.6 x 10 6 cm/sec, a small fraction of the ion thermal speed. The poloidal flow follows the ion diamagnetic direction, also at an average speed of 1.6 x 10 6 cm/sec. In view of certain ordering parameters, the toroidal flow is compared with the predictions of neoclassical transport theory in the collisional regime. For the poloidal motion, however, it appears that an (E/sub r/ x B)/B 2 drift in a positive radial electric field, approaching a stable ambipolar state (STRINGER, 1970) is responsible. Mechanisms for the time evolution of the rotation are also examined. The radial electric field responsible for the (E/sub r/ x B)/B 2 drift is determined from the theory using the measured poloidal velocity

High Beta Tokamak research

International Nuclear Information System (INIS)

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

1993-01-01

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

High performance operational limits of tokamak and helical systems

International Nuclear Information System (INIS)

Yamazaki, Kozo; Kikuchi, Mitsuru

2003-01-01

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

Simulations of finite beta turbulence in tokamaks and stellarators

International Nuclear Information System (INIS)

Jenko, F.

2002-01-01

One of the central open questions in our attempt to understand microturbulence in fusion plasmas concerns the role of finite beta effects. Nonlinear codes trying to investigate this issue must go beyond the commonly used adiabatic electron approximation - a task which turns out to be a serious computational challenge. This step is necessary because the electrons are the prime contributor to the parallel currents which in turn produce the magnetic field fluctuations. Results at both ion and electron space-time scales from gyrokinetic and gyrofluid models are presented which shed light on the character of finite beta turbulence in tokamaks and stellarators. (author)

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

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

The high beta tokamak-extended pulse magnetohydrodynamic mode control research program

International Nuclear Information System (INIS)

Maurer, D A; Bialek, J; Byrne, P J; De Bono, B; Levesque, J P; Li, B Q; Mauel, M E; Navratil, G A; Pedersen, T S; Rath, N; Shiraki, D

2011-01-01

The high beta tokamak-extended pulse (HBT-EP) magnetohydrodynamic (MHD) mode control research program is studying ITER relevant internal modular feedback control coil configurations and their impact on kink mode rigidity, advanced digital control algorithms and the effects of plasma rotation and three-dimensional magnetic fields on MHD mode stability. A new segmented adjustable conducting wall has been installed on the HBT-EP and is made up of 20 independent, movable, wall shell segments instrumented with three distinct sets of 40 saddle coils, totaling 120 in-vessel modular feedback control coils. Each internal coil set has been designed with varying toroidal angular coil coverage of 5, 10 and 15 0 , spanning the toroidal angle range of an ITER port plug based internal coil to test resistive wall mode (RWM) interaction and multimode MHD plasma response to such highly localized control fields. In addition, we have implemented 336 new poloidal and radial magnetic sensors to quantify the applied three-dimensional fields of our control coils along with the observed plasma response. This paper describes the design and implementation of the new control shell incorporating these control and sensor coils on the HBT-EP, and the research program plan on the upgraded HBT-EP to understand how best to optimize the use of modular feedback coils to control instability growth near the ideal wall stabilization limit, answer critical questions about the role of plasma rotation in active control of the RWM and the ferritic resistive wall mode, and to improve the performance of MHD control systems used in fusion experiments and future burning plasma systems.

Analytic, High-beta Solutions of the Helical Grad-Shafranov Equation

International Nuclear Information System (INIS)

Smith, D.R.; Reiman, A.H.

2004-01-01

We present analytic, high-beta (β ∼ O(1)), helical equilibrium solutions for a class of helical axis configurations having large helical aspect ratio, with the helix assumed to be tightly wound. The solutions develop a narrow boundary layer of strongly compressed flux, similar to that previously found in high beta tokamak equilibrium solutions. The boundary layer is associated with a strong localized current which prevents the equilibrium from having zero net current

Scaling of energy confinement and poloidal beta in high density tokamaks

NARCIS (Netherlands)

Schram, D.C.; Schüller, F.C.

1980-01-01

A semi-empirical analysis of the heat balance of ohmically heated, high density Tokamak plasmas, shows that the observed heat transport can be explained by neoclassical (plateau) ion heat conduction in the central part of the plasma. Experimental values for Te, ß¿e, and tEe and the variation of

Study of critical beta non-circular tokamak equilibria sustained in steady state by beam driven currents

International Nuclear Information System (INIS)

Okano, K.; Ogawa, Y.; Naitou, H.

1988-07-01

A new MHD-equilibrium/current-drive analysis code was developed to analyse the high beta tokamak equilibria consistent with the beam driven current profiles. In this new code, the critical beta equilibrium, which is stable against the ballooning mode, the kink mode and the Mercier mode, is determined first using MHD equilibrium and stability analysis codes (EQLAUS/ERATO). Then, the current drive parameters and the plasma parameters, required to sustain this critical beta equilibrium, are determined by iterative calculations. The beam driven current profiles are evaluated by the Fokker-Planck calculations on individual flux surfaces, where the toroidal effects on the beam ion and plasma electron trajectories are considered. The pressure calculation takes into account the beam ion and fast alpha components. A peculiarity of our new method is that the obtained solution is not only consistent with the MHD equilibrium but also consistent with the critical beta limit conditions, in the current profile and the pressure profile. Using this new method, β ∼ 21 % bean and β ∼ 6 % D-type critical beta equilibria were scanned for various parameters; the major radius, magnetic field, temperature, injection energy, etc. It was found that the achievable Q value for the bean type was always about 30 % larger than for the D-type cases, where Q = fusion power/beam power. With strong beanness, Q ∼ 6 for DEMO type tokamaks (∼500 MWth) and Q ∼ 20 for power reactor size (4.5 GWth) are achievable. On the other hand, the Q value would not exceed sixteen for the D-type machines. (author)

System assessment of helical reactors in comparison with tokamaks

International Nuclear Information System (INIS)

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

2002-10-01

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

Recent advancement in research and planning toward high beta steady state operation in KSTAR

International Nuclear Information System (INIS)

Park, Hyeon Keo; Hong, S.; Humphreys, D.

2015-01-01

The goal of Korean Superconducting Tokamak Advanced Research (KSTAR) research is to explore stable improved confinement regimes and technical challenge for superconducting tokamak operation and thus, to establish the basis for predictable high beta steady state tokamak plasma operation. To fulfil the goal, the current KSTAR research program is composed of three elements: 1) Exploration of anticipated engineering and technology for a stable long pulse operation of high beta plasmas including Edge Localized Mode (ELM) control with the low n (=1, 2) Resonant Magnetic Perturbation (RMP) using in-vessel control coils and innovative non-inductive current drives. The achieved long pulse operation up to ∼50s and fully non-inductive current drive will be combined in the future. Study of efficient heat exhaust will be combined with an innovative divertor design/operation. 2) Exploration of the operation boundary through establishment of true stability limits of the harmful MagnetoHydroDynamic (MHD) instabilities and confinement of the tokamak plasmas in KSTAR, making use of the lowest error field and magnetic ripple simultaneously achieved among all tokamaks ever built. The intrinsic machine error field has a long history of research as the source of MHD instabilities and magnetic ripple is known to be a cause of energy loss in the plasma. The achieved high beta discharges at β N ∼4 and stable discharges at q 95 (∼2) will be further improved. 3) Validation of theoretical modeling of MHD instabilities and turbulence toward predictive capability of stable high beta plasmas. In support of these research goals, the state of the art diagnostic systems, such as Electron Cyclotron Emission Imaging (ECEI) system in addition to accurate profile diagnostics, are deployed not only to provide precise 2D/3D information of the MHD instabilities and turbulence but also to challenge unresolved physics problems such as the nature of ELMs, ELM-crash dynamics and the role of the core

Observation of finite-β MHD phenomena in Tokamaks

International Nuclear Information System (INIS)

McGuire, K.M.

1985-01-01

Stable high beta plasmas are required for the tokamak to attain an economical fusion reactor. Recently, intense neutral beam heating experiments in tokamaks have shown new effects on plasma stability and confinement associated with high beta plasmas. The observed spectrum of MHD fluctuations at high beta is clearly dominated by the n = 1 mode when the q = 1 surface is in the plasma. The m/n = 1/1 mode drives other n = 1 modes through toroidal coupling and n > 1 modes through nonlinear coupling. On PDX, with near perpendicular injection, a resonant interaction between the n = 1 internal kink and the trapped fast ions results in loss of beam particles and heating power. Key parameters in the theory are the value of qsub(o) and the injection angle. High frequency broadband magnetic fluctuations have been observed on ISX-B and D-III and a correlation with the deterioration of plasma confinement was reported. During enhanced confinement (H-mode) discharges in divertor plasmas two new edge instabilities were observed, both localized radially near the separatrix. By assembling results from the different tokamak experiments, it is found that the simple theoretical ideal MHD beta limit has not been exceeded

Observations of toroidal and poloidal rotation in the high beta tokamak Torus II

International Nuclear Information System (INIS)

Kostek, C.A.

1983-01-01

The macroscopic rotation of plasma in a toroidal containment device is an important feature of the equilibrium. Toroidal and poloidal rotation in the high beta tokamak Torus II is measured experimentally by examining the Doppler shift of the 4685.75 A He II line emitted from the plasma. The toroidal flow at an average velocity of 1.6 x 10 6 cm/sec, a small fraction of the ion thermal speed, moves in the same direction as the toroidal plasma current. The poloidal flow follows the ion diamagnetic current direction, also at an average speed of 1.6 x 10 6 cm/sec. In view of certain ordering parameters, the toroidal flow is compared with predictions from neoclassical theory in the collosional, Pfirsch-Schluter regime. The poloidal motion, however results from an E x B drift in a positive radial electric field, approaching a stable ambipolar state. This radial electric field is determined from theory by using the measured poloidal velocity. Mechanisms for the time evolution of rotation are also examined. It appears that the circulation damping is governed by a global decay of the temperature and density gradients which, in turn, may be functions of radiative cooling, loss of equilibrium due to external field decay, or the emergence of a growing instability, occasionally observed in CO 2 interferometry measurements

ZORNOC: a 1 1/2-D tokamak data analysis code for studying noncircular high beta plasmas

International Nuclear Information System (INIS)

Zurro, B.; Wieland, R.M.; Murakami, M.; Swain, D.W.

1980-03-01

A new tokamak data analysis code, ZORNOC, was developed to study noncircular, high beta plasmas in the Impurity Study Experiment (ISX-B). These plasmas exhibit significant flux surface shifts and elongation in both ohmically heated and beam-heated discharges. The MHD equilibrium flux surface geometry is determined by solving the Grad-Shafranov equation based on: (1) the shape of the outermost flux surface, deduced from the magnetic loop probes; (2) a pressure profile, deduced by means of Thomson scattering data (electrons), charge exchange data (ions), and a Fokker-Planck model (fast ions); and (3) a safety factor profile, determined from the experimental data using a simple model (Z/sub eff/ = const) that is self-consistently altered while the plasma equilibrium is iterated. For beam-heated discharches the beam deposition profile is determined by means of a Monte Carlo scheme and the slowing down of the fast ions by means of an analytical solution of the Fokker-Planck equation. The code also carries out an electron power balance and calculates various confinement parameters. The code is described and examples of its operation are given

Development path of low aspect ratio tokamak power plants

International Nuclear Information System (INIS)

Stambaugh, R.D.; Chan, V.S.; Miller, R.L.

1997-03-01

Recent advances in tokamak physics indicate the spherical tokamak may offer a magnetic fusion development path that can be started with a small size pilot plant and progress smoothly to larger power plants. Full calculations of stability to kink and ballooning modes show the possibility of greater than 50% beta toroidal with the normalized beta as high as 10 and fully aligned 100% bootstrap current. Such beta values coupled with 2--3 T toroidal fields imply a pilot plant about the size of the present DIII-D tokamak could produce ∼ 800 MW thermal, 160 MW net electric, and would have a ratio of gross electric power over recirculating power (Q PLANT ) of 1.9. The high beta values in the ST mean that E x B shear stabilization of turbulence should be 10 times more effective in the ST than in present tokamaks, implying that the required high quality of confinement needed to support such high beta values will be obtained. The anticipated beta values are so high that the allowable neutron flux at the blanket sets the device size, not the physics constraints. The ST has a favorable size scaling so that at 2--3 times the pilot plant size the Q PLANT rises to 4--5, an economic range and 4 GW thermal power plants result. Current drive power requirements for 10% of the plasma current are consistent with the plant efficiencies quoted. The unshielded copper centerpost should have an adequate lifetime against nuclear transmutation induced resistance change and the low voltage, high current power supplies needed for the 12 turn TF coil appear reasonable. The favorable size scaling of the ST and the high beta mean that in large sizes, if the copper TF coil is replaced with a superconducting TF coil and a shield, the advanced fuel D-He 3 could be burned in a device with Q PLANT ∼ 4

Computational studies of tokamak plasmas

International Nuclear Information System (INIS)

Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

1981-02-01

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

Sustained high βN plasmas on EAST tokamak

Science.gov (United States)

Gao, Xiang; the EAST team

2018-05-01

Sustained high normalized beta (βN ∼ 1.9) plasmas with an ITER-like tungsten divertor have been achieved on EAST tokamak recently. The high power NBI heating system of 4.8 MW and the 4.6 GHz lower hybrid wave of 1 MW were developed and applied to produce edge and internal transport barriers in high βN discharges. The central flat q profile with q (ρ) ∼ 1 at ρ safety factor q95 = 4.7 is identified by the multi-channel far-infrared laser polarimeter and the EFIT code. The fraction of non-inductive current is about 40%. The relation between fishbone activity and ITB formation is observed and discussed.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Progress Toward Long Pulse, High Performance Plasmas in the DIII-D Tokamak

International Nuclear Information System (INIS)

P.A. Politzer; T.C. Luce; M.E. Austin; J.R. Ferron, A.M. Garofalo; C.M. Greenfield; A.W. Hyatt; R.J. La Haye; L.L. Lao; E.A. Lazarus; M.A. Makowski; M. Murakami; C.C. Petty; R.I. Pinsker; B.W. Rice; E.J. Strait, M.R. Wade; J.G. Watkins

2000-01-01

A major portion of the research program of the DIII-D tokamak collaboration is devoted to the development and demonstration of high performance advanced tokamak plasmas, with profiles as close as possible to those anticipated for steady-state operation. The work during the 1999 campaign has resulted in significant progress toward this goal. High normalized performance ((beta)(sub N)(approx) 4 and(beta)(sub N) H(sub 89)(approx) 9) discharges have been sustained for up to 2 s. These plasmas are in H-mode with rapid ELMs. The most common limiting phenomena are resistive wall modes (RWMs) rather than neoclassical tearing modes (NTMs). NTMs do occur, apparently triggered by the RWMs. The observed pressure is well above the calculated beta limit without a wall, and(beta)(sub N) and gt; 4(ell)(sub i) throughout the high performance phase. The bootstrap current is estimated to be and gt;50% of the total, and measurements of the internal loop voltage show that only about 25% of the current is inductively driven. The central q profile is flat, as is the calculated bootstrap current profile, due to the absence of any localized pressure gradients. The residual inductive current is localized around r/a(approx) 0.5. To demonstrate quasi-stationary operation, it will be necessary to replace the residual inductive current with ECCD at the same minor radius. To effectively apply ECH and ECCD to these discharges, density control will be needed. Preliminary experiments using the DIII-D cryopump have reduced the density by(approx)20%. A new EC power system and a new private flux cryopump will be available for the 2000 campaign

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

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.

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

Advanced tokamak physics in DIII-D

Energy Technology Data Exchange (ETDEWEB)

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

2000-12-01

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

Physics issues of high bootstrap current tokamaks

International Nuclear Information System (INIS)

Ozeki, T.; Azumi, M.; Ishii, Y.

1997-01-01

Physics issues of a tokamak plasma with a hollow current profile produced by a large bootstrap current are discussed based on experiments in JT-60U. An internal transport barrier for both ions and electrons was obtained just inside the radius of zero magnetic shear in JT-60U. Analysis of the toroidal ITG microinstability by toroidal particle simulation shows that weak and negative shear reduces the toroidal coupling and suppresses the ITG mode. A hard beta limit was observed in JT-60U negative shear experiments. Ideal MHD mode analysis shows that the n = 1 pressure-driven kink mode is a plausible candidate. One of the methods to improve the beta limit against the kink mode is to widen the negative shear region, which can induce a broader pressure profile resulting in a higher beta limit. The TAE mode for the hollow current profile is less unstable than that for the monotonic current profile. The reason is that the continuum gaps near the zero shear region are not aligned when the radius of q min is close to the region of high ∇n e . Finally, a method for stable start-up for a plasma with a hollow current profile is describe, and stable sustainment of a steady-state plasma with high bootstrap current is discussed. (Author)

Wall stabilization of high beta plasmas in DIII-D

International Nuclear Information System (INIS)

Taylor, T.S.; Strait, E.J.; Lao, L.L.; Turnbull, A.D.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Groebner, R.J.; La Haye, R.J.; Mauel, M.

1995-02-01

Detailed analysis of recent high beta discharges in the DIII-D tokamak demonstrates that the resistive vacuum vessel can provide stabilization of low n magnetohydrodynamic (MHD) modes. The experimental beta values reaching up to β T = 12.6% are more than 30% larger than the maximum stable beta calculated with no wall stabilization. Plasma rotation is essential for stabilization. When the plasma rotation slows sufficiently, unstable modes with the characteristics of the predicted open-quotes resistive wallclose quotes mode are observed. Through slowing of the plasma rotation between the q = 2 and q = 3 surfaces with the application of a non-axisymmetric field, the authors have determined that the rotation at the outer rational surfaces is most important, and that the critical rotation frequency is of the order of Ω/2π = 1 kHz

D-D tokamak reactor assessment

International Nuclear Information System (INIS)

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

1983-01-01

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

Physics design requirements for the Tokamak Physics Experiment (TPX)

International Nuclear Information System (INIS)

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

1993-01-01

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

Mercier criterion for high-β tokamaks

International Nuclear Information System (INIS)

Galvao, R.M.O.

1984-01-01

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

Transport of carbon ion test particles and hydrogen recycling in the plasma of the Columbia tokamak ''HBT'' [High Beta Tokamak

International Nuclear Information System (INIS)

Wang, Jian-Hua.

1990-01-01

Carbon impurity ion transport is studied in the Columbia High Beta Tokamak (HBT), using a carbon tipped probe which is inserted into the plasma (n e ∼ 1 - 5 x 10 14 (cm -3 ), T e ∼ 4 - 10 (eV), B t ∼ 0.2 - 0.4(T)). Carbon impurity light, mainly the strong lines of C II (4267A, emitted by the C + ions) and C III (4647A, emitted by the C ++ ions), is formed by the ablation or sputtering of plasma ions and by the discharge of the carbon probe itself. The diffusion transport of the carbon ions is modeled by measuring the space-and-time dependent spectral light emission of the carbon ions with a collimated optical beam and photomultiplier. The point of emission can be observed in such a way as to sample regions along and transverse to the toroidal magnetic field. The carbon ion diffusion coefficients are obtained by fitting the data to a diffusion transport model. It is found that the diffusion of the carbon ions is ''classical'' and is controlled by the high collisionality of the HBT plasma; the diffusion is a two-dimensional problem and the expected dependence on the charge of the impurity ion is observed. The measurement of the spatial distribution of the H α emissivity was obtained by inverting the light signals from a 4-channel polychromator, the data were used to calculate the minor-radial influx, the density, and the recycling time of neutral hydrogen atoms or molecules. The calculation shows that the particle recycling time τ p is comparable with the plasma energy confinement time τ E ; therefore, the recycling of the hot plasma ions with the cold neutrals from the walls is one of the main mechanisms for loss of plasma energy

Continuous tokamaks

International Nuclear Information System (INIS)

Peng, Y.K.M.

1978-04-01

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

Survey of linear MHD stability in tokamak configurations

International Nuclear Information System (INIS)

Wakatani, M.

1977-01-01

The results found by MHD stability studies for both low-beta and high-beta tokamaks are reviewed. The stability against kink-ballooning modes in equilibria surrounded by vacuum or a layer of force free currents is considered. Internal kink modes and the relation to interchange modes, which should be considered after external kink modes are suppressed, are surveyed

Stability Limits of High-Beta Plasmas in DIII-D

International Nuclear Information System (INIS)

Strait, E.J.

2005-01-01

Stability at high beta is an important requirement for a compact, economically attractive fusion reactor. DIII-D experiments have shown that ideal magnetohydrodynamic (MHD) theory is an accurate predictor of the ultimate stability limits for tokamaks, and the Troyon scaling law has provided a useful approximation of ideal stability limits for discharges with 'conventional' profiles. However, variation of the discharge shape, pressure profile, and current density profile can lead to ideal MHD beta limits that differ significantly from simple Troyon scaling. The need for profiles consistent with steady-state operation places an important additional constraint on plasma stability. Nonideal effects can also be important and must be taken into account. For example, neoclassical tearing modes (NTMs), resulting from plasma resistivity and the nonlinear effects of the bootstrap current, can become unstable at beta values well below the ideal MHD limit. DIII-D experiments are now entering a new era of unprecedented control over plasma stability, including suppression of NTMs by localized current drive at the island location, and direct feedback stabilization of kink modes with a resistive wall. The continuing development of physics understanding and control tools holds the potential for stable, steady-state fusion plasmas at high beta

System studies for quasi-steady-state advanced physics tokamak

International Nuclear Information System (INIS)

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

1983-11-01

Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated

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

High beta and confinement studies on TFTR

International Nuclear Information System (INIS)

Navratil, G.A.; Bhattacharjee, A.; Iacono, R.; Mauel, M.E.; Sabbagh, S.A.; Kesner, J.

1992-01-01

A new regime of high poloidal beta operation in TFTR was developed in the course of the first two years of this project (9/25/89 to 9/24/91). Our proposal to continue this successful collaboration between Columbia University and the Massachusetts Institute of Technology with the Princeton Plasma Physics Laboratory for a three year period (9/25/91 to 9/24/94) to continue to investigate improved confinement and tokamak performance in high poloidal beta plasmas in TFTR through the DT phase of operation was approved by the DOE and this is a report of our progress during the first 9 month budget period of the three year grant (9/25/91 to 6/24/92). During the approved three year project period we plan to (1) extend and apply the low current, high QDD discharges to the operation of TFTR using Deuterium and Tritium plasma; (2) continue the analysis and plan experiments on high poloidal beta phenomena in TFTR including: stability properties, enhanced global confinement, local transport, bootstrap current, and divertor formation; (3) plan and carry out experiments on TFTR which attempt to elevate the central q to values > 2 where entry to the second stability regime is predicted to occur; and (4) collaborate on high beta experiments using bean-shaped plasmas with a stabilizing conducting shell in PBX-M. In the seven month period covered by this report we have made progress in each of these four areas through the submission of 4 TFTR Experimental Proposals and the partial execution of 3 of these using a total of 4.5 run days during the August 1991 to February 1992 run

INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

International Nuclear Information System (INIS)

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

2004-03-01

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

PDX modification to produce a bean-shaped high-beta plasma

International Nuclear Information System (INIS)

Materna, P.; Chrzanowski, J.; Heitzenroeder, P.; Lee, K.; Pereira, M.

1983-01-01

Princeton's PDX tokamak is being converted to produce bean-shaped plasmas which hopefully will reach beta of 10%. The work, which is nearly complete, involves repositioning active coils, adding passive coils, and making external modifications

Computational Study of Anomalous Transport in High Beta DIII-D Discharges with ITBs

Science.gov (United States)

Pankin, Alexei; Garofalo, Andrea; Grierson, Brian; Kritz, Arnold; Rafiq, Tariq

2015-11-01

The advanced tokamak scenarios require a large bootstrap current fraction and high β. These large values are often outside the range that occurs in ``conventional'' tokamak discharges. The GLF23, TGLF, and MMM transport models have been previously validated for discharges with parameters associated with ``conventional'' tokamak discharges. It has been demonstrated that the TGLF model under-predicts anomalous transport in high β DIII-D discharges [A.M. Garofalo et al. 2015 TTF Workshop]. In this research, the validity of MMM7.1 model [T. Rafiq et al. Phys. Plasmas 20 032506 (2013)] is tested for high β DIII-D discharges with low and high torque. In addition, the sensitivity of the anomalous transport to β is examined. It is shown that the MMM7.1 model over-predicts the anomalous transport in the DIII-D discharge 154406. In particular, a significant level of anomalous transport is found just outside the internal transport barrier. Differences in the anomalous transport predicted using TGLF and MMM7.1 are reviewed. Mechanisms for quenching of anomalous transport in the ITB regions of high-beta discharges are investigated. This research is supported by US Department of Energy.

Improvement of the tokamak concept

Energy Technology Data Exchange (ETDEWEB)

Laurent, L

1994-12-31

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

'Crescent'-shaped tokamak for compact ignition

International Nuclear Information System (INIS)

Yamazaki, K.; Reiersen, W.T.

1985-12-01

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

'Crescent'-shaped tokamak for compact ignition

International Nuclear Information System (INIS)

Yamazaki, K.; Reiersen, W.T.

1986-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

100% noninductive operation at high beta using off-axis ECCD

International Nuclear Information System (INIS)

Murakami, M.; Greenfield, C.M.; Wade, M.R.

2005-01-01

The Advanced Tokamak (AT) program on DIII-D is developing the scientific basis for steady-state, high-performance operation in future devices. The key element of the program is to demonstrate sustainment of 100% noninductive current for several seconds at high beta. Guided by integrated modeling, recent experiments using up to 2.5 MW of off-axis electron cyclotron current drive (ECCD) and up to 15 MW neutral beam injection (NBI) with q 95 ∼ 5 have sustained ∼ 100% of the plasma current noninductively for 1 s at high beta (β ∼ 3.6%, β N ∼ 3.4, above the no-wall limit) with q min ≥ 1.5 and good confinement (H 89 ∼ 2.3). Integrated modeling using both empirical and theory-based models is used to design experiments and to interpret their results. These experiments have achieved the parameters required for the ITER Q=5 steady-state scenario, and the same modeling tools are applied to ITER AT scenario development. (author)

100% NONINDUCTIVE OPERATION AT HIGH BETA USING OFF-AXIS ECCD

International Nuclear Information System (INIS)

MURAKAMI, M.; GREENFIELD, C.M.; WADE, M.R.; LUCE, T.C.; FERRON, J.R.; ST JOHN, H.E.; MAKOWSKI, M.A.; AUSTIN, M.E.; ALLEN, S.L.; BRENNAN, D.P.; BURRELL, K.H.; CASPER, T.A.; DeBOO, J.C.; DOYLE, E.J.; GAROFALO, A.M.; GOHIL, P.; GORELOV, I.A.; GOEBNER, R.J.; HOBIRK, J.; HYATT, A.W.; JAYAKUMAR, R.J.; KAJIWARA, K.; KESSEL, C.E.; KINSEY, J.E.; LA HAYE, R.J.; KIM, J.Y.; LAO, L.L.; LOHR, J.; MENARD, J.E.; PETTY, C.C.; PETRIE, T.W.; PINSKER, R.I.; POLITZER, P.A.; PRATER, R.; RHODES, T.L.; SIPS, A.C.C.; STAEBLER, G.M.; TAYLOR, T.S.; WANG, G.; WEST, W.P.; ZENG, L.

2004-01-01

The Advanced Tokamak (AT) program on DIII-D is developing the scientific basis for steady-state, high-performance operation in future devices. The key element of the program is to demonstrate sustainment of 100% noninductive current for several seconds at high beta. Guided by integrated modeling, recent experiments using up to 2.5 MW of off-axis electron cyclotron current drive (ECCD) and up to 15 MW neutral beam injection (NBI) with q 95 ∼ 5 have sustained ∼ 100% of the plasma current noninductively for 1 s at high beta (β ∼ 3.6%, β N ∼ 3.4, above the no-wall limit) with q min (ge) 1.5 and good confinement (H 89 ∼ 2.3). Integrated modeling using both empirical and theory-based models is used to design experiments and to interpret their results. These experiments have achieved the parameters required for the ITER Q=5 steady-state scenario, and the same modeling tools are applied to ITER AT scenario development

Heating properties of merging/reconnection startup of high-beta ST

International Nuclear Information System (INIS)

Ono, Yasushi

2005-01-01

The high-power reconnection heating of spherical tokamak (ST) has been studied in the TS-3 experiment by use of axial merging of two STs. In this method, the merging/magnetic reconnection transformed a part of magnetic energy of merging STs into their thermal energy within short reconnection time. Our present low-field merging (0.3-0.5kG, R∼0.2m) attained the maximum heating power of 4-10MW and increased the beta-values of STs by factor 2-3 transiently. The ion heating energy was found to increase inversely with the q-value (B t component) of two STs. The most probable cause for this dependence is fast reconnection speed/ outflow speed due to large anomalous resistivity of current sheet in low-q state. After merging startup, decrease in beta was observed especially in low-q STs, indicating that the final beta value of ST increased with the q-value. (author)

Metastable beta limit in DIII-D

International Nuclear Information System (INIS)

La Haye, R.J.; Callen, J.D.; Gianakon, T.A.

1997-06-01

The long-pulse, slowly evolving single-null divertor (SND) discharges in DIII-D with H-mode, ELMs, and sawteeth are found to be limited significantly below (factor of 2) the predicted ideal limit β N = 4l i by the onset of tearing modes. The tearing modes are metastable in that they are explained by the neoclassical bootstrap current (high β θ ) destabilization of a seed island which occurs even if Δ' θ , there is a region of the modified Rutherford equation such that dw/dt > 0 for w larger than a threshold value; the plasma is metastable, awaiting the critical perturbation which is then amplified to the much larger saturated island. Experimental results from a large number of tokamaks indicate that the high beta operational envelope of the tokamak is well defined by ideal magnetohydrodynamic (MHD) theory. The highest beta values achieved have historically been obtained in fairly short pulse discharges, often <1-2 sawteeth periods and < 1-2 energy replacement times. The maximum operational beta in single-null divertor (SND), long-pulse discharges in DIII-D with a cross-sectional shape similar to the proposed ITER tokamak is found to be limited significantly below the threshold for ideal instabilities by the onset of resistive MHD instabilities

HIGH PERFORMANCE ADVANCED TOKAMAK REGIMES FOR NEXT-STEP EXPERIMENTS

International Nuclear Information System (INIS)

GREENFIELD, C.M.; MURAKAMI, M.; FERRON, J.R.; WADE, M.R.; LUCE, T.C.; PETTY, C.C.; MENARD, J.E; PETRIE, T.W.; ALLEN, S.L.; BURRELL, K.H.; CASPER, T.A; DeBOO, J.C.; DOYLE, E.J.; GAROFALO, A.M; GORELOV, Y.A; GROEBNER, R.J.; HOBIRK, J.; HYATT, A.W; JAYAKUMAR, R.J; KESSEL, C.E; LA HAYE, R.J; JACKSON, G.L; LOHR, J.; MAKOWSKI, M.A.; PINSKER, R.I.; POLITZER, P.A.; PRATER, R.; STRAIT, E.J.; TAYLOR, T.S; WEST, W.P.

2003-01-01

OAK-B135 Advanced Tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic (MHD) stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half-radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding (ELMing) H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results of recent experiment and simulation efforts

IR and FIR laser polarimetry as a diagnostic tool in high-. beta. and Tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Pereira, D; Machida, M; Scalabrin, A

1986-03-01

The change of the polarization state of an electromagnetic wave (EMW) propagating across a magnetized plasma may be used to determine plasma parameters. In a plasma machine of the Tokamak type, the Faraday rotation of the EMW allows for the determination of the product of the plasma electronic density by the poloidal magnetic field. A novel optical configuration which permits simultaneous measurements of these two parameters without the use of an auxiliary interferometric set up is proposed. By choosing appropriate laser wave length this method can be used in Tokamaks (lambda >= 1mm) and also in theta-pinch plasmas (lambda approx. 10..mu..m). The application of these results is discussed to plasma machines now in operation in Brazil, like the Tokamak/USP and theta-pinch/UNICAMP, using lasers developed at UNICAMP.

Limiting beta of stellarators with no net current

International Nuclear Information System (INIS)

Strauss, H.R.; Monticello, D.A.

1981-01-01

Using reduced nonlinear MHD equations, we find finite beta, resistive, l = 2 stellarator equilibria with no net current. We then investigate stability to low mode number internal MHD modes, and find beta limits comparable to tokamaks. Low shear equilibria appear to be substantially more stable than high shear

Progress toward fully noninductive, high beta conditions in DIII-D

International Nuclear Information System (INIS)

Murakami, M.; Wade, M.R.; Greenfield, C.M.; Luce, T.C.; Ferron, J.R.; St John, H.E.; DeBoo, J.C.; Osborne, T.H.; Petty, C.C.; Politzer, P.A.; Burrell, K.H.; Gohil, P.; Gorelov, I.A.; Groebner, R.J.; Hyatt, A.W.; Kajiwara, K.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lohr, J.

2006-01-01

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, β T =3.6%, normalized beta, β N =3.5, and confinement factor, H 89 =2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1 s or half of a current relaxation time (τ CR ). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2 s duration (1τ CR ) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have

Transport and stability studies in negative central shear advanced tokamak plasmas

International Nuclear Information System (INIS)

Jayakumar, R.J.

2003-01-01

Achieving high performance for long duration is a key goal of Advanced Tokamak (AT) research around the world. To this end, tokamak experiments are focusing on obtaining (a) a high fraction of well-aligned non-inductive plasma current (b) wide internal transport barriers (ITBs) in the ion and electron transport channels to obtain high temperatures (c) control of resistive wall modes and neoclassical Tearing Modes which limit the achievable beta. A current profile that yields a negative central magnetic shear (NCS) in the core is consistent with the above focus; Negative central shear is conducive for obtaining internal transport barriers, for high degree of bootstrap current alignment and for reaching the second stability region for ideal ballooning modes, while being stable to ideal kink modes at high beta with wall stabilization. Much progress has been made in obtaining AT performance in several tokamaks through an increasing understanding of the stability and transport properties of tokamak plasmas. RF and neutral beam current drive scenarios are routinely developed and implemented in experiments to access new advanced regimes and control plasma profiles. Short duration and sustained Internal Transport Barriers (ITB) have been obtained in the ion and electron channels. The formation of an ITB is attributable to the stabilization of ion and electron temperature gradient (ITG and ETG) and trapped electron modes (TEM), enhancement of E x B flow shear rate and rarefaction of resonant surfaces near the rational q min values. (orig.)

Transport and stability studies in negative central shear advanced tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Jayakumar, R.J. [Lawrence Livermore National Laboratory (United States)

2003-07-01

Achieving high performance for long duration is a key goal of Advanced Tokamak (AT) research around the world. To this end, tokamak experiments are focusing on obtaining (a) a high fraction of well-aligned non-inductive plasma current (b) wide internal transport barriers (ITBs) in the ion and electron transport channels to obtain high temperatures (c) control of resistive wall modes and neoclassical Tearing Modes which limit the achievable beta. A current profile that yields a negative central magnetic shear (NCS) in the core is consistent with the above focus; Negative central shear is conducive for obtaining internal transport barriers, for high degree of bootstrap current alignment and for reaching the second stability region for ideal ballooning modes, while being stable to ideal kink modes at high beta with wall stabilization. Much progress has been made in obtaining AT performance in several tokamaks through an increasing understanding of the stability and transport properties of tokamak plasmas. RF and neutral beam current drive scenarios are routinely developed and implemented in experiments to access new advanced regimes and control plasma profiles. Short duration and sustained Internal Transport Barriers (ITB) have been obtained in the ion and electron channels. The formation of an ITB is attributable to the stabilization of ion and electron temperature gradient (ITG and ETG) and trapped electron modes (TEM), enhancement of E x B flow shear rate and rarefaction of resonant surfaces near the rational q{sub min} values. (orig.)

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

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

Tokamak power systems studies at ANL

International Nuclear Information System (INIS)

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

1986-01-01

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

Synergism between profile and cross section shape optimization for negative central shear advanced tokamaks

International Nuclear Information System (INIS)

Turnbull, A.D.; Taylor, T.S.; Lao, L.L.

1996-01-01

The Advanced Tokamak (AT) concept is aimed at achieving high beta, high confinement, and a well aligned high bootstrap current fraction in a tokamak configuration consistent with steady state operation. The required improvements over the simple O-D scaling laws, normally used to predict standard, pulsed tokamak performance, axe obtained by taking into account the dependence of the stability and confinement on the 2-D equilibrium; the planned TPX experiment was designed to take full advantage of both advanced profiles and advanced cross-section shaping. Systematic stability studies of the promising Negative Central Shear (NCS) configuration have been performed for a wide variety of cross-section shapes and profile variations. The ideal MHD beta limit is found to be strongly dependent on both and, in fact, there is a clear synergistic relationship between the gains in beta from optimizing the profiles and optimizing the shape. Specifically, for a circular cross-section with highly peaked profiles, β is limited to normalized β values of β N = β/(I/aB) ∼ 2% (mT/MA). A small gain in beta can be achieved by broadening the pressure; however, the root-mean-square beta (β*) is slightly reduced. With peaked pressure profiles, a small increase in β N over that in a circular cross-section is also obtained by strong shaping. At fixed q, this translates to a much larger gain in β and β*. With both optimal profiles and strong shaping, however, the gain in all the relevant fusion performance parameters is dramatic; β and β* can be increased a factor 5 for example. Moreover, the bootstrap alignment is improved. For an optimized strongly shaped configuration, confinement, beta values, and bootstrap alignment adequate for a practical AT power plant appear to be realizable. Data from DIII-D supports these predictions and analysis of the DIII-D data will be presented

Stability analysis for the Big Dee upgrade of the Doublet III tokamak

International Nuclear Information System (INIS)

Helton, F.J.; Luxon, J.L.

1987-01-01

Ideal magnetohydrodynamic stability analysis has been carried out for configurations expected in the Big Dee tokamak, an upgrade of the Doublet III tokamak into a non-circular cross-section device which began operation early in 1986. The results of this analysis support theoretical predictions as follows: Since the maximum value of beta stable to ballooning and Mercier modes, which we denote β c , increases with inverse aspect ratio, elongation and triangularity, the Big Dee is particularly suited to obtain high values of β c and there exist high β c Big Dee equilibria for large variations in all relevant plasma parameters. The beta limits for the Big Dee are consistent with established theory as summarized in present scaling laws. High beta Big Dee equilibria are continuously accessible when approached through changes in all relevant input parameters and are structurally stable with respect to variations of input plasma parameters. Big Dee beta limits have a smooth dependence on plasma parameters such as β p and elongation. These calculations indicate that in the actual running of the device the Big Dee high beta equilibria should be smoothly accessible. Theory predicts that the limiting plasma parameters, such as beta, total plasma current and plasma pressure, which can be obtained within the operating limits of the Big Dee are reactor relevant. Thus the Big Dee should be able to use its favourable ideal MHD scaling and controlled plasma shaping to attain reactor relevant parameters in a moderate sized device. (author)

Economic comparison of MHD equilibrium options for advanced steady state tokamak power plants

International Nuclear Information System (INIS)

Ehst, D.A.; Kessel, C.E.; Jardin, S.C.; Krakowski, R.A.; Bathke, C.G.; Mau, T.K.; Najmabadi, F.

1998-01-01

Progress in theory and in tokamak experiments leads to questions of the optimal development path for commercial tokamak power plants. The economic prospects of future designs are compared for several tokamak operating modes: (high poloidal beta) first stability, second stability and reverse shear. Using a simplified economic model and selecting uniform engineering performance parameters, this comparison emphasizes the different physics characteristics - stability and non- inductive current drive - of the various equilibria. The reverse shear mode of operation is shown to offer the lowest cost of electricity for future power plants. (author)

Economic considerations of commercial tokamak options

International Nuclear Information System (INIS)

Dabiri, A.E.

1986-05-01

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

Microinstability theory in tokamaks: a review

International Nuclear Information System (INIS)

Tang, W.M.

1977-06-01

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

Microinstability theory in tokamaks: a review

Energy Technology Data Exchange (ETDEWEB)

Tang, W.M.

1977-06-01

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

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

Ballooning mode second stability region for sequences of tokamak equilibria

International Nuclear Information System (INIS)

Sugiyama, L.; Mark, J.W.K.

A numerical study of several sequences of tokamak equilibria derived from two flux conserving sequences confirms the tendency of high n ideal MHD ballooning modes to stabilize for values of the plasma beta greater than a second critical beta, for sufficiently favorable equilibria. The major stabilizing effect of increasing the inverse rotational transform profile q(Psi) for equilibria with the same flux surface geometry is shown. The unstable region shifts toward larger shear d ln q/d ln γ and the width of the region measured in terms of the poloidal beta or a pressure gradient parameter, for fixed shear, decreases. The smaller aspect ratio sequences are more sensitive to changes in q and have less stringent limits on the attainable value of the plasma beta in the high beta stable region. Finally, the disconnected mode approximation is shown to provide a reasonable description of the second high beta stability boundary

Implications of rf current drive theory for next step steady-state tokamak design

International Nuclear Information System (INIS)

Schultz, J.H.

1985-06-01

Two missions have been identified for a next-step tokamak experiment in the United States. The more ambitious Mission II device would be a superconducting tokamak, capable of doing long-pulse ignition demonstrations, and hopefully capable of also being able to achieve steady-state burn. A few interesting lines of approach have been identified, using a combination of logical design criteria and parametric system scans [SC85]. These include: (1) TIBER: A point-design suggested by Lawrence Livermore, that proposes a machine with the capability of demonstrating ignition, high beta (10%) and high Q (=10), using high frequency, fast-wave current drive. The TIBER topology uses moderate aspect ratio and high triangularity to achieve high beta. (2) JET Scale-up. (3) Magic5: It is argued here that an aspect ratio of 5 is a magic number for a good steady-state current drive experiment. A moderately-sized machine that achieves ignition and is capable of high Q, using either fast wave or slow wave current drive is described. (4) ET-II: The concept of a highly elongated tokamak (ET) was first proposed as a low-cost approach to Mission I, because of the possibility of achieving ohmic ignition with low-stress copper magnets. We propose that its best application is really for commercial tokamaks, using fast-wave current drive, and suggest a Mission II experiment that would be prototypical of such a reactor

Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta

International Nuclear Information System (INIS)

Schlossberg, David J.; Bodner, Grant M.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.

2017-01-01

Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β t ), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β t up to ~100% with a minimum |B| well spanning up to ~50% of the plasma volume.

Super high field ohmically heated tokamak operation

International Nuclear Information System (INIS)

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

1986-01-01

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

Plea for stellarator funding raps tokamaks

International Nuclear Information System (INIS)

Blake, M.

1992-01-01

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

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

High-beta studies with beam-heated, non-circular plasmas in ISX-B

International Nuclear Information System (INIS)

Lazarus, E.A.; Bates, S.C.; Bush, C.E.

1981-01-01

In this paper we describe some preliminary results of high beta studies on ISX-B for mildly D shaped discharges. ISX-B is a modest size tokamak (R 0 = 93 cm, a = 27 cm) equipped with two tangantially-aligned neutral beam injectors giving a total power up to 3 MW. The poloidal coil system allows choice of plasma boundary shapes from circular to elongated (kappa less than or equal to 1.8), with D, elliptical, or inverse D cross sections. The non-circular work discussed here is for kappa approx. = 1.5

Ideal-MHD beta limits: scaling laws and comparison with Doublet III high-beta plasmas

International Nuclear Information System (INIS)

Bernard, L.C.; Bhadra, D.K.; Helton, F.J.; Lao, L.L.; Todd, T.N.

1983-06-01

Doublet III (DIII) recently has achieved a value for #betta#, the ratio of volume averaged plasma to magnetic pressure, of 4.5%. This #betta# value is in the range required for an economically attractive tokamak reactor, and also close to the relevant limit predicted by ideal-MHD theory. It is therefore of great interest to assess the validity of the theory by comparison with experiment and thus to have a basis for the prediction of future reactor performance. A large variety of plasma shapes have been obtained in DIII. These shapes can be divided into two classes: (1) limiter discharges, and (2) diverted discharges, which are of great interest because of their good confinement in the H-mode operation. We derive simple scaling laws from the variation of optimized ideal-MHD beta limits (#betta#/sub c/) with plasma shape parameters. The current profile is optimized for fixed plasma shapes, separately for the high-n (ballooning) and the low-n (kink) modes. Results are presented in the form of suitability normalized curves of #betta# versus poloidal beta, #betta#/sub p/, for both ballooning and kink modes in order to simultaneously compare all the DIII experimental data

Confinement of a self-stabilized tokamak under average magnetic well conditions

International Nuclear Information System (INIS)

Demchenko, V.V.; Fu, G.Y.; Van Dam, J.W.

1987-05-01

It is well known that the average favorable magnetic curvature of a tokamak is stabilizing with respect to pressure-driven magnetohydrodynamic instabilities at low beta and that self-stabilization occurs at finite beta in the so-called second stability regime. Here we self-consistently investigate how these two effects, viz., the mean magnetic well and the self-stabilization, influence the energy confinement time in a tokamak, using the ballooning mode transport model

A stable route to high-{beta}{sub p} plasmas with non-monotonic q-profiles

Energy Technology Data Exchange (ETDEWEB)

Soeldner, F X; Baranov, Y; Bhatnagar, V P; Bickley, A J; Challis, C D; Fischer, B; Gormezano, C; Huysmans, G T.A.; Kerner, W; Rimini, F; Sips, A C.C.; Springmann, R; Taroni, A [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Goedbloed, J P; Holties, H A [Institute for Plasmas Physics, Nieuwegein (Netherlands); Parail, V V; Pereverzev, G V [Kurchatov Institute of Atomic Energy, Moscow (Russian Federation)

1994-07-01

Steady-state operation of tokamak reactors seems feasible in so-called Advanced Scenarios with high bootstrap current in high-beta{sub p} operation. The stabilization of such discharges with noninductive profile control will be attempted on JET in pursuit of previous high bootstrap current studies. Results of modelling studies of full noninductive current drive scenarios in JET and ITER are presented. Fast Waves (FW), Lower Hybrid (LH) Waves and Neutral Beam Injection (NBI) are used for heating and current drive, alternatively or in combination. A stable route to nonmonotonic q-profiles has been found with a specific ramp-up scenario which combines LH-current drive (LHCD) and a fast Ohmic ramp-up. A hollow current profile with deep shear reversal over the whole central region is thereby formed in an early low-beta phase and frozen in by additional heating. (authors). 5 refs., 4 figs.

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

Research using small tokamaks

International Nuclear Information System (INIS)

1991-01-01

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

Stability of high β large aspect ratio tokamaks

International Nuclear Information System (INIS)

Cowley, S.C.

1991-10-01

High β(β much-gt ε/q 2 ) large aspect ratio (ε much-gt 1) tokamak equilibria are shown to be always stable to ideal M.H.D. modes that are localized about a flux surface. Both the ballooning and interchange modes are shown to be stable. This work uses the analytic high β large aspect ratio tokamak equilibria developed by Cowley et.al., which are valid for arbitrary pressure and safety factor profiles. The stability results make no assumption about these profiles or the shape of the boundary. 14 refs., 4 figs

Accessibility of high β tokamak states

International Nuclear Information System (INIS)

Hogan, J.T.

1978-05-01

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

Physics evaluation of compact tokamak ignition experiments

International Nuclear Information System (INIS)

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

1985-01-01

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

High-β steady-state advanced tokamak regimes for ITER and FIRE

International Nuclear Information System (INIS)

Meade, D.M.; Sauthoff, N.R.; Kessel, C.E.; Budny, R.V.; Gorelenkov, N.; Jardin, S.C.; Schmidt, J.A.; Navratil, G.A.; Bialek, J.; Ulrickson, M.A.; Rognlein, T.; Mandrekas, J.

2005-01-01

An attractive tokamak-based fusion power plant will require the development of high-β steady-state advanced tokamak regimes to produce a high-gain burning plasma with a large fraction of self-driven current and high fusion-power density. Both ITER and FIRE are being designed with the objective to address these issues by exploring and understanding burning plasma physics both in the conventional H-mode regime, and in advanced tokamak regimes with β N ∼ 3 - 4, and f bs ∼50-80%. ITER has employed conservative scenarios, as appropriate for its nuclear technology mission, while FIRE has employed more aggressive assumptions aimed at exploring the scenarios envisioned in the ARIES power-plant studies. The main characteristics of the advanced scenarios presently under study for ITER and FIRE are compared with advanced tokamak regimes envisioned for the European Power Plant Conceptual Study (PPCS-C), the US ARIES-RS Power Plant Study and the Japanese Advanced Steady-State Tokamak Reactor (ASSTR). The goal of the present work is to develop advanced tokamak scenarios that would fully exploit the capability of ITER and FIRE. This paper will summarize the status of the work and indicate critical areas where further R and D is needed. (author)

Deuterium-tritium TFTR plasmas in the high poloidal beta regime

International Nuclear Information System (INIS)

Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.

1995-03-01

Deuterium-tritium plasmas with enhanced energy confinement and stability have been produced in the high poloidal beta, advanced tokamak regime in TFTR. Confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.46 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasma,s exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Fusion power exceeding 6.7 MW with a fusion power gain Q DT = 0.22 has been produced with reduced alpha particle first orbit loss provided by the increased l i

High resolution detection and excitation of resonant magnetic perturbations in a wall-stabilized tokamak

Energy Technology Data Exchange (ETDEWEB)

Maurer, David A. [Physics Department, Auburn University, Auburn, Alabama 36849 (United States); Shiraki, Daisuke; Levesque, Jeffrey P.; Bialek, James; Angelini, Sarah; Byrne, Patrick; DeBono, Bryan; Hughes, Paul; Mauel, Michael E.; Navratil, Gerald A.; Peng Qian; Rhodes, Dov; Rath, Nickolaus; Stoafer, Christopher [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

2012-05-15

We report high-resolution detection of the 3D plasma magnetic response of wall-stabilized tokamak discharges in the High Beta Tokamak-Extended Pulse [T. H. Ivers et al., Phys. Plasmas 3, 1926 (1996)] device. A new adjustable conducting wall has been installed on HBT-EP made up of 20 independent, movable, wall segments instrumented with three distinct sets of 40 modular coils that can be independently driven to generate a wide variety of magnetic perturbations. High-resolution detection of the plasma response is made with 216 poloidal and radial magnetic sensors that have been located and calibrated with high-accuracy. Static and dynamic plasma responses to resonant and non-resonant magnetic perturbations are observed through measurement of the step-response following a rapid change in the toroidal phase of the applied perturbations. Biorthogonal decomposition of the full set of magnetic sensors clearly defines the structures of naturally occurring external kinks as being composed of independent m/n = 3/1 and 6/2 modes. Resonant magnetic perturbations were applied to discharges with pre-existing, saturated m/n = 3/1 external kink mode activity. This m/n = 3/1 kink mode was observed to lock to the applied perturbation field. During this kink mode locked period, the plasma resonant response is characterized by a linear, a saturated, and a disruptive plasma regime dependent on the magnitude of the applied field and value of the edge safety factor and plasma rotation.

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

Geodesic acoustic modes excited by finite beta drift waves

DEFF Research Database (Denmark)

Chakrabarti, Nikhil Kumar; Guzdar, P.N.; Kleva, R.G.

2008-01-01

Presented in this paper is a mode-coupling analysis for the nonlinear excitation of the geodesic acoustic modes (GAMs) in tokamak plasmas by finite beta drift waves. The finite beta effects give rise to a strong stabilizing influence on the parametric excitation process. The dominant finite beta...... effect is the combination of the Maxwell stress, which has a tendency to cancel the primary drive from the Reynolds stress, and the finite beta modification of the drift waves. The zonal magnetic field is also excited at the GAM frequency. However, it does not contribute to the overall stability...... of the three-wave process for parameters of relevance to the edge region of tokamaks....

The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1989-10-01

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

Review of the beta situation

International Nuclear Information System (INIS)

Sheffield, J.

1982-01-01

This note lists some of the possible causes of beta limitation in tokamak and discusses what is known and what is involved in investigating them. The motivation for preparing this note is the observed degradation of confinement with increasing beta poloidal β/sub p/ and beam power P/sub b/ in ISX-B

BETA SCALING OF TRANSPORT ON THE DIII-D TOKAMAK: IS TRANSPORTELECTROSTATIC OR ELECTROMAGNETIC?

International Nuclear Information System (INIS)

PETTY, C.C; LUCE, T.C; McDONALD, D.C; MANDREKAS, J; WADE, M.R; CANDY, J; CORDEY, J.G; DROZDOV, V; EVANS, T.E; FERRON, J.R; GROEBNER, R.J; HYATT, A.W; JACKSON, G.L; LA HAYE, R.J; OSBORNE, T.H; WALTZ, R.E.

2003-01-01

Determining the scaling of transport with (β), the ratio of the plasma kinetic pressure to the magnetic pressure, helps to differentiate between various proposed theories of turbulent transport since mechanisms that are primarily electrostatic show little change in transport with increasing β, while primarily electromagnetic mechanisms generally have a strong unfavorable β scaling. Experiments on the DIII-D tokamak have measured the β scaling of heat transport with all of the other dimensionless parameters held constant in high confinement mode (H-mode) plasmas with edge localized modes (ELMs). A four point scan varied β from 30% to 85% of the ideal ballooning stability limit (normalized beta from 1.0 to 2.8) and found no change in the normalized confinement time, i.e., Bτ th ∞ β -0.01 ± 0.09. The measured thermal diffusivities, normalized to the Bohm diffusion coefficient, also did not vary during the β can to within the experimental uncertainties, whereas the normalized helium particle transport decreased with increasing β. The H-mode pedestal β varied in concert with the core β and showed no signs of saturation. This weak, possibly non-existent, β scaling of transport favors primarily electrostatic mechanisms such as E x B transport, and is in marked disagreement with the strong unfavorable β dependence contained in empirical scaling relations derived from multi-machine H-mode confinement databases

Merging startup experiments on the UTST spherical tokamak

International Nuclear Information System (INIS)

Yamada, Takuma; Kamio, Shuji; Imazawa, Ryota

2010-01-01

The University of Tokyo Spherical Tokamak (UTST) was constructed to explore the formation of ultrahigh-beta spherical tokamak (ST) plasmas using double null plasma merging. The main feature of the UTST is that the poloidal field coils are located outside the vacuum vessel to demonstrate startup in a reactor-relevant situation. Initial operations used partially completed power supplies to investigate the appropriate conditions for plasma merging. The plasma current of the merged ST reached 100 kA when the central solenoid coil was used to assist plasma formation. Merging of two ST plasmas through magnetic reconnection was successfully observed using two-dimensional pickup coil arrays, which directly measure the toroidal and axial magnetic fields inside the UTST vacuum vessel. The resistivity of the current sheet was found to be anomalously high during merging. (author)

Pellet-plasma interaction: Local disturbances caused by pellet ablation in tokamaks

International Nuclear Information System (INIS)

Lengyel, L.L.

1989-01-01

The local disturbance amplitudes caused by ablating pellets in tokamaks are computed in the framework of a magnetohydrodynamic model supplemented by the neutral gas plasma shielding ablation model. The model computes, for a given number of pellet particles locally deposited, the time histories of the ablatant cloud parameters, such as cloud radius, cloud length, electron density, temperature and cloud beta, at a succession of magnetic flux surfaces. The cloud radius thus determined may be fed back into the ablation model, thus adjusting the effect of the shielding cloud on the ablation rate. The model is applied to typical plasma parameter ranges of existing and planned tokamaks. The results show that the ablating pellets may cause massive local disturbances in tokamaks, depending upon the number of particles locally deposited. The peaks of these disturbances are of a spike nature, lasting only a few microseconds (Alfven time-scale). The characteristic decay time of the 'quasi-steady' disturbance values that characterize the after-spike period is of the order of several milliseconds (hydrodynamic time-scale). The peak electron density values may be as high as 10 23 to 10 25 m -3 , with the associated beta peaks exceeding unity. The 'quasi-steady' values of the electron density and the ablatant beta may be of the order of 10 22 to 10 24 m -3 and unity, respectively. Furthermore, the results show the strong dependence of the ablation rate on the dynamic characteristics of the ablatant cloud surrounding the pellet. (author). 25 refs, 6 figs, 2 tabs

Physics Issues in the Design of Low Aspect-Ratio, High-Beta, Quasi-Axisymmetric Stellarators

International Nuclear Information System (INIS)

Zarnstorff, M.C.; Berry, L.A.; Boozer, A.; Brooks, A.; Cooper, W.A.

2000-01-01

Compact stellarators have the potential to combine the best features of the stellarator and the advanced tokamak, offering steady state operation without current drive and potentially without disruptions at an aspect ratio similar to tokamaks. A quasi-axisymmetric stellarator is developed that is consistent with the boot-strap current and passively stable to the ballooning, kink, Mercier, vertical, and neoclassical tearing modes at b=4.1 % without need for conducting walls or external feedback. The configuration has good flux surfaces and fast ion confinement. Thermal transport analysis indicates that the confinement should be similar to tokamaks of the same size, allowing access to the b-limit with moderate power. Coils have been designed to reproduce the physics properties. Initial analysis indicates the coils have considerable flexibility to manipulate the configuration properties. Simulations of the current evolution indicate the kink-mode can remain stable during the approach to h igh-beta

Study of a compact reversed shear Tokamak reactor

International Nuclear Information System (INIS)

Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.

1998-01-01

A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)

High beta experiments in CHS

International Nuclear Information System (INIS)

Okamura, S.; Matsuoka, K.; Nishimura, K.

1994-09-01

High beta experiments were performed in the low-aspect-ratio helical device CHS with the volume-averaged equilibrium beta up to 2.1 %. These values (highest for helical systems) are obtained for high density plasmas in low magnetic field heated with two tangential neutral beams. Confinement improvement given by means of turning off gas puffing helped significantly to make high betas. Magnetic fluctuations increased with increasing beta, but finally stopped to increase in the beta range > 1 %. The coherent modes appearing in the magnetic hill region showed strong dependence on the beta values. The dynamic poloidal field control was applied to suppress the outward plasma movement with the plasma pressure. Such an operation gave fixed boundary operations of high beta plasmas in helical systems. (author)

Energy confinement of high-density tokamaks

NARCIS (Netherlands)

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

1977-01-01

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

Report on the high magnetic field tokamak TRIAM-1

Energy Technology Data Exchange (ETDEWEB)

Ito, T; Kawai, Y; Toi, K; Hiraki, N; Nakamure, K [Kyushu Univ., Fukuoke (Japan). Research Inst. for Applied Mechanics

1981-02-01

A high magnetic field tokamak has been constructed at Kyushu University to study the confinement of high magnetic field tokamak plasma and turbulent heating. The tokamak device consists of toroidal field coils, vertical field coils, horizontal field coils, primary windings, a transformer iron core, turbulent heating coils, and a vacuum chamber. For the observation of plasma, plasma monitors, a micro-wave interferometer, a laser scattering system, a neutral particle energy analyzer, a soft X-ray detector, and a visible spectrometer were installed on the vacuum chamber. The experimental results showed that the central electron temperature was about 640 eV, the central ion temperature 280 eV and mean electron density 2.2 x 10/sup 14//cm/sup 3/. It was found that the proportionality law of electron density and confinement time was valid for this small plasma system. By the turbulent heating, the central ion temperature increased from 170 eV to 580 eV.

Relative merits of size, field, and current on ignited tokamak performance

International Nuclear Information System (INIS)

Uckan, N.A.

1988-01-01

A simple global analysis is developed to examine the relative merits of size (L = a or R/sub 0 /), field (B/sub 0 /), and current (I) on ignition regimes of tokamaks under various confinement scaling laws. Scalings of key parameters with L, B/sub 0 /, and I are presented at several operating points, including (a) optimal path to ignition (saddle point), (b) ignition at minimum beta, (c) ignition at 10 keV, and (d) maximum performance at the limits of density and beta. Expressions for the saddle point and the minimum conditions needed for ohmic ignition are derived analytically for any confinement model of the form tau/sub E/ ∼ n/sup x/T/sup y/. For a wide range of confinement models, the ''figure of merit'' parameters and I are found to give a good indication of the relative performance of the devices where q* is the cylindrical safety factor. As an illustration, the results are applied to representative ''CIT'' (as a class of compact, high-field ignition tokamaks) and ''Super-JETs'' [a class of large-size (few x JET), low-field, high-current (≥20-MA) devices.

Workshop on High Power ICH Antenna Designs for High Density Tokamaks

Science.gov (United States)

Aamodt, R. E.

1990-02-01

A workshop in high power ICH antenna designs for high density tokamaks was held to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of RF auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made.

Integrated plasma control for high performance tokamaks

International Nuclear Information System (INIS)

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

2005-01-01

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

Kinetic analysis of MHD ballooning modes in tokamaks

International Nuclear Information System (INIS)

Tang, W.M.; Rewoldt, G.; Cheng, C.Z.; Chance, M.S.

1984-10-01

A comprehensive analysis of the stability properties of the appropriate kinetically generalized form of MHD ballooning modes together with the usual trapped-particle drift modes is presented. The calculations are fully electromagnetic and include the complete dynamics associated with compressional ion acoustic waves. Trapped-particle effects along with all forms of collisionless dissipation are taken into account without approximations. The influence of collisions is estimated with a model Krook operator. Results from the application of this analysis to realistic tokamak operating conditions indicate that unstable short-wavelength modes with significant growth rates can extend from β = 0 to value above the upper ideal-MHD-critical-beta associated with the so-called second stability regime. Since the strength of the relevant modes appears to vary gradually with β, these results support a soft beta limit picture involving a continuous (rather than abrupt or hard) modification of anomalous transport already present in low-β-tokamaks. However, at higher beta the increasing dominance of the electromagnetic component of the perturbations indicated by these calculations could also imply significantly different transport scaling properties

Unified Ideal Stability Limits for Advanced Tokamak and Spherical Torus Plasmas

International Nuclear Information System (INIS)

Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Sabbagh, S.A.; Fredrickson, E.D.; Jardin, S.C.; Maingi, R.; Manickam, J.; Mueller, D.; Ono, M.; Paoletti, F.; Peng, Y.-K.M.; Soukhanovskii, V.; Stutman, D.; Synakowski, E.J.

2003-01-01

Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal nonrotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants

Magnetic analysis of tokamak plasma with approximate MHD equilibrium solution

International Nuclear Information System (INIS)

Moriyama, Shin-ichi; Hiraki, Naoji

1993-01-01

A magnetic analysis method for determining equilibrium configuration parameters (plasma shape, poloidal beta and internal inductance) on a non-circular tokamak is described. The feature is to utilize an approximate MHD equilibrium solution which explicitly relates the configuration parameters with the magnetic fields picked up by magnetic sensors. So this method is suitable for the real-time analysis performed during a tokamak discharge. A least-squares fitting procedure is added to the analytical algorithm in order to reduce the errors in the magnetic analysis. The validity is investigated through the numerical calculation for a tokamak equilibrium model. (author)

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

A measurement of deuterium neutral by the Balmer-series in the STP-2 high beta screw pinch tokamak

International Nuclear Information System (INIS)

Yamaguchi, S.; Hirano, K.

1980-06-01

The Balmer-alpha and beta are measured with a calibrated spectrograph in STP-2 screw pinch tokamak operated under the maximum toroidal field being 9.2 kG, peak plasma current 30 kA and filling pressure 5 mtorr. The electron temperature and density profiles are obtained by ruby laser Thomson scattering. It is shown that electron temperature is about 10 eV and density is of the order of 10 14 /cm 3 . A non-cylindrical symmetric Abel-inversion technique is used to deduce the emission coefficient profiles from that of the line intensity of the Balmer's. In the present parameter range the neutral deuterium density is almost equal to the population density of the ground state, so that it is obtainable from measured intensities of D sub(α) and D sub(β) which give the population densities of the upper levels i = 3 and 4. The Collisional Radiative (CR) model is applied to the rate equations to estimate the ground state population density. It is found that at 4 μsec from the start of the discharge the deuterium neutral density may be approximately 2 x 10 12 /cm 3 at the center of plasma and 2 x 10 14 /cm 3 at the periphery. These values may contain an error of about factor two. Time history of neutral deuterium density is consistent with the increase of plasma density. (author)

Workshop on high power ICH antenna designs for high density tokamaks

International Nuclear Information System (INIS)

Aamodt, R.E.

1990-01-01

A workshop in high power ICH antenna designs for high density tokamaks was held in Boulder, Colorado on January 31 through February 2, 1990. The purposes of the workshop were to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of rf auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made

Increasing the beta limit due to neoclassical tearing modes by raising the axial safety factor q(0) > 1

International Nuclear Information System (INIS)

La Haye, R.J.; Strait, E.J.; Rice, B.W.

2000-01-01

Resistive neoclassical tearing modes (NTMs) can cause the stable beta value in long pulse, high confinement plasmas to fall significantly below that predicted by ideal theory. The NTM islands which degrade confinement and limit beta are induced and sustained by helically perturbed bootstrap currents. A combination of shaping and q-profile modification is used in the DIII-D tokamak to increase this beta limit. (author)

Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime

Science.gov (United States)

Ren, Q.

2015-11-01

Recent DIII-D experiments have advanced the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Fully noninductive plasmas with extremely high values of the poloidal beta, βp >= 4 , have been sustained at βT >= 2 % for long durations with excellent energy confinement quality (H98y,2 >= 1 . 5) and internal transport barriers (ITBs) generated at large minor radius (>= 0 . 6) in all channels (Te, Ti, ne, VTf). Large bootstrap fraction (fBS ~ 80 %) has been obtained with high βp. ITBs have been shown to be compatible with steady state operation. Because of the unusually large ITB radius, normalized pressure is not limited to low βN values by internal ITB-driven modes. βN up to ~4.3 has been obtained by optimizing the plasma-wall distance. The scenario is robust against several variations, including replacing some on-axis with off-axis neutral beam injection (NBI), adding electron cyclotron (EC) heating, and reducing the NBI torque by a factor of 2. This latter observation is particularly promising for extension of the scenario to EAST, where maximum power is obtained with balanced NBI injection, and to a reactor, expected to have low rotation. However, modeling of this regime has provided new challenges to state-of-the-art modeling capabilities: quasilinear models can dramatically underpredict the electron transport, and the Sauter bootstrap current can be insufficient. The analysis shows first-principle NEO is in good agreement with experiments for the bootstrap current calculation and ETG modes with a larger saturated amplitude or EM modes may provide the missing electron transport. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA

Physics of energetic particle-driven instabilities in the START spherical tokamak

International Nuclear Information System (INIS)

McClements, K.G.; Gryaznevich, M.P.; Akers, R.J.; Appel, L.C.; Counsell, G.F.; Roach, C.M.; Sharapov, S.E.; Majeski, R.

1999-01-01

The recent use of neutral beam injection (NBI) in the UKAEA small tight aspect ratio tokamak (START) has provided the first opportunity to study experimentally the physics of energetic ions in spherical tokamak (ST) plasmas. In such devices the ratio of major radius to minor radius R 0 /a is of order unity. Several distinct classes of NBI-driven instability have been observed at frequencies up to 1 MHz during START discharges. These observations are described, and possible interpretations are given. Equilibrium data, corresponding to times of beam-driven wave activity, are used to compute continuous shear Alfven spectra: toroidicity and high plasma beta give rise to wide spectral gaps, extending up to frequencies of several times the Alfven gap frequency. In each of these gaps Alfvenic instabilities could, in principle, be driven by energetic ions. Chirping modes observed at high beta in this frequency range have bandwidths comparable to or greater than the gap widths. Instability drive in START is provided by beam ion pressure gradients (as in conventional tokamaks), and also by positive gradients in beam ion velocity distributions, which arise from velocity-dependent charge exchange losses. It is shown that fishbone-like bursts observed at a few tens of kHz can be attributed to internal kink mode excitation by passing beam ions, while narrow-band emission at several hundred kHz may be due to excitation of fast Alfven (magnetosonic) eigenmodes. In the light of our understanding of energetic particle-driven instabilities in START, the possible existence of such instabilities in larger STs is discussed. (author)

Topics in stability and transport in tokamaks: Dynamic transition to second stability with auxiliary heating; stability of global Alfven waves in an ignited plasma

International Nuclear Information System (INIS)

Fu, G.

1988-01-01

The problem of access to the high-beta ballooning second-stability regime by means of auxiliary heating and the problem of the stability of global-shear Alfven waves in an ignited tokamak plasma are theoretically investigated. These two problems are related to the confinement of both the bulk plasma as well as the fusion-product alpha particles an dare fundamentally important to the operation of ignited tokamak plasma. First, a model that incorporates both transport and ballooning mode stability was developed in order to estimate the auxiliary heating power required for tokamak plasma to evolve in time self-consistently into a high-beta, globally self-stabilized equilibrium. The critical heating power needed for access to second stability is found to be proportional to the square root of the anomalous diffusivity induced by the ballooning instability. Next, the full effects of toroidicity are retained in a theoretical description of global-type-shear Alfven modes whose stability can be modified by the fusion-product alpha particles that will present in an ignited tokamak plasma. Toroidicity is found to induce mode coupling and to stabilize the so-called Global Alfven Eigenmodes (GAE)

Stability of Tokamaks with respect to slip motions

International Nuclear Information System (INIS)

Rebhan, E.; Salat, A.

1976-06-01

Using the energy principle in Tokamaks we investigate a class of perturbations which, if unstable, cannot be stabilized by the toroidal main field. On the assumptions of usual Tokamak ordering and in the limit of infinite aspect ratio, these perturbations are shown to be minimizing among all axisymmetric perturbations. In the case of finite aspect ratio, a detailed stability analysis is carried out using a constant pressure surface current model with elliptic, triangular or rectangular plasma cross-section. Definite stabilization by toroidal effects and by beta poloidal is demonstrated. (orig.) [de

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

Alpha-particle effects on high-n instabilities in tokamaks

International Nuclear Information System (INIS)

Rewoldt, G.

1988-06-01

Hot α-particles and thermalized helium ash particles in tokamaks can have significant effects on high toroidal mode number instabilities such as the trapped-electron drift mode and the kinetically calculated magnetohydrodynamic ballooning mode. In particular, the effects can be stabilizing, destabilizing, or negligible, depending on the parameters involved. In high-temperature tokamaks capable of producing significant numbers of hot α-particles, the predominant interaction of the mode with the α-particles is through resonances of various sorts. In turn, the modes can cause significant anomalous transport of the α-particles and the helium ash. Here, results of comprehensive linear eigenfrequency-eigenfunction calculations are presented for relevant realistic cases to show these effects. 24 refs., 12 figs., 6 tabs

Collective processes in a tokamak with high-energy particles: general problems of the linear theory of Alfven instabilities of a tokamak with high-energy ions

International Nuclear Information System (INIS)

Mikhailovskii, A.B.

1986-01-01

Some general problems of the theory of Alfven instabilities of a tokamak with high-energy ions are considered. It is assumed that such ions are due to either ionization of fast neutral atoms, injected into the tokamak, or production of them under thermo-nuclear conditions. Small-oscillation equations are derived for the Alfven-type waves, which allow for both destabilizing effects, associated with the high-energy particles, and stabilizing ones, such as effects of shear and bulk-plasm dissipation. A high-energy ion contribution is calculated into the growth rate of the Alfven waves. The author considers the role of trapped-electron collisional dissipation

Impurity screening in high density plasmas in tokamaks with a limiter configuration

International Nuclear Information System (INIS)

Ferro, C.; Zanino, R.

1992-01-01

Impurity screening in high density plasmas in tokamaks with a limiter configuration is investigated by means of a simple semi-analytical model. An iterative scheme is devised, in order to determine self-consistently the values of scrape-off layer thickness, edge electron density and temperature, and main plasma contamination parameter Z eff , as a function of given average electron density and temperature in the main plasma and given input power. The model is applied to the poloidal limiter case of the Frascati Tokamak Upgrade, and results are compared with experimental data. A reasonable agreement between the trends is found, emphasizing the importance of a high edge plasma density for obtaining a clean main plasma in limiter tokamaks. (orig.)

First Results from Tests of High Temperature Superconductor Magnets on Tokamak

Energy Technology Data Exchange (ETDEWEB)

Gryaznevich, M.; Todd, T.T., E-mail: mikhail.gryaznevich@ccfe.ac.uk [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Svoboda, V.; Markovic, T.; Ondrej, G. [Czech Technical University, Prague (Czech Republic); Stockel, J.; Duran, I.; Kovarik, K. [IPP Prague, Czech Technical University, Prague (Czech Republic); Sykes, A.; Kingham, D. [Tokamak Solutions, Culham Science Centre, Abingdon (United Kingdom); Melhem, Z.; Ball, S.; Chappell, S. [Oxford Instruments, Abingdon (United Kingdom); Lilley, M. K.; De Grouchy, P.; Kim, H. -T. [Imperial College, London (United Kingdom)

2012-09-15

Full text: It has long been known that high temperature superconductors (HTS) could have an important role to play in the future of tokamak fusion research. Here we report on first results of the use of HTS in a tokamak magnet and on the progress in design and construction of the first fully-HTS tokamak. In the experiment, the two copper vertical field coils of the small tokamak GOLEM were replaced by two coils each with 6 turns of HTS (Re)BCO tape. Liquid nitrogen was used to cool the coils to below the critical temperature at which HTS becomes superconducting. Little effect on the HTS critical current has been observed for perpendicular field up to 0.5 T and superconductivity has been achieved at {approx} 90.5K during bench tests. There had been concerns that the plasma pulses and pulsed magnetic fields might cause a 'quench' in the HTS, i.e., a sudden and potentially damaging transition from superconductor to normal conductor. However, many plasma pulses were fired without any quenches even when disruptions occurred with corresponding induced electrical fields. In addition, experiments without plasma have been performed to study properties of the HTS in a tokamak environment, i.e., critical current and its dependence on magnetic and electrical fields generated in a tokamak both in DC and pulsed operations, maximum current ramp-up speed, performance of the HTS tape after number of artificially induced quenches etc. No quench has been observed at DC currents up to 200 A (1.2 kA-turns through the coil). In short pulses, current up to 1 kA through the tape (6 kA-turns) has been achieved with no subsequent degradation of the HTS performance with a current ramp rate up to 0.6 MA/s. In future experiments, increases in both the plasma current and pulse duration are planned. Considerable experience has been gained during design and fabrication of the cryostat, coils, isolation and insulation, feeds and cryosystems, and GOLEM is now routinely operated with HTS coils. The

Experimental study on practicability of self-created spherical tokamak in coilless STPC-EX machine

International Nuclear Information System (INIS)

Sinman, S.

2002-01-01

The aim of this study is to recognize the physical basis of the alternative self organization mechanism occurred STPC-EX machine. The conventional diagnostic tools are used in this study and for photographic recording, open shutter integrated post-fogging method is preferred. The annular coaxial two plasma current sheets one within other at the same direction are created and flowed on the surface of floating conductive central rod. Consequently, spherical tokamak configurated by new creation mechanism of Dual Axial Z-Pinch. (DAZP) yields fairly high beta of 0.4-0.6 at self created spherical tokamak plasma. Sustainment time of DAZP is 5.6-6.3 mili second. (author)

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

Stability at high performance in the MAST spherical tokamak

International Nuclear Information System (INIS)

Buttery, R.J.; Akers, R.; Arends, E. =

2003-01-01

The development of reliable H-modes on MAST, together with advances in heating power and a range of powerful diagnostics, has provided a platform to enable MAST to address some of he most important issues of tokamak stability. In particular the high β potential of the ST is highlighted with stable operation at β N ∼5-6 , β T ∼ 16% and β p as high as 1.9, confirmed by a range of profile diagnostics. Calculations indicate that β N levels are in the vicinity of no-wall stability limits. Studies have provided the first identification of the Neoclassical Tearing Mode (NTM) in the ST, using its behaviour to quantitatively validate predictions of NTM theory, previously only applied to conventional tokamaks. Experiments have demonstrated that sawteeth play a strong role in triggering NTMs - by avoiding large sawteeth much higher β N can, and has, been reached. Further studies have confirmed the NTM's significance, with large islands observed using the 300 point Thomson diagnostic, and locking of large n=1 modes frequently leading to disruptions. H-mode plasmas are also limited by ELMs, with confinement degraded as ELM frequency rises. However, unlike the conventional tokamak, the ELMs in high performing regimes on MAST (H IPB98Y2 ∼1) appear to be type III in nature. Modelling identifies instability to peeling modes, consistent with a type III interpretation, and shows considerable scope to raise pressure gradients (despite n=∞ ballooning theory predictions of instability) before ballooning type modes (perhaps associated with type I ELMs) occur. Finally sawteeth are shown not to remove the q=1 surface in the ST - other promising models are being explored. Thus research on MAST is not only demonstrating stable operation at high performance levels, and developing methods to control instabilities; it is also providing detailed tests of the stability physics and models applicable to conventional tokamaks, such as ITER. (author)

Neutral-beam deposition in large, finite-beta noncircular tokamak plasmas

International Nuclear Information System (INIS)

Wieland, R.M.; Houlberg, W.A.

1982-02-01

A parametric pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross section and noncircular shifted plasma cross sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is benchmarked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radius a = 120 cm and elongation b/a = 1.6. Deposition profiles are compared for deuterium beam energies of 120 to 150 keV, central plasma densities of 8 x 10 13 - 2 x 10 14 cm -3 , and beam orientation ranging from perpendicular to tangential to the inside wall

On the minimum circulating power of steady state tokamaks

International Nuclear Information System (INIS)

Itoh, K.; Itoh, S.; Fukuyama, A.; Yagi, M.

1995-07-01

Circulating power for the sustenance and profile control of the steady state tokamak plasmas is discussed. The simultaneous fulfillment of the MHD stability at high beta value, the improved confinement and the stationary equilibrium requires the rotation drive as well as the current drive. In addition to the current drive efficiency, the efficiency for the rotation drive is investigated. The direct rotation drive by the external torque, such as the case of beam injection, is not efficient enough. The mechanism and the magnitude of the spontaneous plasma rotation are studied. (author)

Study on fast ion loss in HL-2A tokamak

International Nuclear Information System (INIS)

Liu Yi; Sun Tengfei; Ji Xiaoquan

2012-01-01

Experiments with a high-energy deuterium neutral beam (NB) injection (30 keV, about 0.6 MW) were performed on the HL-2A tokamak. Analysis of neutron decay following the NB 'blip' injection indicates that tangentially injected beam ions are well confined, slowing down classically in the HL-2A. Anomalous losses of beam ions were observed when a beta-induced Alfven acoustic (BAAE) mode was present in the plasma. Such a high energetic particle driven mode led to fast-ion loss, showing a strong influence of the energetic particle driven mode on the fast-ion transport. (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

Physics and Control of Locked Modes in the DIII-D Tokamak

International Nuclear Information System (INIS)

Volpe, Francesco

2017-01-01

This Final Technical Report summarizes an investigation, carried out under the auspices of the DOE Early Career Award, of the physics and control of non-rotating magnetic islands (''locked modes'') in tokamak plasmas. Locked modes are one of the main causes of disruptions in present tokamaks, and could be an even bigger concern in ITER, due to its relatively high beta (favoring the formation of Neoclassical Tearing Mode islands) and low rotation (favoring locking). For these reasons, this research had the goal of studying and learning how to control locked modes in the DIII-D National Fusion Facility under ITER-relevant conditions of high pressure and low rotation. Major results included: the first full suppression of locked modes and avoidance of the associated disruptions; the demonstration of error field detection from the interaction between locked modes, applied rotating fields and intrinsic errors; the analysis of a vast database of disruptive locked modes, which led to criteria for disruption prediction and avoidance.

Physics and Control of Locked Modes in the DIII-D Tokamak

Energy Technology Data Exchange (ETDEWEB)

Volpe, Francesco [Columbia Univ., New York, NY (United States). Dept. of Applied Physics and Applied Mathematics

2017-01-30

This Final Technical Report summarizes an investigation, carried out under the auspices of the DOE Early Career Award, of the physics and control of non-rotating magnetic islands (“locked modes”) in tokamak plasmas. Locked modes are one of the main causes of disruptions in present tokamaks, and could be an even bigger concern in ITER, due to its relatively high beta (favoring the formation of Neoclassical Tearing Mode islands) and low rotation (favoring locking). For these reasons, this research had the goal of studying and learning how to control locked modes in the DIII-D National Fusion Facility under ITER-relevant conditions of high pressure and low rotation. Major results included: the first full suppression of locked modes and avoidance of the associated disruptions; the demonstration of error field detection from the interaction between locked modes, applied rotating fields and intrinsic errors; the analysis of a vast database of disruptive locked modes, which led to criteria for disruption prediction and avoidance.

Non-Inductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta in the Pegasus Toroidal Experiment

Science.gov (United States)

Reusch, Joshua

2017-10-01

A major goal of the spherical tokamak research program is accessing a state of low internal inductance li, high elongation κ, high toroidal and normalized beta (βt and βN) , and low collisionality without solenoidal current drive. A new local helicity injection (LHI) system in the lower divertor region of the ultra-low aspect ratio Pegasus ST provides non-solenoidally driven plasmas that exhibit most of these characteristics. LHI utilizes compact, edge-localized current sources (Ainj 4 cm2, Iinj 8 kA, Vinj 1.5 kV) for plasma startup and sustainment, and can sustain more than 200 kA of plasma current. Plasma growth via LHI is enhanced by a transition from a regime of high kink-like MHD activity to one of reduced MHD activity at higher frequencies and presumably shorter wavelengths. The strong edge current drive provided by LHI results in a hollow current density profile with low li. The low aspect ratio (R0 / a 1.2) of Pegasus allows ready access to high κ and MHD stable operation at very high normalized plasma currents (IN =Ip /aBT> 15). Thomson scattering measurements indicate Te 100 eV and ne 1 ×19 m-3. The impurity Ti evolution is correlated in time with high frequency magnetic fluctuations, implying substantial reconnection ion heating is driven by the applied helicity injection. Doppler spectroscopy indicates Ti >=Te and that the anomalous ion heating scales consistently with two fluid reconnection theory. Taken together, these features provide access to very high βt plasmas. Equilibrium analyses indicate βt up to 100% and βN 6.5 is achieved. At increasingly low BT, the discharge disrupts at the no-wall ideal stability limit. In these high βt discharges, a minimum |B| well forms over 50% of the plasma volume. This unique magnetic configuration may be of interest for testing predictions of stabilizing drift wave turbulence and/or improving energetic particle confinement. This work supported by US DOE Grants DE-FG02-96ER54375 and DE-SC0006928.

Tokamak power system studies at ANL

International Nuclear Information System (INIS)

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

1986-06-01

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

On steady poloidal and toroidal flows in tokamak plasmas

International Nuclear Information System (INIS)

McClements, K. G.; Hole, M. J.

2010-01-01

The effects of poloidal and toroidal flows on tokamak plasma equilibria are examined in the magnetohydrodynamic limit. ''Transonic'' poloidal flows of the order of the sound speed multiplied by the ratio of poloidal magnetic field to total field B θ /B can cause the (normally elliptic) Grad-Shafranov (GS) equation to become hyperbolic in part of the solution domain. It is pointed out that the range of poloidal flows for which the GS equation is hyperbolic increases with plasma beta and B θ /B, thereby complicating the problem of determining spherical tokamak plasma equilibria with transonic poloidal flows. It is demonstrated that the calculation of the hyperbolicity criterion can be easily modified when the assumption of isentropic flux surfaces is replaced with the more tokamak-relevant one of isothermal flux surfaces. On the basis of the latter assumption, a simple expression is obtained for the variation of density on a flux surface when poloidal and toroidal flows are simultaneously present. Combined with Thomson scattering measurements of density and temperature, this expression could be used to infer information on poloidal and toroidal flows on the high field side of a tokamak plasma, where direct measurements of flows are not generally possible. It is demonstrated that there are four possible solutions of the Bernoulli relation for the plasma density when the flux surfaces are assumed to be isothermal, corresponding to four distinct poloidal flow regimes. Finally, observations and first principles-based theoretical modeling of poloidal flows in tokamak plasmas are briefly reviewed and it is concluded that there is no clear evidence for the occurrence of supersonic poloidal flows.

High beta and second stability region transport and stability analysis

International Nuclear Information System (INIS)

Hughes, M.H.; Phillps, M.W.; Todd, A.M.M.; Krishnaswami, J.; Hartley, R.

1992-09-01

This report describes ideal and resistive studies of high-beta plasmas and of the second stability region. Emphasis is focused on ''supershot'' plasmas in TFIR where MHD instabilities are frequently observed and which spoil their confinement properties. Substantial results are described from the analysis of these high beta poloidal plasmas. During these studies, initial pressure and safety factor profiles were obtained from the TRANSP code, which is used extensively to analyze experimental data. Resistive MBD stability studies of supershot equilibria show that finite pressure stabilization of tearing modes is very strong in these high βp plasmas. This has prompted a detailed re-examination of linear tearing mode theory in which we participated in collaboration with Columbia University and General Atomics. This finite pressure effect is shown to be highly sensitive to small scale details of the pressure profile. Even when an ad hoc method of removing this stabilizing mechanism is implemented, however, it is shown that there is only superficial agreement between resistive MBD stability computation and the experimental data. While the mode structures observed experimentally can be found computationally, there is no convincing correlation with the experimental observations when the computed results are compared with a large set of supershot data. We also describe both the ideal and resistive stability properties of TFIR equilibria near the transition to the second region. It is shown that the highest β plasmas, although stable to infinite-n ideal ballooning modes, can be unstable to the so called ''infernal'' modes associated with small shear. The sensitivity of these results to the assumed pressure and current density profiles is discussed. Finally, we describe results from two collaborative studies with PPPL. The first involves exploratory studies of the role of the 1/1 mode in tokamaks and, secondly, a study of sawtooth stabilization using ICRF

Shear Alfven waves in tokamaks

International Nuclear Information System (INIS)

Kieras, C.E.

1982-12-01

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

Experimental study of the β-limit in ohmic H-mode in the TUMAN-3M tokamak

International Nuclear Information System (INIS)

Lebedev, S.V.; Andreiko, M.V.; Askinazi, L.G.; Golant, V.E.; Kornev, V.A.; Krikunov, S.V.; Levin, L.S.; Rozhdestvensky, V.V.; Tukachinsky, A.S.; Yaroshevich, S.P.

1998-01-01

Because of its high confinement properties, the H-mode provides good opportunities to achieve high beta values in a tokamak. In this paper the results of an experimental study of β T and β N limits in the H-mode, obtained in a circular cross section tokamak without auxiliary heating are presented. The experiments were performed in the TUMAN-3M tokamak. The device has the following parameters: R 0 =0.53m, a s =0.22m (limiter configuration), B T ≤1.2T, I p ≤175kA, n-bar e ≤6.2x10 19 m -3 . The stored energy was measured using diamagnetic loops and compared with W calculated from kinetic data obtained by Thomson scattering and microwave interferometry. Measurements of the stored energy and of the β were performed in the ohmic H-mode before and after boronization and in the scenario with fast current ramp-down in ohmic H-mode. A maximum value of β T of 2.0% and β N of 2.0 were achieved. The β N limit achieved reveals itself as a 'soft' (non-disruptive) limit. The stored energy slowly decays after the current ramp-down. No correlation was found between beta restriction and MHD phenomena. Internal transport barrier (ITB) formation was observed in ohmic H-mode. An enhancement factor of 2.0 over ITER93H(ELM-free) was found in the ohmic H-mode with ITB. (author)

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

International Nuclear Information System (INIS)

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

1986-01-01

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

About the Toroidal Magnetic Field of a Tokamak Burning Plasma Experiment with Superconducting Coils

International Nuclear Information System (INIS)

Mazzucato, E.

2002-01-01

In tokamaks, the strong dependence on the toroidal magnetic field of both plasma pressure and energy confinement is what makes possible the construction of small and relatively inexpensive burning plasma experiments using high-field resistive coils. On the other hand, the toroidal magnetic field of tokamaks using superconducting coils is limited by the critical field of superconductivity. In this article, we examine the relative merit of raising the magnetic field of a tokamak plasma by increasing its aspect ratio at a constant value of the peak field in the toroidal magnet. Taking ITER-FEAT as an example, we find that it is possible to reach thermonuclear ignition using an aspect ratio of approximately 4.5 and a toroidal magnetic field of 7.3 T. Under these conditions, fusion power density and neutron wall loading are the same as in ITER [International Thermonuclear Experimental Reactor], but the normalized plasma beta is substantially smaller. Furthermore, such a tokamak would be able to reach an energy gain of approximately 15 even with the deterioration in plasma confinement that is known to occur near the density limit where ITER is forced to operate

ICRF Mode Conversion Current Drive for Plasma Stability Control in Tokamaks

International Nuclear Information System (INIS)

Grekov, D.; Kock, R.; Lyssoivan, A.; Noterdaeme, J. M.; Ongena, J.

2007-01-01

There is a substantial incentive for the International Thermonuclear Experimental Reactor (ITER) to operate at the highest attainable beta (plasma pressure normalized to magnetic pressure), a point emphasized by requirements of attractive economics in a reactor. Recent experiments aiming at stationary high beta discharges in tokamak plasmas have shown that maximum achievable beta value is often limited by the onset of instabilities at rational magnetic surfaces (neoclassical tearing modes). So, methods of effective control of these instabilities have to be developed. One possible way for neoclassical tearing modes control is an external current drive in the island to locally replace the missing bootstrap current and thus to suppress the instability. Also, a significant control of the sawtooth behaviour was demonstrated when the magnetic shear was modified by driven current at the magnetic surface where safety factor equals to 1. In the ion cyclotron range of frequencies (ICRF), the mode conversion regime can be used to drive the local external current near the position of the fast-to-slow wave conversion layer, thus providing an efficient means of plasma stability control. The slow wave energy is effectively absorbed in the vicinity of mode conversion layer by electrons with such parallel to confining magnetic field velocities that the Landau resonance condition is satisfied. For parameters of present day tokamaks and for ITER parameters the slow wave phase velocity is rather low, so the large ratio of momentum to energy content would yield high current drive efficiency. In order to achieve noticeable current drive effect, it is necessary to create asymmetry in the ICRF power absorption between top and bottom parts of the plasma minor cross-section. Such asymmetric electron heating may be realized using: - shifted from the torus midplane ICRF antenna in TEXTOR tokamak; - plasma displacement in vertical direction that is feasible in ASDEX-Upgrade; - the

Methods for the design and optimization of shaped tokamaks

International Nuclear Information System (INIS)

Haney, S.W.

1988-05-01

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

A minimum-size tokamak concept for conditions near ignition

International Nuclear Information System (INIS)

Lehnert, B.

1983-01-01

Based on a combination of Alcator scaling and a recent theory on the Murakami density limit, a minimum-size tokamak concept (Minitor) is proposed. Even if this concept does not aim at alpha particle containment, it has the important goal of reaching plasma core temperatures and Lawson parameter values required for ignition, by ohmic heating alone and under macroscopically stable conditions. The minimized size, and the associated enhancement of the plasma current density, are found to favour high plasma temperatues, average densities, and beta values. The goal of this concept appears to be realizable by relatively modest technical means. (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.; 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

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

High n ballooning modes in highly elongated tokamaks

International Nuclear Information System (INIS)

An, C.H.; Bateman, G.

1980-02-01

An analytic study of stability against high n ballooning modes in highly elongated axisymmetric plasmas is presented and compared with computational results. From the equation for the marginal pressure gradient, it is found that the local shear plays an important role on the stability of elongated and shifted plasma, and that high elongation deteriorates the stability by decreasing the stabilizing effects of field line bending and local shear. The net contribution of the local shear to stability decreases with elongation and shift for strongly ballooning modes (eigenfunctions strongly localized near the outer edge of the toroidal flux surfaces) but increases for interchange modes (eigenfunctions more uniform along the flux surfaces). The computational study of high n ballooning modes in a highly elongated plasma reveals that lowering the aspect ratio and broadening the pressure profile enhance the marginal beta for β/sub p/ less than unity but severely reduce the marginal beta for β/sub p/ larger than unity

High gain requirements and high field Tokamak experiments

International Nuclear Information System (INIS)

Cohn, D.R.

1994-01-01

Operation at sufficiently high gain (ratio of fusion power to external heating power) is a fundamental requirement for tokamak power reactors. For typical reactor concepts, the gain is greater than 25. Self-heating from alpha particles in deuterium-tritium plasmas can greatly reduce ητ/temperature requirements for high gain. A range of high gain operating conditions is possible with different values of alpha-particle efficiency (fraction of alpha-particle power that actually heats the plasma) and with different ratios of self heating to external heating. At one extreme, there is ignited operation, where all of the required plasma heating is provided by alpha particles and the alpha-particle efficiency is 100%. At the other extreme, there is the case of no heating contribution from alpha particles. ητ/temperature requirements for high gain are determined as a function of alpha-particle heating efficiency. Possibilities for high gain experiments in deuterium-tritium, deuterium, and hydrogen plasmas are discussed

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-03-01

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

ADX: a high field, high power density, advanced divertor and RF tokamak

Science.gov (United States)

LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

2015-05-01

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

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.

Development of high thermal flux components for continuous operation in Tokamaks

International Nuclear Information System (INIS)

Schlosser, J.; Chappuis, P.; Coston, J.F.; Deschamps, P.; Lipa, M.

1991-01-01

High heat flux plasma facing components are under development and appropriate experimental evaluations have been carried out in order to operate during cycles of several hundred seconds. In Tore Supra, a large tokamak with a plasma nominal duration in excess of 30 seconds, solutions are tested that could be later applied to the NET/ITER tokamak, where peaked heat flux values of 15 MW/m 2 on the divertor plates are foreseen. The proposed concept is a swirl square tube design protected with brazed CFC flat tiles. Development programs and validation tests are presented. The tests results are compared with calculations

Control, pressure perturbations, displacements, and disruptions in highly elongated tokamak plasmas

International Nuclear Information System (INIS)

Marcus, F.B.; Hofmann, F.; Tonetti, G.; Jardin, S.C.; Noll, P.

1989-06-01

The control and evolution of highly elongated tokamak plasmas with large growth rates are simulated with the axisymmetric, resistive MHD code TSC in the geometry of the TCV tokamak. Pressure perturbations such as sawteeth and externally programmed displacements create initial velocity perturbations which may be stabilized by low power, rapid response coils inside the passively stabilizing vacuum vessel, together with slower shaping coils outside the vessel. Vertical disruption induced voltages and forces on the rapid coils and vessel are investigated, and a model is proposed for an additional vertical force due to poloidal currents. (author) 6 figs., 1 tab., 26 refs

MHD stability limits in the TCV Tokamak

International Nuclear Information System (INIS)

Reimerdes, H.

2001-07-01

of this limit with elongation is also in qualitative agreement with ideal MHD theory. Edge localised modes (ELMs), occurring in TCV Ohmic high-confinement mode discharges, were observed to be preceded by coherent magnetic oscillations. The detected poloidal and toroidal mode structures are consistent with a resonant flux surface close to the plasma edge. Unlike conventional MHD modes, these precursors start at a random toroidal location and then grow in amplitude and toroidal extent until they encompass the whole toroidal circumference. Thus, the asymmetry causing and maintaining the toroidal localisation of the ELM precursor must be intrinsic to the plasma. Soft X-ray measurements show that the localised precursor always coincides with a central m = 1 mode, which can usually be associated with the sawtooth pre- or postcursor mode. A comparison of the phases indicates a correlation with the maximum of the central mode preceding the toroidal location of the ELM precursor and, therefore, a hitherto unobserved coupling between central modes and ELMs. Highly elongated plasmas promise several advantages, among them higher current and beta limits. During TCV experiments dedicated to an increasing of the plasma elongation, a new disruptive current limit, at values well below the conventional current limit corresponding to q a > 2, was encountered for κ > 2.3. This limit, which is preceded by a kink-type mode, is found to be consistent with ideal MHD stability calculations. The TCV observations, therefore, provide the first experimental confirmation of a deviation of the linear Troyon-scaling of the ideal beta limit with normalised current at high elongation, which was predicted over 10 years ago. Neoclassical tearing modes (NTMs), which have been observed to limit the achievable beta in a number of tokamaks, arise from a helical perturbation of the bootstrap current caused by an existing seed island. Neoclassical m/n = 2/1 tearing modes have been identified in TCV

MHD stability limits in the TCV Tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

observed decrease of this limit with elongation is also in qualitative agreement with ideal MHD theory. Edge localised modes (ELMs), occurring in TCV Ohmic high-confinement mode discharges, were observed to be preceded by coherent magnetic oscillations. The detected poloidal and toroidal mode structures are consistent with a resonant flux surface close to the plasma edge. Unlike conventional MHD modes, these precursors start at a random toroidal location and then grow in amplitude and toroidal extent until they encompass the whole toroidal circumference. Thus, the asymmetry causing and maintaining the toroidal localisation of the ELM precursor must be intrinsic to the plasma. Soft X-ray measurements show that the localised precursor always coincides with a central m = 1 mode, which can usually be associated with the sawtooth pre- or postcursor mode. A comparison of the phases indicates a correlation with the maximum of the central mode preceding the toroidal location of the ELM precursor and, therefore, a hitherto unobserved coupling between central modes and ELMs. Highly elongated plasmas promise several advantages, among them higher current and beta limits. During TCV experiments dedicated to an increasing of the plasma elongation, a new disruptive current limit, at values well below the conventional current limit corresponding to q{sub a} > 2, was encountered for {kappa} > 2.3. This limit, which is preceded by a kink-type mode, is found to be consistent with ideal MHD stability calculations. The TCV observations, therefore, provide the first experimental confirmation of a deviation of the linear Troyon-scaling of the ideal beta limit with normalised current at high elongation, which was predicted over 10 years ago. Neoclassical tearing modes (NTMs), which have been observed to limit the achievable beta in a number of tokamaks, arise from a helical perturbation of the bootstrap current caused by an existing seed island. Neoclassical m/n = 2/1 tearing modes have been

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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)

Advanced tokamak research in DIII-D

International Nuclear Information System (INIS)

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

2004-01-01

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

Equilibrium and ballooning mode stability of an axisymmetric tensor pressure tokamak

International Nuclear Information System (INIS)

Cooper, W.A.; Bateman, G.; Nelson, D.B.; Kammash, T.

1980-08-01

A force balance relation, a representation for the poloidal beta (β/sub p/), and expressions for the current densities are derived from the MHD equilibrium relations for an axisymmetric tensor pressure tokamak. Perpendicular and parallel beam pressure components are evaluated from a distribution function that models high energy neutral particle injection. A double adiabatic energy principle is derived from that of Kruskal and Oberman, with correction terms added. The energy principle is then applied to an arbitrary cross-section axisymmetric tokamak to examine ballooning instabilities of large toroidal mode number. The resulting Euler equation is remarkably similar to that of ideal MHD. Although the field-bending term is virtually unaltered, the driving term is modified because the pressures are no longer constant on a flux surface. Either a necessary or a sufficient marginal stability criterion for a guiding center plasma can be derived from this equation whenever an additional stabilizing element unique to the double adiabatic theory is either kept or neglected, respectively

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

System studies of compact ignition tokamaks

International Nuclear Information System (INIS)

Galambos, J.D.; Peng, Y.K.M.; Blackfield, D.T.

1986-01-01

A new version of the FEDC Tokamak System Code (TSC) has been developed to analyze the Compact Ignition Tokamak (CIT). These proposed experiments have small (major radius F 1.5m) and high magnetic fields (B J 10T), and are characterized by reduced cost. Key design constraints of CIT include limits to the high stress levels in the magnetic coils, limits to the large temperature rises in the coils and on the first wall or divertor plate, minimizing power supply requirements, and assuring adequate plasma performance in fusion ignition and burn time consistent with the latest physics understanding. We present systems code level studies of CIT parameter space here for a range of design options with various design constraints. The present version of the TSC incorporates new models for key components of CIT. For example, new algorithms have been incorporated for calculating stress levels in the TFC and ohmic solenoid, temperature rise in the magnetic coils, peak power requirements, plasma MHD equilibrium and volt-second capability. The code also incorporates a numerical optimizer to find combinations of engineering quantities (device size, coil sizes, coil current densities etc.) and physics quantities (plasma density temperature, and beta, etc.) which satisfy all the constraints and can minimize or maximize a figure of merit (e.g., the major radius). This method was recently used in a mirror reactor system code (3) for the Minimara concept development

Fishbone mode in high-β discharges of spherical tokamaks

International Nuclear Information System (INIS)

Kolesnichenko, Ya.I.; Lutsenko, V.V.; Marchenko, V.S.

2000-01-01

Using Hamiltonian formalism, it has been shown that well-trapped energetic ions moving outwards consume the energy of MHD perturbations through the precessional resonance provided that the plasma pressure is sufficiently high. This supports the conclusion of recent publication that the fishbone mode is stabilized in high-β discharges of spherical tokamaks. It has also been found that the presence of the velocity anisotropy of energetic ions does not change this conclusion. (author)

Turbulence of high-beta plasma

International Nuclear Information System (INIS)

Khvesyuk, V.I.; Chirkov, A.Y.

1999-01-01

Principals of numerical modelling of turbulence in high-beta plasma (β > 0.1) are discussed. Creation of transport model for axial symmetric nonuniform confining magnetic field is considered. Numerical model of plasma turbulence in FRC is presented. The physical and mathematical models are formulated from nonuniform axial symmetric high-beta plasma. It is shown that influence of waves arise under this plasma conditions lead to chaotic motion of charged particles across magnetic field. (author)

Status of the tokamak program

Science.gov (United States)

Sheffield, J.

1981-08-01

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

Research using small tokamaks

International Nuclear Information System (INIS)

1991-05-01

discharges, production and self-organization of a turbulent plasma column in a spheromak (''SK-CG-1''), and (iv) a planned large-aspect ratio, high-beta tokamak (HBT-EP) experiment. Refs, figs and tabs

Design and operation of Alvand II C tokamak

International Nuclear Information System (INIS)

Avakian, M.; Azodi, H.; Naraghi, M.; Tabatabaie, H.; Rezvani, B.; Shahnazi, Sh.; Behrozinia, J.; Solaimani, M.

1984-01-01

ALVAND IIC is a research tokamak having a circular cross section with major radius of 45.5 cm and minor radius of 12.6 cm, which has been designed and constructed in the plasma physics division of NRC of AEOI. It is used for developing various diagnostics for low beta plasmas and study of the physical characteristics of these plasmas. This paper discusses the design of ALVAND IIC. (Author)

Tokamak

International Nuclear Information System (INIS)

Wesson, John.

1996-01-01

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

HYFIRE: a tokamak/high-temperature electrolysis system

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.P.; Benenati, R.; Varljen, T.C.; Chi, J.W.H.; Karbowski, J.S.

1981-01-01

The HYFIRE studies to date have investigated a number of technical approaches for using the thermal energy produced in a high-temperature Tokamak blanket to provide the electrical and thermal energy required to drive a high-temperature (> 1000 0 C) water electrolysis process. Current emphasis is on two design points, one consistent with electrolyzer peak inlet temperatures of 1400 0 C, which is an extrapolation of present experience, and one consistent with a peak electrolyzer temperature of 1100 0 C. This latter condition is based on current laboratory experience with high-temperature solid electrolyte fuel cells. Our major conclusion to date is that the technical integration of fusion and high-temperature electrolysis appears to be feasible and that overall hydrogen production efficiencies of 50 to 55% seem possible

Design of Tokamak plasma with high Tc superconducting coils

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

1992-09-01

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

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

Simulation of saturated tearing modes in tokamaks

International Nuclear Information System (INIS)

Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.

2004-01-01

A quasi-linear model, which includes the effect of the neoclassical bootstrap current, is developed for saturated tearing modes in order to compute magnetic island widths in axisymmetric toroidal plasmas with arbitrary aspect ratio and cross-sectional shape. The model is tested in a simple stand-alone code and is implemented in the BALDUR [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1982)] predictive modeling code. It is found that the widths of tearing mode islands increase with decreasing aspect ratio and with increasing elongation. Also, the island widths increase when the gradient of the current density increases at the edge of the islands and when the current density inside the islands is suppressed, such as the suppression caused by the near absence of the bootstrap current within the islands. In simulations of tokamak discharges, it is found that tearing mode island widths oscillate in time in response to periodic sawtooth crashes. The local enhancements in the transport produced by magnetic islands have a noticeable effect on global plasma confinement in simulations of low aspect ratio, high beta tokamaks, where saturated tearing mode islands can occur with widths that are greater than 15% of the plasma minor radius

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

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

Accelerator technology in tokamaks

International Nuclear Information System (INIS)

Kustom, R.L.

1977-01-01

This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

STATIONARY HIGH-PERFORMANCE DISCHARGES IN THE DIII-D TOKAMAK

International Nuclear Information System (INIS)

LUCE, TC; WADE, MR; FERRON, JR; HYATT, AW; KELLMAN, AG; KINSEY, JE; LAHAYE, RJ; LASNIER, CJ; MURAKAMI, M; POLITZER, PA; SCOVILLE, JT

2002-01-01

A271 STATIONARY HIGH-PERFORMANCE DISCHARGES IN THE DII-D TOKAMAK. Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak under stationary conditions at relatively low plasma current (q 95 > 4). A figure of merit for fusion gain (β N H 89 /q 95 2 ) has been maintained at values corresponding to | = 10 operation in a burning plasma for > 6 s or 36τ E and 2τ R . The key element is the relaxation of the current profile to a stationary state with q min > 1. In the absence of sawteeth and fishbones, stable operation has been achieved up to the estimated no-wall β limit. Feedback control of the energy content and particle inventory allow reproducible, stationary operation. The particle inventory is controlled by gas fueling and active pumping; the wall plays only a small role in the particle balance. The reduced current lessens significantly the potential for structural damage in the event of a major disruption. In addition, the pulse length capability is greatly increased, which is essential for a technology testing phase of a burning plasma experiment where fluence (duty cycle) is important

Divertor design for the Tokamak Physics Experiment

International Nuclear Information System (INIS)

Hill, D.N.; Braams, B.

1994-05-01

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

Ideal MHD beta-limits of poloidally asymmetric equilibria

Energy Technology Data Exchange (ETDEWEB)

Todd, A.M.M.; Miller, A.E.; Grimm, R.C.; Okabayashi, M.; Dalhed, H.E. Jr.

1981-05-01

The ideal MHD stability of poloidally asymmetric equilibria, which are typical of a tokamak reactor design with a single-null poloidal divertor is examined. As with symmetric equilibria, stability to non-axisymmetric modes improves with increasing triangularity and ellipticity, and with lower edge safety factor. Pressure profiles optimized with respect to ballooning stability are obtained for an asymmetric shape, resulting in ..beta../sub critical/ approx. = 5.7%. The corresponding value for an equivalent symmetric shape is ..beta../sub critical/ approx. = 6.5%.

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

Planned upgrade to the coaxial plasma source facility for high heat flux plasma flows relevant to tokamak disruption simulations

International Nuclear Information System (INIS)

Caress, R.W.; Mayo, R.M.; Carter, T.A.

1995-01-01

Plasma disruptions in tokamaks remain serious obstacles to the demonstration of economical fusion power. In disruption simulation experiments, some important effects have not been taken into account. Present disruption simulation experimental data do not include effects of the high magnetic fields expected near the PFCs in a tokamak major disruption. In addition, temporal and spatial scales are much too short in present simulation devices to be of direct relevance to tokamak disruptions. To address some of these inadequacies, an experimental program is planned at North Carolina State University employing an upgrade to the Coaxial Plasma Source (CPS-1) magnetized coaxial plasma gun facility. The advantages of the CPS-1 plasma source over present disruption simulation devices include the ability to irradiate large material samples at extremely high areal energy densities, and the ability to perform these material studies in the presence of a high magnetic field. Other tokamak disruption relevant features of CPS-1U include a high ion temperature, high electron temperature, and long pulse length

Optimized profiles for improved confinement and stability in the Dill-D tokamak

International Nuclear Information System (INIS)

Taylor, T.S.; St John, H.; Turnbull, A.D.

1994-01-01

Simultaneous achievement of high energy confinement, τ E , and high plasma beta, β, leads to an economically attractive compact tokamak fusion reactor. High confinement enhancement, H τ E /τ E -ITER89P 4, and high normalized beta β N β/(I/aB) = 6%-m-T/MA. have been obtained in DIII-D experimental discharges. These improved confinement and/or improved stability limits are observed in several DIII-D high performance operational regimes: VH-mode, high l i H-mode, second stable core, and high beta poloidal. We have identified several important features of the improved performance in these discharges: details of the plasma shape, toroidal rotation or ExB flow profile, q profile and current density profile, and pressure profile. From our improved physics understanding of these enhanced performance regimes, we have developed operational scenarios which maintain the essential features of the improved confinement and which increase the stability limits using localized current profile control. The stability limit is increased by modifying the interior safety factor profile to be nonmonotonic with high central q, while maintaining the edge current density consistent with the improved transport regimes and the high edge bootstrap current. We have calculated high beta equilibria with β N 6.5, stable to ideal n=1 kinks and stable to ideal ballooning modes. The safety factor at the 95% flux surface is 6, the central q value is 3.9 and the minimum in q is 2.6. The current density profile is maintained by the natural profile of the bootstrap current, and a modest amount of electron cyclotron current drive. (Author)

Optimized profiles for improved confinement and stability in the DIII-D tokamak

International Nuclear Information System (INIS)

Taylor, T.S.; St. John, H.; Turnbull, A.D.

1995-02-01

Simultaneous achievement of high energy confinement, τ E , and high plasma beta, β, leads to an economically attractive compact tokamak fusion reactor. High confinement enhancement, H = τ E /τ E-ITER89P = 4, and high normalized beta β N = β/(I/aB) = 6%-m-T/MA, have been obtained in DIII-D experimental discharges. These improved confinement and/or improved stability limits are observed in several DIII-D high performance operational regimes: VH-mode, high ell i H-mode, second stable core, and high beta poloidal. The authors have identified several important features of the improved performance in these discharges: details of the plasma shape, toroidal rotation or ExB flow profile, q profile and current density profile, and pressure profile. From the improved physics understanding of these enhanced performance regimes, they have developed operational scenarios which maintain the essential features of the improved confinement and which increase the stability limits using localized current profile control. The stability limit is increased by modifying the interior safety factor profile to be nonmonotonic with high central q, while maintaining the edge current density consistent with the improved transport regimes and the high edge bootstrap current. They have calculated high beta equilibria with β N = 6.5, stable to ideal n = 1 kinks and stable to ideal ballooning modes. The safety factor at the 95% flux surface is 6, the central q value is 3.9 and the minimum in q is 2.6. The current density profile is maintained by the natural profile of the bootstrap current, and a modest amount of electron cyclotron current drive

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

Neoclassical transport analysis for a class of high-β tokamak equilibria

International Nuclear Information System (INIS)

Rieser, H.; Werthmann, H.; Kuhn, S.

1995-01-01

Balescu's neoclassical transport theory is extended to the case of non-circular flux-surface geometries. Modified classical and neoclassical transport equations, governing particle and heat fluxes in the short- and long-mean-free-path regimes, are derived. These equations are shown to coincide to leading order with the corresponding equations given by Hirshman and Sigmar. They are then applied to an ideal MHD equilibrium, suitable as a simplified but analytically tractable model of a high-β tokamak. Numerical results for the radial profiles of the global (i.e. flux-surface integrated) particle and heat fluxes in the classical, Pfirsch-Schlueter and banana regimes are presented for geometry and plasma parameters realized in some tokamaks, like the divertor and injection tokamak experiment (DITE). This spatial representation provides direct insight into the overall collisional transport behaviour of a given equilibrium, whereas the anomalous transport problem is not addressed here. Our results demonstrate that for a given pressure profile the global neoclassical fluxes may depend very sensitively on the temperature profiles and that, in particular, the global classical and neoclassical ion heat fluxes exhibit a characteristic non-monotonic behaviour. (author)

DAMAVAND - An Iranian tokamak with a highly elongated plasma cross-section

International Nuclear Information System (INIS)

Amrollahi, R.

1997-01-01

The ''DAMAVAND'' facility is an Iranian Tokamak with a highly elongated plasma cross-section and with a poloidal divertor. This Tokamak has the advantage to allow the plasma physics research under the conditions similar to those of ITER magnetic configuration. For example, the opportunity to reproduce partially the plasma disruptions without sacrificing the studies of: equilibrium, stability and control over the elongated plasma cross-section; processes in the near-wall plasma; auxiliary heating systems, etc. The range of plasma parameters, the configuration of ''DAMAVAND'' magnetic coils and passive loops, and their location within the vacuum chamber allow the creation of the plasma at the center of the vacuum chamber and the production of two poloidal volumes (upper and lower) for the divertor. (author)

Stability of tearing modes in tokamak plasmas

International Nuclear Information System (INIS)

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

1994-02-01

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

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

Analytic, high β, flux conserving equilibria for cylindrical tokamaks

International Nuclear Information System (INIS)

Sigmar, D.J.; Vahala, G.

1978-09-01

Using Grad's theory of generalized differential equations, the temporal evolution from low to high β due to ''adiabatic'' and nonadiabatic (i.e., neutral beam injection) heating of a cylindrical tokamak plasma with circular cross section and peaked current profiles is calculated analytically. The influence of shaping the initial safety factor profile and the beam deposition profile and the effect of minor radius compression on the equilibrium is analyzed

Analytic, high β, flux conserving equilibria for cylindrical tokamaks

International Nuclear Information System (INIS)

Sigmar, D.J.; Vahala, G.

1978-01-01

Using Grad's theory of generalized differential equations, the temporal evolution from low to high β due to ''adiabatic'' and nonadiabatic (i.e., neutral beam injection) heating of a cylindrical tokamak plasma with circular cross section and peaked current profiles is calculated analytically. The influence of shaping the initial safety factor profile and the beam deposition profile and the effect of minor radius compression on the equilibrium is analyzed

DEALS magnet concept and its applcations to high density, high field tokamak systems

International Nuclear Information System (INIS)

Hsieh, S.Y.; Powell, J.; Lehner, J.; Bezler, P.; Laverick, C.; Finkelman, M.; Brown, T.; Bundy, J.

1977-01-01

The goal of the DEALS program is to develop a demountable TF magnet system concept that will reduce construction and life cycle costs, enhance the accessibility of components inside the coil system, and increase the chances for being able to use large high-field magnet systems in post TFTR reactor experiments. These experiments are projected to occur during the mid 1980's, with conceptual designs beginning in two or three years. A number of recent studies have highlighted the need for Tokamak fusion reactor systems with reasonable down time for maintenance and repair and realistic operating capacity factors, as well as the need for smaller, lower cost reactors. Two scoping studies were carried out of recent Tokamak system concepts incorporating conventionally wound coils to assess the possibilities of using demountable coils of rectangular section with an active support system and a third more intensive study using a passive support with slight movement of the joints. These studies are described briefly

Resistive MHD studies of high-β-tokamak plasmas

International Nuclear Information System (INIS)

Lynch, V.E.; Carreras, B.A.; Hicks, H.R.; Holmes, J.A.; Garcia, L.

1981-01-01

Numerical calculations have been performed to study the MHD activity in high-β tokamaks such as ISX-B. These initial value calculations built on earlier low β techniques, but the β effects create several new numerical issues. These issues are discussed and resolved. In addition to time-stepping modules, our system of computer codes includes equilibrium solvers (used to provide an initial condition) and output modules, such as a magnetic field line follower and an X-ray diagnostic code. The transition from current driven modes at low β to predominantly pressure driven modes at high β is described. The nonlinear studies yield X-ray emissivity plots which are compared with experiment

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.)

Impact of maximum TF magnetic field on performance and cost of an advanced physics tokamak

International Nuclear Information System (INIS)

Reid, R.L.

1983-01-01

Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of variation in the maximum value of the field at the toroidal field (TF) coils on the performance and cost of a low q/sub psi/, quasi-steady-state tokamak. Marginal ignition, inductive current startup plus 100 s of inductive burn, and a constant value of epsilon (inverse aspect ratio) times beta poloidal were global conditions imposed on this study. A maximum TF field of approximately 10 T was found to be appropriate for this device

High-beta experiments with neutral-beam injection on PDX

International Nuclear Information System (INIS)

Johnson, D.; Bell, M.; Bitter, M.

1983-01-01

Experimental investigations of high-beta plasmas produced in PDX with near-perpendicular neutral-beam injection are reported. Systematic power scans have been performed over a wide range of toroidal fields (νsub(T)q.7 T< Bsub(T)<2.2 T) and plasma currents (200 kA< Isub(p)<500 kA). At high toroidal fields, the change in total stored energy due to beam injection increases linearly with input power and also increases with plasma current. At lower toroidal fields and low injection power levels, the stored energy also increases with power and plasma current. However, at high power and low toroidal fields, a saturation in heating is observed. This result suggests the onset of a νsub(T) limit for circular cross-section tokamaks with near-perpendicular injection. Scaling experiments indicate that this νsub(T) limit increases with rising 1/q. Values of νsub(T)approx.=3% at qsub(PSI)=1.8 have been achieved. At high values of νsub(T)q, short bursts of MHD activity are observed, synchronized with sharply increased fluxes of perpendicular charge-exchange neutrals and rapid decreases in the rate of beam-driven neutron production. When strong bursts occur, there is a significant depletion of the fast-ion population. Estimates of the fast-ion loss indicate that it could explain the observed decrease in heating, although an additional reduction in thermal-plasma confinement cannot be ruled out. Numerical studies using measured pressure profiles predict that the equilibria obtained become unstable to the ideal n=1 internal mode, at about the same value of 0 where the new fluctuations are observed. (author)

Beta decay of highly charged ions

International Nuclear Information System (INIS)

Litvinov, Yuri A; Bosch, Fritz

2011-01-01

Beta decay of highly charged ions has attracted much attention in recent years. An obvious motivation for this research is that stellar nucleosynthesis proceeds at high temperatures where the involved atoms are highly ionized. Another important reason is addressing decays of well-defined quantum-mechanical systems, such as one-electron ions where all interactions with other electrons are excluded. The largest modifications of nuclear half-lives with respect to neutral atoms have been observed in beta decay of highly charged ions. These studies can be performed solely at ion storage rings and ion traps, because there high atomic charge states can be preserved for extended periods of time (up to several hours). Currently, all experimental results available in this field originate from experiments at the heavy-ion complex GSI in Darmstadt. There, the fragment separator facility FRS allows the production and separation of exotic, highly charged nuclides, which can then be stored and investigated in the storage ring facility ESR. In this review, we present and discuss in particular two-body beta decays, namely bound-state beta decay and orbital electron capture. Although we focus on experiments conducted at GSI, we will also attempt to provide general requirements common to any other experiment in this context. Finally, we address challenging but not yet performed experiments and we give prospects for the new radioactive beam facilities, such as FAIR in Darmstadt, IMP in Lanzhou and RIKEN in Wako.

Confinement of ohmically heated plasmas and turbulent heating in high-magnetic field tokamak TRIAM-1

Energy Technology Data Exchange (ETDEWEB)

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

1979-12-01

TRIAM-1, the tokamak device with high toroidal magnetic field, has been constructed to establish the scaling laws of advanced tokamak devices such as Alcator, and to study the possibility of the turbulent heating as a further economical heating method of the fusion oriented plasmas. The plasma parameters obtained by ohmic heating alone are as follows; central electron temperature T sub(e0) = 640 eV, central ion temperature T sub(i0) = 280 eV and line-average electron density n average sub(e) = 2.2 x 10/sup 14/ cm/sup -3/. The empirical scaling laws are investigated concerning T sub(e0), T sub(i0) and n average sub(e). The turbulent heating has been carried out by applying the high electric field in the toroidal direction to the typical tokamak discharge with T sub(i0) asymptotically equals 200 eV. The efficient ion heating is observed and T sub(i0) attains to about 600 eV.

Achieving high fusion reactivity in high poloidal beta discharges in TFTR

International Nuclear Information System (INIS)

Manuel, M.E.; Navratil, G.A.; Sabbagh, S.A.; Batha, S.; Bell, M.G.; Bell, R.; Budny, R.V.; Bush, C.E.; Cavallo, A.; Chance, M.S.; Cheng, C.Z.; Efthimion, P.C.; Fredrickson, E.D.; Fu, G.Y.; Hawryluk, R.J.; Janos, A.C.; Jassby, D.L.; Levinton, F.; Mikkelsen, D.R.; Manickam, J.; McCune, D.C.; McGuire, K.M.; Medley, S.S.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Park, H.K.; Ramsey, A.T.; Stratton, B.C.; Synakowski, E.J.; Taylor, G.; Wieland, R.M.; Yamada, M.; Zarnstorff, M.C.: Zweben, S.; Kesner, J.; Marmar, E.; Snipes, J.; Terry, J.

1993-04-01

High poloidal beta discharges have been produced in TFTR that achieved high fusion reactivities at low plasma currents. By rapidly decreasing the plasma current just prior to high-power neutral beam injection, relatively peaked current profiles were created having high l i > 2, high Troyon-normalized beta, βN > 3, and high poloidal beta. β p ≥ 0.7 R/a. The global energy confinement time after the current ramp was comparable to supershots, and the combination of improved MHD stability and good confinement produced a new high εβ p high Q DD operating mode for TFTR. Without steady-state current profile control, as the pulse lengths of high βp discharges were extended, l i decreased, and the improved stability produced immediately after by the current ramp deteriorated. In four second, high εβ p discharges, the current profile broadened under the influence of bootstrap and beam-drive currents. When the calculated voltage throughout the plasma nearly vanished, MHD instabilities were observed with β N as low as 1.4. Ideal MHD stability calculations showed this lower beta limit to be consistent with theoretical expectations

Application studies of spherical tokamak plasma merging

International Nuclear Information System (INIS)

Ono, Yasushi; Inomoto, Michiaki

2012-01-01

The experiment of plasma merging and heating has long history in compact torus studies since Wells. The study of spherical tokamak (ST), starting from TS-3 plasma merging experiment of Tokyo University in the late 1980s, is followed by START of Culham laboratory in the 1900s, TS-4 and UTST of Tokyo University and MAST of Culham laboratory in the 2000s, and last year by VEST of Soul University. ST has the following advantages: 1) plasma heating by magnetic reconnection at a MW-GW level, 2) rapid start-up of high beta plasma, 3) current drive/flux multiplication and distribution control of ST plasma, 4) fueling and helium-ash exhaust. In the present article, we emphasize that magnetic reconnection and plasma merging phenomena are important in ST plasma study as well as in plasma physics. (author)

Measurement of local, internal magnetic fluctuations via cross-polarization scattering in the DIII-D tokamak (invited)

Energy Technology Data Exchange (ETDEWEB)

Barada, K., E-mail: kshitish@ucla.edu; Rhodes, T. L.; Crocker, N. A.; Peebles, W. A. [University of California-Los Angeles, P.O. Box 957099, Los Angeles, California 90095 (United States)

2016-11-15

We present new measurements of internal magnetic fluctuations obtained with a novel eight channel cross polarization scattering (CPS) system installed on the DIII-D tokamak. Measurements of internal, localized magnetic fluctuations provide a window on an important physics quantity that we heretofore have had little information on. Importantly, these measurements provide a new ability to challenge and test linear and nonlinear simulations and basic theory. The CPS method, based upon the scattering of an incident microwave beam into the opposite polarization by magnetic fluctuations, has been significantly extended and improved over the method as originally developed on the Tore Supra tokamak. A new scattering geometry, provided by a unique probe beam, is utilized to improve the spatial localization and wavenumber range. Remotely controllable polarizer and mirror angles allow polarization matching and wavenumber selection for a range of plasma conditions. The quasi-optical system design, its advantages and challenges, as well as important physics validation tests are presented and discussed. Effect of plasma beta (ratio of kinetic to magnetic pressure) on both density and magnetic fluctuations is studied and it is observed that internal magnetic fluctuations increase with beta. During certain quiescent high confinement operational regimes, coherent low frequency modes not detected by magnetic probes are detected locally by CPS diagnostics.

HYFIRE: a tokamak-high-temperature electrolysis system

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.; Steinberg, M.; Benenati, R.; Horn, F.; Isaacs, H.; Lazareth, O.W.; Makowitz, H.; Usher, J.

1980-01-01

Brookhaven National Laboratory (BNL) is carrying out a comprehensive conceptual design study called HYFIRE of a commercial fusion Tokamak reactor, high-temperature electrolysis system. The study is placing particular emphasis on the adaptability of the STARFIRE power reactor to a synfuel application. The HYFIRE blanket must perform three functions: (a) provide high-temperature (approx. 1400 0 C) process steam at moderate pressures (in the range of 10 to 30 atm) to the high-temperature electrolysis (HTE) units; (b) provide high-temperature (approx. 700 0 to 800 0 C) heat to a thermal power cycle for generation of electricity to the HTE units; and (c) breed enough tritium to sustain the D-T fuel cycle. In addition to thermal energy for the decomposition of steam into its constituents, H 2 and O 2 , electrical input is required. Fourteen hundred degree steam coupled with 40% power efficiency results in a process efficiency (conversion of fusion energy to hydrogen chemical energy) of 50%

HYFIRE: a tokamak-high-temperature electrolysis system

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.; Steinberg, M.; Benenati, R.; Horn, F.; Isaacs, H.; Lazareth, O.W.; Makowitz, H.; Usher, J.

1980-01-01

Brookhaven National Laboratory (BNL) is carrying out a comprehensive conceptual design study called HYFIRE of a commercial fusion Tokamak reactor, high-temperature electrolysis system. The study is placing particular emphasis on the adaptability of the STARFIRE power reactor to a synfuel application. The HYFIRE blanket must perform three functions: (a) provide high-temperature (approx. 1400 0 C) process steam at moderate pressures (in the range of 10 to 30 atm) to the high-temperature electrolysis (HTE) units; (b) provide high-temperature (approx. 700 0 to 800 0 C) heat to a thermal power cycle for generation of electricity to the HTE units; and (c) breed enough tritium to sustain the D-T fuel cycle. In addition to thermal energy for the decomposition of steam into its constituents, H 2 and O 2 , electrical input is required. Fourteen hundred degree steam coupled with 40% power cycle efficiency results in a process efficiency (conversion of fusion energy to hydrogen chemical energy) of 50%

Equilibrium and stability of high-beta toroidal plasmas with toroidal and poloidal flow in reduced magnetohydrodynamic models

International Nuclear Information System (INIS)

Ito, A.; Nakajima, N.

2010-11-01

Effects of flow, finite ion temperature and pressure anisotropy on equilibrium and stability of a high-beta toroidal plasma are studied in the framework of reduced magnetohydrodynamics (MHD). A set of reduced equilibrium equations for high-beta tokamaks with toroidal and poloidal flow comparable to the poloidal sound velocity is derived in a unified form of single-fluid and Hall MHD models and a two-fluid MHD model with ion finite Larmor radius (FLR) terms. Pressure anisotropy is introduced with equations for the parallel heat flux which are closed by a fluid closure model. It is solved analytically for the single-fluid model and the solutions shows complicated characteristics in the region around the poloidal sound velocity due to pressure anisotropy and the parallel heat flux. Numerical solutions are found by using the finite element method for the two-fluid model with FLR effects in the case of isotropic, adiabatic pressure and indicate the following features of two-fluid equilibria: the isosurfaces of the magnetic flux, the pressure and the ion stream function do not coincide with each other, and the solutions depend on the sign of the radial electric field. Reduced single-fluid MHD equations with time evolution that are consistent with the above equilibria are also derived in order to study their stability. They conserve the energy up to the order required by the equilibria. (author)

Drift-kinetic Alfven modes in high performance tokamaks

International Nuclear Information System (INIS)

Jaun, A.; Fasoli, A.F.; Testa, D.; Vaclavik, J.; Villard, L.

2001-01-01

The stability of fast-particle driven Alfven eigenmodes is modeled in high performance tokamaks, successively with a conventional shear, an optimized shear and a tight aspect ratio plasma. A large bulk pressure yields global kinetic Alfven eigenmodes that are stabilized by mode conversion in the presence of a divertor. This suggests how conventional reactor scenarii could withstand significant pressure gradients from the fusion products. A large safety factor in the core q 0 >2.5 in deeply shear reversed configurations and a relatively large bulk ion Larmor radius in a low magnetic field can trigger global drift-kinetic Alfven eigenmodes that are unstable in high performance JET, NSTX and ITER plasmas. (author)

Compression experiments on the TOSKA tokamak

International Nuclear Information System (INIS)

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

1980-10-01

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

Toroidal microinstability studies of high temperature tokamaks

International Nuclear Information System (INIS)

Rewoldt, G.; Tang, W.M.

1989-07-01

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

Strong Scattering of High Power Millimeter Waves in Tokamak Plasmas with Tearing Modes

DEFF Research Database (Denmark)

Westerhof, E.; Nielsen, Stefan Kragh; Oosterbeek, J.W.

2009-01-01

In tokamak plasmas with a tearing mode, strong scattering of high power millimeter waves, as used for heating and noninductive current drive, is shown to occur. This new wave scattering phenomenon is shown to be related to the passage of the O point of a magnetic island through the high power...

Progress in Developing a High-Availability Advanced Tokamak Pilot Plant

Energy Technology Data Exchange (ETDEWEB)

Brown, T.; Goldston, R.; Kessel, C.; Neilson, G.; Menard, J.; Prager, S.; Scott, S.; Titus, P.; Zarnstorff, M., E-mail: tbrown@pppl.gov [Princeton University, Princeton Plasma Physics Laboratory, Princeton (United States); Costley, A. [Henley on Thames (United Kingdom); El-Guebaly, L. [University of Wisconsin, Madison (United States); Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Waganer, L. [St. Louis (United States)

2012-09-15

Full text: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The mission of the pilot plant was set to encompass component test and fusion nuclear science missions yet produce net electricity with high availability in a device designed to be prototypical of the commercial device. The objective of the study was to evaluate three different magnetic configuration options, the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS) in an effort to establish component characteristics, maintenance features and the general arrangement of each candidate device. With the move to look beyond ITER the fusion community is now beginning to embark on DEMO reactor studies with an emphasis on defining configuration arrangements that can meet a high availability goal. In this paper the AT pilot plant design will be presented. The selected maintenance approach, the device arrangement and sizing of the in-vessel components and details of interfacing auxiliary systems and services that impact the ability to achieve high availability operations will be discussed. Efforts made to enhance the interaction of in-vessel maintenance activities, the hot cell and the transfer process to develop simplifying solutions will also be addressed. (author)

Numerical studies of edge localized instabilities in tokamaks

International Nuclear Information System (INIS)

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

2002-01-01

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

Plasma heating and fuelling in the Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.K.; Barsukov, A.G.; Belyakov, V.A.

2005-01-01

The results of the last two years of plasma investigations at Globus-M are presented. Described are improvements helping to achieve high performance OH plasmas, which are used as the target for auxiliary heating and fuelling experiments. Increased energy content, high beta poloidal and good confinement are reported. Experiments on NBI plasma heating with a wide range of plasma parameters were performed. Some results are presented and analyzed. Experiments on RF plasma heating in the frequency range of fundamental ion cyclotron harmonics are described. In some experiments which were performed for the first time in spherical tokamaks, promising results were achieved. Noticeable ion heating was recorded at low launched power and a high concentration of hydrogen minority in deuterium plasmas. Simulations of RF wave absorption are briefly discussed. Described also are modification of the plasma gun and test-stand experiments. Fuelling experiments performed at Globus-M are discussed. (author)

Sawtooth crashes at high beta on JET

Energy Technology Data Exchange (ETDEWEB)

Alper, B; Huysmans, G T.A.; Sips, A C.C. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Nave, M F.F. [Universidade Tecnica, Lisbon (Portugal). Inst. Superior Tecnico

1994-07-01

The sawtooth crashes on JET display features which depend on beta. The main observation is a transient bulging of flux surfaces (duration inferior to 30 microsec.), which is predominantly on the low field side and extends to larger radii as beta increases. This phenomenon reaches the plasma boundary when beta{sub N} exceeds 0.5 and in these cases is followed by an ELM within 50 microsec. These sawtooth/ELM events limit plasma performance. Modelling of mode coupling shows qualitative agreement between observations of the structure of the sawtooth precursor and the calculated internal kink mode at high beta. (authors). 6 refs., 5 figs.

Modelling of advanced tokamak physics scenarios in ALCATOR C-Mod

International Nuclear Information System (INIS)

Bonoli, P.T.; Porkolab, M.; Ramos, J.

2001-01-01

Advanced tokamak modes of operation in Alcator C-Mod have been investigated using a simulation model which combines an MHD equilibrium and current profile control calculation with an ideal MHD stability analysis. Stable access to high β t operating modes with reversed shear current density profiles has been demonstrated using 2.4-3.0 MW of off-axis lower hybrid current drive (LHCD). Here β t =2μ 0 (p)/B 2 0 is the volume averaged toroidal plasma beta. Current profile control at the β-limit and beyond has also been demonstrated. The effects of LH power level as well as changes in the profiles of density and temperature on shear reversal radius have been quantified and are discussed. (author)

Construction of a high beta plasma source

International Nuclear Information System (INIS)

Naraghi, M.; Torabi-Fard, A.

1976-02-01

A high beta plasma source has been designed and constructed. This source will serve as a means of developing and exercising different diagnostic techniques as required for ALVAND I, linear theta pinch experiment. Also, it will serve to acquaint the technicians with some of the techniques and safety rules of high voltage and capacitor discharge experiments. The operating parameters of the theta pinch and Z-pinch preionization is presented and the program of diagnostic measurements on the high beta plasma source is discussed

High-beta linac structures

International Nuclear Information System (INIS)

Schriber, S.O.

1979-01-01

Accelerating structures for high-beta linacs that have been and are in use are reviewed in terms of their performance. Particular emphasis is given to room-temperature structures and the disk-and-washer structure. The disk-and-washer structure has many attractive features that are discussed for pulsed high-gradient linacs, for 100% duty-cycle medium-gradient linacs and for high-current linacs requiring maximal amounts of stored energy in the electric fields available to the beam

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

Multi-megajoule heating of large tokamaks with high energy heavy ion beams

International Nuclear Information System (INIS)

Dei-Cas, R.

1981-07-01

The fast neutral injection heating and RF heating for tokamak like plasmas are now well established. We consider in this paper the use of high energy (approximately 1 GeV) heavy ions (Xe 132 ) to reach ignition in JET or INTOR like tokamaks. The main advantages of such a method will be outlined. The capture and the confinement of heavy ions have been analysed in a particular case and with the described RF linac it seems possible to inject in the order of 50 MJ in 1 sec with a modest increase of the effective charge Zsub(eff)<1.05 in a JET-like plasma for a particle life time of 1 sec and then the additional radiated power should be maintained at a relatively low level in comparison to the injected power

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

Czech Academy of Sciences Publication Activity Database

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

2012-01-01

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

Optimization study of normal conductor tokamak for commercial neutron source

Science.gov (United States)

Fujita, T.; Sakai, R.; Okamoto, A.

2017-05-01

The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

Active and passive kink mode studies in a tokamak with a movable ferromagnetic wall

Energy Technology Data Exchange (ETDEWEB)

Levesque, J. P.; Hughes, P. E.; Bialek, J.; Byrne, P. J.; Mauel, M. E.; Navratil, G. A.; Peng, Q.; Rhodes, D. J.; Stoafer, C. C. [Department of Applied Physics and Applied Mathematics, Columbia University, 500 W. 120th Street, New York, New York 10027 (United States)

2015-05-15

High-resolution active and passive kink mode studies are conducted in a tokamak with an adjustable ferromagnetic wall near the plasma surface. Ferritic tiles made from 5.6 mm thick Hiperco{sup ®} 50 alloy have been mounted on the plasma-facing side of half of the in-vessel movable wall segments in the High Beta Tokamak-Extended Pulse device [D. A. Maurer et al., Plasma Phys. Controlled Fusion 53, 074016 (2011)] in order to explore ferritic resistive wall mode stability. Low-activation ferritic steels are a candidate for structural components of a fusion reactor, and these experiments examine MHD stability of plasmas with nearby ferromagnetic material. Plasma-wall separation for alternating ferritic and non-ferritic wall segments is adjusted between discharges without opening the vacuum vessel. Amplification of applied resonant magnetic perturbations and plasma disruptivity are observed to increase when the ferromagnetic wall is close to plasma surface instead of the standard stainless steel wall. Rapidly rotating m/n=3/1 external kink modes have higher growth rates with the nearby ferritic wall. Feedback suppression of kinks is still as effective as before the installation of ferritic material in vessel, in spite of increased mode growth rates.

Economic analyses of alpha channeling in tokamak power plants

International Nuclear Information System (INIS)

Ehst, D.A.

1998-01-01

The hot-ion-mode of operation [1] has long been thought to offer optimized performance for long-pulse or steady-state magnetic fusion power plants. This concept was revived in recent years when theoretical considerations suggested that nonthermal fusion alpha particles could be made to channel their power density preferentially to the fuel ions [2,3]. This so-called anomalous alpha particle slowing down can create plasmas with fuel ion temperate T i somewhat larger than the electron temperature T e , which puts more of the beta-limited plasma pressure into the useful fuel species (rather than non-reacting electrons). As we show here, this perceived benefit may be negligible or nonexistent for tokamaks with steady state current drive. It has likewise been argued [2,3] that alpha channeling could be arranged such that little or no external power would be needed to generate the steady state toroidal current. Under optimistic assumptions we show that such alpha-channeling current drive would moderately improve the economic performance of a first stability tokamak like ARIES-I [4], however a reversed-shear (advanced equilibrium) tokamak would likely not benefit since traditional radio-wave (rf) electron-heating current drive power would already be quite small

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

International Nuclear Information System (INIS)

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

1983-04-01

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

Three novel tokamak plasma regimes in TFTR

International Nuclear Information System (INIS)

Furth, H.P.

1985-10-01

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

Advanced tokamak burning plasma experiment

International Nuclear Information System (INIS)

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

2001-01-01

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

Tokamak experimental power reactor

International Nuclear Information System (INIS)

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

1978-01-01

A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

Resistive instabilities in tokamaks

International Nuclear Information System (INIS)

Rutherford, P.H.

1985-10-01

Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed

Tokamak experiments

International Nuclear Information System (INIS)

Robinson, D.C.

1987-01-01

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

High-frequency gyrotrons and their application to tokamak plasma heating

International Nuclear Information System (INIS)

Kreischer, K.E.

1981-01-01

A comprehensive analysis of high frequency (100 to 200 GHz) and high power (> 100 kW) gyrotrons has been conducted. It is shown that high frequencies will be required in order for electron cyclotron radiation to propagate to the center of a compact tokamak power reactor. High power levels will be needed in order to ignite the plasma with a reasonable number of gyrotron units. In the first part of this research, a set of analytic expressions, valid for all TE cavity modes and all harmonics, is derived for the starting current and frequency detuning using the Vlasov-Maxwell equations in the weakly relativistic limit. The use of an optical cavity is also investigated

Beta and current limits in the Doublet III tokamak

International Nuclear Information System (INIS)

Strait, E.J.; Chu, M.S.; Jahns, G.L.

1986-04-01

Neutral-beam heated discharges in Doublet III exhibit an operational beta limit, β/sub T/(%) less than or equal to 3.5 I(MA)/a(m)B(T), in good agreement with several theoretical predictions for ideal external kink or ballooning modes. These theories predict that the β limit has no explicit dependence on plasma shape (for nominal dee shapes). This aspect of the theory was confirmed in Doublet III by varying the elongation (kappa) from 1.0 to 1.6 and the triangularity (delta) from -0.1 to 0.9 and finding in all cases the same β limit. The maximum achievable beta thus depends on the minimum achievable value of the safety factor q. In Doublet III, the operational current limit is given by q greater than or equal to 1.7 for limiter-defined discharges and q greater than or equal to 2.7 for separatrix-defined discharges. Operation with q approx.2 was achieved for 1.0 less than or equal to kappa less than or equal to 1.6. Both β and q limits are characterized by major disruptions which usually terminate the discharge. In both cases, the disruptions often have a precursor oscillation with toroidal mode number n = 1, poloidal mode number m = 2 or 3, a frequency of zero to a few kHz, and a growth time on the order of a millisecond. These observations suggest that the proximate cause of these disruptions is a kink or tearing mode, pressure-driven in one case and current-driven in the other. Theoretical analyses of discharges at both limits will be compared. Modes with a high toroidal mode number, 3 less than or equal to n less than or equal to 5, and ballooning character have been observed near the β/sub T/ limit. These modes do not appear to be closely connected with the disruptions. Heating efficiency, ΔW/ΔP, remains constant up to the limiting disruption. Fishbone modes appear to be mainly a feature of high β/sub p/ operation and not connected to the β/sub T/ limit

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Long pulse high performance discharges in the DIII-D tokamak

International Nuclear Information System (INIS)

Luce, T.C.; Wade, M.R.; Politzer, P.A.

2001-01-01

Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 , indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β min >1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with β N H 89 ∼ 7 for up to 6.3 s or ∼34τ E . These discharges appear to have stationary current profiles with q min ∼1.05, in agreement with the current profile relaxation time ∼1.8 s. (author)

Updated tokamak systems code and applications to high-field ignition devices

International Nuclear Information System (INIS)

Reid, R.L.; Galambos, J.D.; Peng, Y-K.M.; Strickler, D.J.; Selcow, E.C.

1985-01-01

This paper describes revisions made to the Tokamak Systems Code to more accurately model high-field copper ignition devices. The major areas of revision were in the plasma physics model, the toroidal field (TF) coil model, and the poloidal field (PF) coil/MHD model. Also included in this paper are results obtained from applying the revised code to a study for a high-field copper ignition device to determine the impact of magnetic field on axis, (at the major radius), on performance, and on cost

Characterization of a beta-glycosidase highly active on disaccharides and of a beta-galactosidase from Tenebrio molitor midgut lumen.

Science.gov (United States)

Ferreira, Alexandre H P; Terra, Walter R; Ferreira, Clélia

2003-02-01

The midgut of the yellow mealworm, Tenebrio molitor L. (Coleoptera: Tenebrionidae) larvae has four beta-glycosidases. The properties of two of these enzymes (betaGly1 and betaGly2) have been described elsewhere. In this paper, the characterization of the other two glycosidases (betaGly3 and betaGly4) is described. BetaGly3 has one active site, hydrolyzes disaccharides, cellodextrins, synthetic substrates and beta-glucosides produced by plants. The enzyme is inhibited by amygdalin, cellotriose, cellotetraose and cellopentaose in high concentrations, probably due to transglycosylation. betaGly3 hydrolyzes beta 1,4-glycosidic linkages with a catalytic rate independent of the substrate polymerization degree (k(int)) of 11.9 s(-1). Its active site is formed by four subsites, where subsites +1 and -1 bind glucose residues with higher affinity than subsite +2. The main role of betaGly3 seems to be disaccharide hydrolysis. BetaGly4 is a beta-galactosidase, since it has highest activity against beta-galactosides. It can also hydrolyze fucosides, but not glucosides, and has Triton X-100 as a non-essential activator (K(a)=15 microM, pH 4.5). betaGly4 has two active sites that can hydrolyze p-nitrophenyl beta-galactoside (NPbetaGal). The one hydrolyzing NPbetaGal with more efficiency is also active against methylumbellipheryl beta-D-galactoside and lactose. The other active site hydrolyzes NPbetaFucoside and binds NPbetaGal weakly. BetaGly4 hydrolyzes hydrophobic substrates with high catalytical efficiency and is able to bind octyl-beta-thiogalactoside in its active site with high affinity. The betaGly4 physiological role is supposed to be the hydrolysis of galactolipids that are found in membranes from vegetal tissues. As the enzyme has a hydrophobic site where Triton X-100 can bind, it might be activated by membrane lipids, thus becoming fully active only at the surface of cell membranes.

MHD Effects of a Ferritic Wall on Tokamak Plasmas

Science.gov (United States)

Hughes, Paul E.

It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency

Tokamak engineering mechanics

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Weiyue; Du, Shijun

2014-01-01

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

Parametric requirements for noncircular tokamak commercial fusion plants

International Nuclear Information System (INIS)

Bourque, R.F.

1978-05-01

Systems analyses have been performed in order to determine the parameter ranges for economical operation of equilibrium commercial fusion power plants using noncircular tokamak reactors. The fully-integrated systems code SCOPE was employed in which costs of electricity were used as the critical dependent variable by which the effects of changes in system parameters were observed. The results of these studies show that many of the operating parameters required for economical operation can be considerably relaxed from what has been thought necessary. For example: (1) Neutron wall loadings over about 2 megawatts per square meter are unnecessary for economical tokamak operation; (2) electrical power levels beyond 1000 MW(e) are not required. In fact, power levels as low as 100 MW(e) may be acceptable; (3) total beta values of 10 to 12 percent are adequate for economic operation; (4) duty cycles as low as 50 percent have only a small effect on plant economics; (5) an acceptable first wall lifetime is 15 MW-yr per square meter

MAST: a Mega Amp Spherical Tokamak

International Nuclear Information System (INIS)

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

1995-01-01

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

Spectroscopy and atomic physics of highly ionized Cr, Fe, and Ni for tokamak plasmas

Science.gov (United States)

Feldman, U.; Doschek, G. A.; Cheng, C.-C.; Bhatia, A. K.

1980-01-01

The paper considers the spectroscopy and atomic physics for some highly ionized Cr, Fe, and Ni ions produced in tokamak plasmas. Forbidden and intersystem wavelengths for Cr and Ni ions are extrapolated and interpolated using the known wavelengths for Fe lines identified in solar-flare plasmas. Tables of transition probabilities for the B I, C I, N I, O I, and F I isoelectronic sequences are presented, and collision strengths and transition probabilities for Cr, Fe, and Ni ions of the Be I sequence are given. Similarities of tokamak and solar spectra are discussed, and it is shown how the atomic data presented may be used to determine ion abundances and electron densities in low-density plasmas.

Beta activity measurements in high, variable gamma backgrounds

International Nuclear Information System (INIS)

Stanga, D.; Sandu, E.; Craciun, L.

1997-01-01

In many cases beta activity measurements must be performed in high and variable gamma backgrounds. In such instances it is necessary to use well-shielded detectors but this technique is limited to laboratory equipment and frequently insufficient. In order to perform in a simple manner beta activity measurements in high and variable backgrounds a software-aided counting technique have been developed and a counting system have been constructed. This technique combines the different counting techniques with traditional method of successive measurement of the sample and background. The counting system is based on a programmable multi-scaler which is endowed with appropriate software and allow all operations to be performed via keyboard in an interactive fashion. Two large - area proportional detectors were selected in order to have the same background and the same gamma response within 5%. A program has been developed for the counting data analysis and beta activity computing. The software-aided counting technique has been implemented for beta activity measurement in high and variable backgrounds. (authors)

Stability analysis of ELMs in long-pulse discharges with ELITE code on EAST tokamak

Science.gov (United States)

Wang, Y. F.; Xu, G. S.; Wan, B. N.; Li, G. Q.; Yan, N.; Li, Y. L.; Wang, H. Q.; Peng, Y.-K. Martin; Xia, T. Y.; Ding, S. Y.; Chen, R.; Yang, Q. Q.; Liu, H. Q.; Zang, Q.; Zhang, T.; Lyu, B.; Xu, J. C.; Feng, W.; Wang, L.; Chen, Y. J.; Luo, Z. P.; Hu, G. H.; Zhang, W.; Shao, L. M.; Ye, Y.; Lan, H.; Chen, L.; Li, J.; Zhao, N.; Wang, Q.; Snyder, P. B.; Liang, Y.; Qian, J. P.; Gong, X. Z.; EAST team

2018-05-01

One challenge in long-pulse and high performance tokamak operation is to control the edge localized modes (ELMs) to reduce the transient heat load on plasma facing components. Minute-scale discharges in H-mode have been achieved repeatedly on Experimental Advanced Superconducting Tokamak (EAST) since the 2016 campaign and understanding the characteristics of the ELMs in these discharges can be helpful for effective ELM control in long-pulse discharges. The kinetic profile diagnostics recently developed on EAST make it possible to perform the pedestal stability analysis quantitatively. Pedestal stability calculation of a typical long-pulse discharge with ELITE code is presented. The ideal linear stability results show that the ELM is dominated by toroidal mode number n around 10–15 and the most unstable mode structure is mainly localized in the steep pressure gradient region, which is consistent with experimental results. Compared with a typical type-I ELM discharge with larger total plasma current (I p = 600 kA), pedestal in the long-pulse H-mode discharge (I p = 450 kA) is more stable in peeling-ballooning instability and its critical peak pressure gradient is evaluated to be 65% of the former. Two important features of EAST tokamak in the long-pulse discharge are presented by comparison with other tokamaks, including a wider pedestal correlated with the poloidal pedestal beta and a smaller inverse aspect ratio and their effects on the pedestal stability are discussed. The effects of uncertainties in measurements on the linear stability results are also analyzed, including the edge electron density profile position, the separatrix position and the line-averaged effective ion charge {Z}{{e}{{f}}{{f}}} value.

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

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

Atomic physics studies of highly charged ions on tokamaks using x-ray spectroscopy

International Nuclear Information System (INIS)

Beiersdorfer, P.; von Goeler, S.; Bitter, M.; Hill, K.W.

1989-07-01

An overview is given of atomic physics issues which have been studied on tokamaks with the help resolution x-ray spectroscopy. The issues include the testing of model calculations predicting the excitation of line radiation, the determination of rate coefficients, and accurate atomic structure measurements. Recent research has focussed primarily on highly charged heliumlike (22 ≤ Z ≤ 28) and neonlike (34 ≤ Z ≤ 63) ions, and results are presented from measurements on the PLT and TFTR tokamaks. Many of the measurements have been aided by improved instrumental design and new measuring techniques. Remarkable agreement has been found between measurements and theory in most cases. However, in this review those areas are stressed where agreement is worst and where further investigations are needed. 19 refs., 13 figs., 2 tabs

MHD activity in the ISX-B tokamak: experimental results and theoretical interpretation

Energy Technology Data Exchange (ETDEWEB)

Carreras, B.A.; Dunlap, J.L.; Bell, J.D.; Charlton, L.A.; Cooper, W.A.; Dory, R.A.; Hender, T.C.; Hicks, H.R.; Holmes, J.A.; Lynch, V.E.

1982-01-01

The observed spectrum of MHD fluctuations in the ISX-B tokamak is clearly dominated by the n=1 mode when the q=1 surface is in the plasma. This fact agrees well with theoretical predictions based on 3-D resistive MHD calculations. They show that the (m=1; n=1) mode is then the dominant instability. It drives other n=1 modes through toroidal coupling and n>1 modes through nonlinear couplings. These theoretically predicted mode structures have been compared in detail with the experimentally measured wave forms (using arrays of soft x-ray detectors). The agreement is excellent. More detailed comparisons between theory and experiment have required careful reconstructions of the ISX-B equilibria. The equilibria so constructed have permitted a precise evaluation of the ideal MHD stability properties of ISX-B. The present results indicate that the high ..beta.. ISX-B equilibria are marginally stable to finite eta ideal MHD modes. The resistive MHD calculations also show that at finite ..beta.. there are unstable resistive pressure driven modes.

Effects of impurities and magnetic divertors on high-temperature tokamaks

International Nuclear Information System (INIS)

Meade, D.M.; Furth, H.P.; Rutherford, P.H.; Seidl, F.G.P.; Duechs, D.F.

1974-10-01

A one-dimensional tokamak plasma transport code has been adapted to include impurity influx, stripping, radiation, and diffusion, as well as the usual processes of hydrogen plasma and heat transport, recycling at the boundary, and multigeneration charge-exchange. Neutral-beam heating, adiabatic compression, and divertor boundary conditions are included as optional features. Illustrative computations are given for present-day and next-generation tokamaks. The problems of impurity control are discussed, and two technical approaches are examined in greater detail: the transient cold-plasma shield, and the poloidal divertor. (auth)

Development of Operation Scenario for Spherical Tokamak at SNU

International Nuclear Information System (INIS)

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

2009-01-01

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

Fusion reactivity, confinement, and stability of neutral-beam heated plasmas in TFTR and other tokamaks

International Nuclear Information System (INIS)

Park, Hyeon, K.

1996-05-01

The hypothesis that the heating beam fueling profile shape connects the edge condition and improved core confinement and fusion reactivity is extensively studied on TFTR and applied to other tokamaks. The derived absolute scalings based on beam fueling profile shape for the stored energy and neutron yield can be applied to the deuterium discharges at different major radii in TFTR. These include Supershot, High poloidal beta, L-mode, and discharges with a reversed shear (RS) magnetic configuration. These scalings are also applied to deuterium-tritium discharges. The role of plasma parameters, such as plasma current, Isdo2(p), edge safety factor, qsdo5(a), and toroidal field, Bsdo2(T), in the performance and stability of the discharges is explicitly studied. Based on practical and externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path for a discharge condition with fusion power gain, Q > 1 is suggested based on this study. Similar physics interpretation is provided for beam heated discharges on other major tokamaks

βp-collapse-induced vertical displacement event in high βp tokamak disruption

International Nuclear Information System (INIS)

Nakamura, Y.; Yoshino, R.; Pomphrey, N.; Jardin, S.C.

1996-01-01

Extremely fast vertical displacement events (VDEs) induced by a strong β p collapse were found in a vertically elongated (κ ∼ 1.5), high β p (β p ∼ 1.7) tokamak with a resistive shell through computer simulations using the tokamak simulation code. Although the plasma current quench which has been shown to be the prime cause of VDEs in a relatively low β p tokamak (β p ∼ 0.2) (Nakamura Y et al 1996 Nucl. Fusion 36 643), was not observed during the VDE evolution, the observed growth rate of VDEs was almost five times (γ ∼ 655 s -1 ) faster than the growth rate of the usual positional instability (γ ∼ 149 s -1 ). The essential mechanism of the β p -collapse-induced VDE was clarified to be the intense enhancement of positional instability due to a large and sudden degradation of the magnetic field decay n-index in addition to the significant destabilization due to a reduction in the stability index n s . The radial shift of the magnetic axis caused by the β p collapse induces eddy currents on the resistive shell, and these eddy currents produce a large degradation of the n-index. (author)

Plasma transport in a compact ignition tokamak

International Nuclear Information System (INIS)

Singer, C.E.; Ku, L.P; Bateman, G.

1987-02-01

Nominal predicted plasma conditions in a compact ignition tokamak are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models which have given almost equally good fits to experimental data. Using a transport model which best fits the data, thermonuclear ignition occurs in a Compact Ignition Tokamak design with major radius 1.32 m, plasma half-width 0.43 m, elongation 2.0, and toroidal field and plasma current ramped in six seconds from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the 3 He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates, have a large effect on ignition and on the maximum beta that can be achieved

Origin of the various beta dependences of ELMy H-mode confinement properties

International Nuclear Information System (INIS)

Takizuka, T; Urano, H; Takenaga, H; Oyama, N

2006-01-01

Dependence of the energy confinement in ELMy H-mode tokamak on the beta has been investigated for a long time, but a common conclusion has not been obtained so far. Recent non-dimensional transport experiments in JT-60U demonstrated clearly the beta degradation. A database for JT-60U ELMy H-mode confinement was assembled. Analysis of this database is carried out, and the strong beta degradation consistent with the above experiments is confirmed. Two subsets of ASDEX Upgrade and JET data in the ITPA H-mode confinement database are analysed to find the origin of the various beta dependences. The shaping of the plasma cross section, as well as the fuelling condition, affects the confinement performance. The beta dependence is not identical for different devices and conditions. The shaping effect, as well as the fuelling effect, is a possible candidate for causing the variation of beta dependence

Measurement of high-energy electrons by means of a Cherenkov detector in ISTTOK tokamak

Energy Technology Data Exchange (ETDEWEB)

Jakubowski, L., E-mail: lech.Jjakubowski@ipj.gov.p [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Zebrowski, J. [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Plyusnin, V.V. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049 - 001 Lisboa (Portugal); Malinowski, K.; Sadowski, M.J.; Rabinski, M. [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Fernandes, H.; Silva, C.; Duarte, P. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049 - 001 Lisboa (Portugal)

2010-10-15

The paper concerns detectors of the Cherenkov radiation which can be used to measure high-energy electrons escaping from short-living plasma. Such detectors have high temporal (about 1 ns) and spatial (about 1 mm) resolution. The paper describes a Cherenkov-type detector which was designed, manufactured and installed in the ISTTOK tokamak in order to measure fast runaway electrons. The radiator of that detector was made of an aluminium nitride (AlN) tablet with a light-tight filter on its front surface. Cherenkov signals from the radiator were transmitted through an optical cable to a fast photomultiplier. It made possible to perform direct measurements of the runaway electrons of energy above 80 keV. The measured energy values and spatial characteristics of the recorded electrons appeared to be consistent with results of numerical modelling of the runaway electron generation process in the ISTTOK tokamak.

Safety assessment document (SAD) for the Princeton Beta Experiment Modification (PBX-M)

International Nuclear Information System (INIS)

Stencel, J.R.; Parsells, R.F.

1988-04-01

The Princeton Beta Experiment-Modification (PBX-M) is an experimental device of the tokamak type. A tokamak is characterized by a strong toroidal magnetic field composed of an externally driven component parallel to the torus centerline modified by the field produced by a transformer-driven current (OH) in the confined plasma. A second magnetic field parallel to the major toroidal axis is added to provide radial equilibrium for the plasma. As an advanced tokamak, PBX-M will have additional magnetic fields to reshape the plasma cross section from a circle into a kidney bean shape; it will also be equipped with 6MW or more of auxiliary heating power provided by four neutral beam injectors, with RF systems, and with an extensive set of diagnostics. Potential hazards associated with PBX-M, which are analyzed in this report, result from energy stored in the magnetic fields, high voltages necessary for the operation of some of the equipment and diagnostics, neutron radiation when the neutral beams are run with deuterium and x-rays, especially those emitted as a result of plasma-wall interaction. This report satisfies the requirements set forth in the PPPL Health and Safety Directives, specifically HSD-5003, and in DOE Order 5481.1B and its Chicago operations supplement (DOE86, DOE82)

The stability margin on EAST tokamak

International Nuclear Information System (INIS)

Jin-Ping, Qian; Bao-Nian, Wan; Biao, Shen; Bing-Jia, Xiao; Walker, M.L.; Humphreys, D.A.

2009-01-01

The experimental advanced superconducting tokamak (EAST) is the first full superconducting tokamak with a D-shaped cross-sectional plasma presently in operation. Its poloidal coils are relatively far from the plasma due to the necessary thermal isolation from the superconducting magnets, which leads to relatively weaker coupling between plasma and poloidal field. This may cause more difficulties in controlling the vertical instability by using the poloidal coils. The measured growth rates of vertical stability are compared with theoretical calculations, based on a rigid plasma model. Poloidal beta and internal inductance are varied to investigate their effects on the stability margin by changing the values of parameters α n and γ n (Howl et al 1992 Phys. Fluids B 4 1724), with plasma shape fixed to be a configuration with k = 1.9 and δ = 0.5. A number of ways of studying the stability margin are investigated. Among them, changing the values of parameters κ and l i is shown to be the most effective way to increase the stability margin. Finally, a guideline of stability margin M s (κ, l i , A) to a new discharge scenario showing whether plasmas can be stabilized is also presented in this paper

Joint research using small tokamaks

International Nuclear Information System (INIS)

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

2005-01-01

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

Optimization design for SST-1 Tokamak insulators

International Nuclear Information System (INIS)

Zhang Yuanbin; Pan Wanjiang

2012-01-01

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

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

Joint research using small tokamaks

International Nuclear Information System (INIS)

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

2005-01-01

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

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.)

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)

Tokamak devices: towards controlled fusion

International Nuclear Information System (INIS)

Trocheris, M.

1975-01-01

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

High-energy ion tail formation due to ion acoustic turbulence in the TRIAM-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Kazuo; Hiraki, Naoji; Nakamura, Yukio; Itoh, Satoshi [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

1982-02-01

The two-component ion energy spectra observed in the TRIAM-1 tokamak are explained as a result of the high-energy ion tail formation due to ion acoustic turbulence driven by a toroidal current pulse for turbulent heating.

Effects of global MHD instability on operational high beta-regime in LHD

International Nuclear Information System (INIS)

Watanabe, K.Y.; Sakakibara, S.; Narushima, Y.; Funaba, H.; Narihara, K.; Tanaka, K.; Toi, K.; Ohdachi, S.; Kaneko, O.; Yamada, H.; Nakajima, N.; Yamada, I.; Kawahata, K.; Tokuzawa, T.; Komori, A.; Yamaguchi, T.; Suzuki, Y.; Cooper, W.A.; Murakami, S.

2005-01-01

In the Large Helical device (LHD), the operational highest averaged beta value has been expanded from 3.2% to 4% in last two years by increasing the heating capability and exploring a new magnetic configuration with a high aspect ratio. Although the MHD stability properties are considered to be unfavourable in the new high aspect configuration, the heating efficiency due to neutral beams and the transport properties are expected to be favourable in a high beta range. In order to make clear the effect of the global ideal MHD unstable mode on the operational regimes in helical systems, specially the beta gradients in the peripheral region and the beta value, the MHD analysis and the transport analysis are done in a high beta range up to 4% in LHD. In a high beta range of more than 3%, the maxima of the observed thermal pressure gradients in the peripheral region are marginally stable to a global ideal MHD instability. Though a gradual degradation of the local transport in the region has been observed as beta increases, a disruptive degradation of the local transport does not appear in the beta range up to 4%. (author)

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

Stabilization of external kink modes in a tokamak with rotating plasma

International Nuclear Information System (INIS)

Mikhailovskii, A.B.; Kuvshinov, B.N.

1995-01-01

An analytical theory of stabilization of external kink modes in a tokamak with rotating plasma is developed, which is of interest in connection with experiments on the DIII-D tokamak demonstrating such a stabilization. It is assumed that, in addition to the main poloidal harmonic, the mode includes one or more side-band poloidal harmonics with singular points lying inside the plasma. Near these singular points, plasma inertia and related toroidal effects, the compressible part of plasma pressure and longitudinal viscosity, are allowed for. These effects are described kinetically taking into account the toroidal trapping of the resonant ions, which is essential if the toroidal velocity is small compared to the ion thermal velocity. Thereby, the theory presented includes both ion Landau damping and its weakening due to toroidal trapping. Near the singular points high-beta effects, which result in the finiteness of the Mercier index s, are allowed for. It is shown that the influence of plasma rotation on the external kink modes is most significant in the case of s<0, i.e., when the development of the instability in a non-rotating plasma is most highly favored. In this case, the plasma rotation plays a stabilizing role, even when the ion Landau damping is neglected. The analysis presented also confirms the hypothesis of Bondeson and Ward on the stabilizing effect of ion Landau damping if this damping is not too small

Modelling of prompt losses of high energy charged particles in Tokamaks

International Nuclear Information System (INIS)

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

1990-01-01

A simple analytical expression for the total prompt loss fraction of high energy charged particles in an axisymmetric Tokamak is derived. The results are compared with predictions obtained from numerical simulations and show good agreement. An application is made to sawtooth induced changes in the losses of fusion generated high energy charged particles. Particular emphasis is given to the importance of sawtooth induced profile changes of the background ion densities and temperature as well as to redistribution of particles which have accumulated during the sawtooth rise but are being lost by redistribution at the sawtooth crash. (au)

Characterizing electrostatic turbulence in tokamak plasmas with high MHD activity

Energy Technology Data Exchange (ETDEWEB)

Guimaraes-Filho, Z O; Santos Lima, G Z dos; Caldas, I L; Nascimento, I C; Kuznetsov, Yu K [Instituto de Fisica, Universidade de Sao Paulo, Caixa Postal 66316, 05315-970, Sao Paulo, SP (Brazil); Viana, R L, E-mail: viana@fisica.ufpr.b [Departamento de Fisica, Universidade Federal do Parana, Caixa Postal 19044, 81531-990, Curitiba, PR (Brazil)

2010-09-01

One of the challenges in obtaining long lasting magnetic confinement of fusion plasmas in tokamaks is to control electrostatic turbulence near the vessel wall. A necessary step towards achieving this goal is to characterize the turbulence level and so as to quantify its effect on the transport of energy and particles of the plasma. In this paper we present experimental results on the characterization of electrostatic turbulence in Tokamak Chauffage Alfven Bresilien (TCABR), operating in the Institute of Physics of University of Sao Paulo, Brazil. In particular, we investigate the effect of certain magnetic field fluctuations, due to magnetohydrodynamical (MHD) instabilities activity, on the spectral properties of electrostatic turbulence at plasma edge. In some TCABR discharges we observe that this MHD activity may increase spontaneously, following changes in the edge safety factor, or after changes in the radial electric field achieved by electrode biasing. During the high MHD activity, the magnetic oscillations and the plasma edge electrostatic turbulence present several common linear spectral features with a noticeable dominant peak in the same frequency. In this article, dynamical analyses were applied to find other alterations on turbulence characteristics due to the MHD activity and turbulence enhancement. A recurrence quantification analysis shows that the turbulence determinism radial profile is substantially changed, becoming more radially uniform, during the high MHD activity. Moreover, the bicoherence spectra of these two kinds of fluctuations are similar and present high bicoherence levels associated with the MHD frequency. In contrast with the bicoherence spectral changes, that are radially localized at the plasma edge, the turbulence recurrence is broadly altered at the plasma edge and the scrape-off layer.

Excitation of Alfvenic instabilities in spherical tokamaks

International Nuclear Information System (INIS)

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

2003-01-01

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

High #betta# and toroidal effects on the internal kink mode in tokamaks

International Nuclear Information System (INIS)

Schmalz, R.

1982-09-01

The inclusion of high-#betta# and first-order toroidal terms in the reduced set of (resistive) MHD equations affords the possibility of improving the study of tokamak plasma behaviour by three-dimensional numerical simulation. A new code, GALA, based on the reduced equations is developed. It is used to analyse the linear and nonlinear behaviour of the internal kink mode in equilibria which are generated by a simple relaxation procedure. We find that the inclusion of toroidal effects in high-#betta# equilibria provides considerable stabilization. (orig.)

Simulation study on dynamics of runaways in tokamaks

International Nuclear Information System (INIS)

Liu Jian; Qin Hong; Fisch, Nathaniel J.

2014-01-01

Electrons with high velocities can be accelerated to very high energies by a strong electric field to form runaway electrons. In tokamak, runaway electrons are produced in many different processes, including the acceleration from the high-energy tail of thermal distribution, through the runaway avalanche, during the rf wave heating and other non-Ohmic current drive, and even in the magnetic reconnection. This proceeding focus on different dynamical problems of runaway electrons in tokamaks. (author)

Structured Cable for High-Current Coils of Tokamaks

Science.gov (United States)

Benson, Christopher; McIntyre, Peter; Sattarov, Akhdiyor; Mann, Thomas

2011-10-01

The 45 kA superconducting cable for the ITER central solenoid coil has yielded questionable results in two recent tests. In both cases the cable Tc increased after cycling only a fraction of the design life, indicating degradation due to fatigue and fracture among the superconducting strands. The Accelerator Research Lab at Texas A&M University is developing a design for a Nb3Sn structured cable suitable for such tokamak coils. The superconductor is configured in 6 sub-cables, and each subcable is supported within a channel of a central support structure within a high-strength armor sheath. The structured cable addresses two issues that are thought to compromise opposition at high current. The strands are supported without cross-overs (which produce stress concentration); and armor sheath and core structure bypass stress through the coil and among subcables so that the stress within each subcable is only what is produced directly upon it. Details of the design and plans for development will be presented.

Numerical models for high beta magnetohydrodynamic flow

International Nuclear Information System (INIS)

Brackbill, J.U.

1987-01-01

The fundamentals of numerical magnetohydrodynamics for highly conducting, high-beta plasmas are outlined. The discussions emphasize the physical properties of the flow, and how elementary concepts in numerical analysis can be applied to the construction of finite difference approximations that capture these features. The linear and nonlinear stability of explicit and implicit differencing in time is examined, the origin and effect of numerical diffusion in the calculation of convective transport is described, and a technique for maintaining solenoidality in the magnetic field is developed. Many of the points are illustrated by numerical examples. The techniques described are applicable to the time-dependent, high-beta flows normally encountered in magnetically confined plasmas, plasma switches, and space and astrophysical plasmas. 40 refs

Developing physics basis for the snowflake divertor in the DIII-D tokamak

Science.gov (United States)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

2018-03-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard

Threshold condition for nonlinear tearing modes in tokamaks

International Nuclear Information System (INIS)

Zabiego, M.F.; Callen, J.D.

1996-03-01

Low-mode-number tearing, mode nonlinear evolution is analyzed emphasizing the need for a threshold condition, to account for observations in tokamaks. The discussion is illustrated by two models recently introduced in the literature. The models can be compared with the available data and/or serve as a basis for planning some experiments in order to either test theory (by means of beta-limit scaling laws, as proposed in this paper) or attempt to control undesirable tearing modes. Introducing a threshold condition in the tearing mode stability analysis is found to reveal some bifurcation points and thus domains of intrinsic stability in the island dynamics operational space

Gyrokinetic theory and dynamics of the tokamak edge

Energy Technology Data Exchange (ETDEWEB)

Scott, B. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

2016-08-15

The validity of modern gyrokinetic field theory is assessed for the tokamak edge. The basic structure of the Lagrangian and resulting equations and their conservation laws is reviewed. The conventional microturbulence ordering for expansion is small potential/arbitrary wavelength. The equilibrium ordering for expansion is long wavelength/arbitrary amplitude. The long-wavelength form of the conventional Lagrangian is derived in detail. The two Lagrangians are shown to match at long wavelength if the E x B Mach number is small enough for its corrections to the gyroaveraging to be neglected. Therefore, the conventional derivation and its Lagrangian can be used at all wavelengths if these conditions are satisfied. Additionally, dynamical compressibility of the magnetic field can be neglected if the plasma beta is small. This allows general use of a shear-Alfven Lagrangian for edge turbulence and self consistent equilibrium-scale phenomena for flows, currents, and heat fluxes for conventional tokamaks without further modification by higher-order terms. Corrections in polarisation and toroidal angular momentum transport due to these higher-order terms for global edge turbulence computations are shown to be small. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

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

Energy composition of high-energy neutral beams on the COMPASS tokamak

Directory of Open Access Journals (Sweden)

Mitosinkova Klara

2016-12-01

Full Text Available The COMPASS tokamak is equipped with two identical neutral beam injectors (NBI for additional plasma heating. They provide a beam of deuterium atoms with a power of up to ~(2 × 300 kW. We show that the neutral beam is not monoenergetic but contains several energy components. An accurate knowledge of the neutral beam power in each individual energy component is essential for a detailed description of the beam- -plasma interaction and better understanding of the NBI heating processes in the COMPASS tokamak. This paper describes the determination of individual energy components in the neutral beam from intensities of the Doppler-shifted Dα lines, which are measured by a high-resolution spectrometer viewing the neutral beam-line at the exit of NBI. Furthermore, the divergence of beamlets escaping single aperture of the last accelerating grid is deduced from the width of the Doppler-shifted lines. Recently, one of the NBI systems was modified by the removal of the Faraday copper shield from the ion source. The comparison of the beam composition and the beamlet divergence before and after this modification is also presented.

Long-pulse high-performance discharges in the DIII-D tokamak

International Nuclear Information System (INIS)

Luce, T.C.; Wade, M.R.; Politzer, P.A.

2001-01-01

Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance ∼10 has been sustained for >5τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and β≤5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with β N H 89 ∼7 for up to 6.3 s or ∼34 τ E . These discharges appear to be in resistive equilibrium with q min ∼1.05, in agreement with the current profile relaxation time of 1.8 s. (author)

LONG-PULSE, HIGH-PERFORMANCE DISCHARGES IN THE DIII-D TOKAMAK

International Nuclear Information System (INIS)

T.C. LUCE; M.R. WADE; P.A. POLITZER; S.L. ALLEN; M E. AUSTIN; D.R. BAKER; B.D. BRAY; D.P. BRENNAN; K.H. BURRELL; T.A. CASPER; M.S. CHU; J.D. De BOO; E.J. DOYLE; J.R. FERRON; A.M. GAROFALO; P.GOHIL; I.A. GORELOV; C.M. GREENFIELD; R.J. GROEBNER; W.W. HEIBRINK; C.-L. HSIEH; A.W. HYATT; R.JAYAKUMAR; J.E.KINSEY; R.J. LA HAYE; L.L. LAO; C.J. LASNIER; E.A. LAZARUS; A.W. LEONARD; Y.R. LIN-LIU; J. LOHR; M.A. MAKOWSKI; M. MURAKAMI; C.C. PETTY; R.I. PINSKER; R. PRATER; C.L. RETTIG; T.L. RHODES; B.W. RICE; E.J. STRAIT; T.S. TAYLOR; D.M. THOMAS; A.D. TURNBULL; J.G. WATKINS; W.P.WEST; K.-L. WONG

2000-01-01

Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance ∼10 has been sustained for >5 τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and β ∼(le) 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate ∼75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with β N H 89 ∼ 7 for up to 6.3 s or ∼ 34 τ E . These discharges appear to be in resistive equilibrium with q min ∼ 1.05, in agreement with the current profile relaxation time of 1.8 s

Minimum dimension of an ITER like Tokamak with a given Q

Energy Technology Data Exchange (ETDEWEB)

Johner, J

2004-07-01

The minimum dimension of an ITER like tokamak with a given amplification factor Q is calculated for two values of the maximum magnetic field in the superconducting toroidal field coils. For ITERH-98P(y,2) scaling of the energy confinement time, it is shown that for a sufficiently large tokamak, the maximum Q is obtained for the operating point situated both at the maximum density and at the minimum margin with respect to the H-L transition. We have shown that increasing the maximum magnetic field in the toroidal field coils from the present 11.8 T to 16 T would result in a strong reduction of the machine size but has practically no effect on the fusion power. Values obtained for {beta}{sub N} are found to be below 2. Peak fluxes on the divertor plates with an ITER like divertor and a multi-machine expression for the power radiated in the plasma mantle, are below 10 MW/m{sup 2}.

Prospects for Tokamak Fusion Reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.

1995-01-01

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

Study on assembly techniques and procedures for ITER tokamak device

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi; Ue, Koichi; Shimizu, Katsusuke; Onozuka, Masanori

2006-06-01

The International Thermonuclear Experimental Reactor (ITER) tokamak is mainly composed of a doughnut-shaped vacuum vessel (VV), four types of superconducting coils such as toroidal field coils (TF coils) arranged around the VV, and in-vessel components, such as blanket and divertor. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of the VV and the TF coil are required to be a high accuracy of ±3 mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements as well as the configuration of the tokamak with large size and heavy weight. Based on the above backgrounds, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The tokamak assembly operations are categorized into six work break down structures (WBS), i.e., (1) preparation for assembly operations, (2) sub-assembly of the 40deg sector composed of 40deg VV sector, two TF coils and thermal shield between VV and TF coil at the assembly hall, (3) completion of the doughnut-shaped tokamak assembly composed of nine 40deg sectors in the cryostat at the tokamak pit, (4) measurement of positioning and accuracy after the completion of the tokamak assembly, (5) installation of the ex-vessel components, and (6) installation of in-vessel components. In the present report, two assembly operations of (2) and (3) in the above six WBS, which are the most critical in the tokamak assembly, are mainly described. The report describes the following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology

Fast computational scheme for feedback control of high current fusion tokamaks

International Nuclear Information System (INIS)

Dong, J.Q.; Khayrutdinov, R.; Azizov, E.; Jardin, S.

1992-01-01

An accurate and fast numerical model of tokamak plasma evolution is presented. In this code (DINA) the equilibrium problem of plasmas with free boundaries in externally changing magnetic fields is solved simultaneously with the plasma transport equation. The circuit equations are solved for the vacuum vessel and passive and active coils. The code includes pellet injection, neutral beam heating, auxiliary heating, and alpha particle heating. Bootstrap and beam-driven plasma currents are accounted for. An inverse variable technique is utilized to obtain the coordinates of the equilibrium magnetic surfaces. This numerical algorithm permits to determine the flux coordinates very quickly and accurately. The authors show that using the fully resistive MHD analysis the region of stability (to vertical motions) is wider than using the rigid displacement model. Comparing plasma motions with the same gain, it is seen that the plasma oscillates more in the rigid analysis than in the MHD analysis. They study the influence of the pick up coil's location and the possibility of control of the plasma vertical position. They use a simple modification of the standard control law that enables the control of the plasma with pick up coils located at any position. This flexibility becomes critical in the design of future complex high current tokamak systems. The fully resistive MHD model permits to obtain accurate estimates of the plasma response. This approach yields computational time savings of one to two orders of magnitude with respect to other existing MHD models. In this sense, conventional numerical algorithms do not provide suitable models for application of modern control techniques into real time expert systems. The proposed inverse variable technique is rather suitable for incorporation in a comprehensive expert system for feedback control of fusion tokamaks in real time

Spherical tokamak power plant design issues

International Nuclear Information System (INIS)

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

2000-01-01

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

Electrons of high perpendicular energy in the low-density regime of Tokamaks

International Nuclear Information System (INIS)

Bornatici, M.; Engelmann, F.

1978-01-01

Effects due to instabilities excited in the low-density regime of tokamaks by runaway electrons via the cyclotron resonance ω+Ω=kV along with the formation of a positive slope in the runaway distribution are considered. Conditions for the production of electrons of high perpendicular energy and their trapping in toroidal field ripples, leading to liner damage, are discussed and found to be rather stringent. Fairly good agreement with the experiments is found

Hard X-ray studies on the Castor tokamak

International Nuclear Information System (INIS)

Mlynar, J.

1990-04-01

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

High energy beta rays and vectors of Bilharzia and Fasciola

International Nuclear Information System (INIS)

Fletcher, J.J.; Akpa, T.C.; Dim, L.A.; Ogunsusi, R.

1988-01-01

Preliminary investigations of the effects of high energy beta rays on Lymnea natalensis, the snail vector of Schistosoma haematobium have been conducted. Results show that in both stream and tap water, about 70% of the snails die when irradiated for up to 18 hours using a 15m Ci Sr-90 beta source. The rest of the snails die without further irradiation in 24 hours. It may then be possible to control the vectors of Bilharzia and Fasciola by using both the direct and indirect effects of high energy betas. (author)

High energy beta rays and vectors of Bilharzia and Fasciola

Energy Technology Data Exchange (ETDEWEB)

Fletcher, J.J.; Akpa, T.C.; Dim, L.A.; Ogunsusi, R.

1988-01-01

Preliminary investigations of the effects of high energy beta rays on Lymnea natalensis, the snail vector of Schistosoma haematobium have been conducted. Results show that in both stream and tap water, about 70% of the snails die when irradiated for up to 18 hours using a 15m Ci Sr-90 beta source. The rest of the snails die without further irradiation in 24 hours. It may then be possible to control the vectors of Bilharzia and Fasciola by using both the direct and indirect effects of high energy betas.

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

Tokamak engineering mechanics

CERN Document Server

Song, Yuntao; Du, Shijun

2013-01-01

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

The tokamak as a neutron source

International Nuclear Information System (INIS)

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

1989-11-01

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

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

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

COMPREHENSIVE GYROKINETIC SIMULATION OF TOKAMAK TURBULENCE AT FINITE RELATIVE GYRORADIUS

International Nuclear Information System (INIS)

WALTZ, R.E.; CANDY, J.; ROSENBLUTH, M.N.

2002-01-01

OAK B202 COMPREHENSIVE GYROKINETIC SIMULATION OF TOKAMAK TURBULENCE AT FINITE RELATIVE GYRORADIUS. A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate turbulent transport in actual experimental profiles and allow direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite beta, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius (ρ*) so as to treat the profile shear stabilization effects which break gyroBohm scaling. The code operates in a cyclic flux tube limit which allows only gyroBohm scaling and a noncyclic radial annulus with physical profile variation. The later requires an adaptive source to maintain equilibrium profiles. Simple ITG simulations demonstrate the broken gyroBohm scaling depends on the actual rotational velocity shear rates competing with mode growth rates, direct comprehensive simulations of the DIII-D ρ*-scaled L-mode experiments are presented as a quantitative test of gyrokinetics and the paradigm

Progress on advanced tokamak and steady-state scenario development on DIII-D and NSTX

Energy Technology Data Exchange (ETDEWEB)

Doyle, E J [Department of Electrical Engineering and PSTI, University of California, Los Angeles, California 90095 (United States); Garofalo, A M [Columbia University, New York, New York 10027 (United States); Greenfield, C M [General Atomics, San Diego, California 92186-5608 (United States); Kaye, S M [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Menard, J E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Murakami, M [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Sabbagh, S A [Columbia University, New York, New York 10027 (United States); Austin, M E [University of Texas-Austin, Austin, Texas 78712 (United States); Bell, R E [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Burrell, K H [General Atomics, San Diego, California 92186-5608 (United States); Ferron, J R [General Atomics, San Diego, California 92186-5608 (United States); Gates, D A [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Groebner, R J; Hyatt, A W; Luce, T C; Petty, C C; Wade, M R; Waltz, R E [General Atomics, San Diego, California 92186-5608 (United States); Jayakumar, R J [Lawrence Livermore National Lab., Livermore, California 94550 (United States); Kinsey, J E [Lehigh Univ., Bethlehem, Pennsylvania 18015 (United States); LeBlanc, B P [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); McKee, G R [Univ. of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Okabayashi, M [Princeton Plasma Physics Lab., Princeton, New Jersey 08543-0451 (United States); Peng, Y-K M [Oak Ridge National Lab., Oak Ridge, Tennessee 37831 (United States); Politzer, P A [General Atomics, San Diego, California 92186-5608 (United States); Rhodes, T L [Dept. of Electrical Engineering and PSTI, Univ. of California, Los Angeles, California 90095 (United States)

2006-12-15

Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f{sub BS} {>=} 60%, q{sub 95} {approx} 4-5 and G {identical_to}{beta}{sub N}H{sub scaling}/q{sub 95}{sup 2} {>=}0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E ? B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at {beta}{sub N} {>=} 4 with q{sub min} {>=} 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. {beta}{sub N} = 4, H{sub 89} = 2.5, with f{sub BS} > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q {>=} 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

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

Preliminary Study of Ideal Operational MHD Beta Limit in HL-2A Tokamak Plasmas

International Nuclear Information System (INIS)

Shen Yong; Dong Jiaqi; He Hongda; Turnbull, A. D.

2009-01-01

Magnetohydrodynamic (MHD) n = 1 kink mode with n the toroidal mode number is studied and the operational beta limit, constrained by the mode, is calculated for the equilibrium of HL-2A by using the GATO code. Approximately the same beta limit is obtained for configurations with a value of the axial safety factor q 0 both larger and less than 1. Without the stabilization of the conducting wall, the beta limit is found to be 0.821% corresponding to a normalized beta value of β c N = 2.56 for a typical HL-2A discharge with a plasma current I p = 0.245 MA, and the scaling of β c N ∼constant is confirmed. (magnetically confined plasma)

High-pressure duo-multichannel soft x-ray spectrometer for tokamak plasma diagnostics

International Nuclear Information System (INIS)

Schwob, J.L.; Wouters, A.W.; Suckewer, S.

1987-03-01

A high-resolution, time-resolving soft X-ray multichannel spectrometer (SOXMOS) that permits the simultaneous measurement of emission in two different spectral ranges has been developed and tested extensively for tokamak plasma diagnostics. The basic instrument is a high-resolution, interferometrically adjusted, extreme grazing incidence Schwob-Fraenkel duochromator. The instrument is equipped with two multichannel detectors that are adjusted interferometrically and scan along the Rowland circle. Each consists of an MgF 2 coated, funneled microchannel plate, associated with a phosphor screen image intensifier that is coupled to a 1024-element photodiode array by a flexible fibrer optic conduit. The total wavelength coverage of the instrument is 5 to 340 0 A with a measured resolution (FWHM) of about 0.2 A when equipped with a 600 g/mm grating, and 5 to 85 A with a resolution of about 0.06 A using a 2400 g/mm grating. The simultaneous spectral coverage of each detector varies from 15 A at the short wavelength limit to 70 A at the long wavelength limit with the lower dispersion grating. The minimum read-out time for a full spectral portion is 17 ms, but several individual lines can be measured with 1 ms time resolution by selected pixel readout. Higher time resolution can be achieved by replacing one multichannel detector with a single channel electron multiplier detector. Examples of data from the PLT and TFTR tokamaks are presented to illustrate the instrument's versatility, high spectral resolution, and high signal-to-noise ratio even in the 10 A region. 44 refs., 20 figs

Prospects for steady-state tokamak reactor operation through feedback control of the current density profile

Energy Technology Data Exchange (ETDEWEB)

Moreau, D

1994-12-31

A brief overview of the most relevant experiments on current profile modifications, strong improvements with respect to the usual L-mode scaling laws and Troyon beta limit is presented, as relevant issues for most tokamaks. Practical means and scenarios for producing and maintaining the optimum current profiles in the various phases of the thermonuclear discharge (profile formation, current ramp-up, burn phase) are proposed. (author). 34 refs., 3 figs.

Improvement of tokamak performance by injection of electrons

International Nuclear Information System (INIS)

Ono, Masayuki.

1992-12-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Compact fusion energy based on the spherical tokamak

Science.gov (United States)

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

2018-01-01

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

A distributed high speed data acquisition system for KT5C tokamak

International Nuclear Information System (INIS)

Sun Xiang; Wang Zhijiang; Lu Ronghua; Wang Jun; Yu Yi; Zhu Zhenghua; Wen Yizhi; Wan Shude; Liu Wandong; Yu Changxuan

2005-01-01

The development of a distributed data acquisition system with low cost to implement high speed data collection through the campus networks for a small tokamak, KT5C, is presented. Data of 512 k bytes at 5 MHz from 5 channels for each can be collected during about 10s after three researchers at different positions demand this system for acquisitions. This system realizes long distance multiuser operations; virtually efficiency of the data acquisition is enhanced. (authors)

Characteristics of MHD stability of high beta plasmas in LHD

International Nuclear Information System (INIS)

Sato, M.; Nakajima, N.; Watanabe, K.Y.; Todo, Y.; Suzuki, Y.

2012-11-01

In order to understand characteristics of the MHD stability of high beta plasmas obtained in the LHD experiments, full MHD simulations have been performed for the first time. Since there is a magnetic hill in a plasma peripheral region, the ballooning modes extending into the plasma peripheral region with a chaotic magnetic field are destabilized. However, in the nonlinear phase, the core region comes under the in influence of the instabilities and the central pressure decreases. There is a tendency that modes are suppressed as the beta value and/or magnetic Reynolds number increase, which is consistent with a result that high beta plasmas enter the second stable region of the ideal ballooning modes as beta increases and remaining destabilized ballooning modes are considered to be resistive type. (author)

Osteocalcin protects pancreatic beta cell function and survival under high glucose conditions

Energy Technology Data Exchange (ETDEWEB)

Kover, Karen, E-mail: kkover@cmh.edu [Division of Endocrine/Diabetes, Children' s Mercy Hospital & Clinics, Kansas City, MO 64108 (United States); University of Missouri-Kansas City School of Medicine, Kansas City, MO 64108 (United States); Yan, Yun; Tong, Pei Ying; Watkins, Dara; Li, Xiaoyu [Division of Endocrine/Diabetes, Children' s Mercy Hospital & Clinics, Kansas City, MO 64108 (United States); University of Missouri-Kansas City School of Medicine, Kansas City, MO 64108 (United States); Tasch, James; Hager, Melissa [Kansas City University Medical Biosciences, Kansas City, MO (United States); Clements, Mark; Moore, Wayne V. [Division of Endocrine/Diabetes, Children' s Mercy Hospital & Clinics, Kansas City, MO 64108 (United States); University of Missouri-Kansas City School of Medicine, Kansas City, MO 64108 (United States)

2015-06-19

Diabetes is characterized by progressive beta cell dysfunction and loss due in part to oxidative stress that occurs from gluco/lipotoxicity. Treatments that directly protect beta cell function and survival in the diabetic milieu are of particular interest. A growing body of evidence suggests that osteocalcin, an abundant non-collagenous protein of bone, supports beta cell function and proliferation. Based on previous gene expression data by microarray, we hypothesized that osteocalcin protects beta cells from glucose-induced oxidative stress. To test our hypothesis we cultured isolated rat islets and INS-1E cells in the presence of normal, high, or high glucose ± osteocalcin for up to 72 h. Oxidative stress and viability/mitochondrial function were measured by H{sub 2}O{sub 2} assay and Alamar Blue assay, respectively. Caspase 3/7 activity was also measured as a marker of apoptosis. A functional test, glucose stimulated insulin release, was conducted and expression of genes/protein was measured by qRT-PCR/western blot/ELISA. Osteocalcin treatment significantly reduced high glucose-induced H{sub 2}O{sub 2} levels while maintaining viability/mitochondrial function. Osteocalcin also significantly improved glucose stimulated insulin secretion and insulin content in rat islets after 48 h of high glucose exposure compared to untreated islets. As expected sustained high glucose down-regulated gene/protein expression of INS1 and BCL2 while increasing TXNIP expression. Interestingly, osteocalcin treatment reversed the effects of high glucose on gene/protein expression. We conclude that osteocalcin can protect beta cells from the negative effects of glucose-induced oxidative stress, in part, by reducing TXNIP expression, thereby preserving beta cell function and survival. - Highlights: • Osteocalcin reduces glucose-induced oxidative stress in beta cells. • Osteocalcin preserves beta cell function and survival under stress conditions. • Osteocalcin reduces glucose

Osteocalcin protects pancreatic beta cell function and survival under high glucose conditions

International Nuclear Information System (INIS)

Kover, Karen; Yan, Yun; Tong, Pei Ying; Watkins, Dara; Li, Xiaoyu; Tasch, James; Hager, Melissa; Clements, Mark; Moore, Wayne V.

2015-01-01

Diabetes is characterized by progressive beta cell dysfunction and loss due in part to oxidative stress that occurs from gluco/lipotoxicity. Treatments that directly protect beta cell function and survival in the diabetic milieu are of particular interest. A growing body of evidence suggests that osteocalcin, an abundant non-collagenous protein of bone, supports beta cell function and proliferation. Based on previous gene expression data by microarray, we hypothesized that osteocalcin protects beta cells from glucose-induced oxidative stress. To test our hypothesis we cultured isolated rat islets and INS-1E cells in the presence of normal, high, or high glucose ± osteocalcin for up to 72 h. Oxidative stress and viability/mitochondrial function were measured by H 2 O 2 assay and Alamar Blue assay, respectively. Caspase 3/7 activity was also measured as a marker of apoptosis. A functional test, glucose stimulated insulin release, was conducted and expression of genes/protein was measured by qRT-PCR/western blot/ELISA. Osteocalcin treatment significantly reduced high glucose-induced H 2 O 2 levels while maintaining viability/mitochondrial function. Osteocalcin also significantly improved glucose stimulated insulin secretion and insulin content in rat islets after 48 h of high glucose exposure compared to untreated islets. As expected sustained high glucose down-regulated gene/protein expression of INS1 and BCL2 while increasing TXNIP expression. Interestingly, osteocalcin treatment reversed the effects of high glucose on gene/protein expression. We conclude that osteocalcin can protect beta cells from the negative effects of glucose-induced oxidative stress, in part, by reducing TXNIP expression, thereby preserving beta cell function and survival. - Highlights: • Osteocalcin reduces glucose-induced oxidative stress in beta cells. • Osteocalcin preserves beta cell function and survival under stress conditions. • Osteocalcin reduces glucose-induced TXNIP

Experimental study on the practicability of a self-created spherical tokamak in the coil less STPC-EX machine

International Nuclear Information System (INIS)

Sinman, S.; Sinman, A.

2003-01-01

The aim of this study is to identify the physical basis of the alternative self-organization mechanism that exists on the STPC-EX machine and to determine complementary features with respect to present compact toroid concepts. In the STPC-EX machine, there exist two solenoids placed inside the central passive floating conductive hollow rod and externally onto flux conserver. These are in a passive state and they do not have an important role in the self-created spherical tokamak plasma (SCSTP) in the STPC-EX machine. In this study, conventional diagnostic tools are used and for photographic recording, the method of open shutter integrated post-fogging is chosen. Two annular coaxial plasma current sheets, one within the other in the same direction, are created and flow on the surface of the central conductive hollow rod. Consequently, the spherical tokamak is configured by a new creation mechanism called the dual-axial z-pinch. High betas of 0.4-0.6 and aspect ratios of up to 1.25 can be obtained. (author)

Demonstration of real-time control for poloidal beta in KSTAR

Energy Technology Data Exchange (ETDEWEB)

Han, Hyunsun, E-mail: hyunsun@nfri.re.kr [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Hahn, S.H.; Bak, J.G. [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of); Hyatt, A.; Johnson, R. [General Atomics, San Diego, CA (United States); Woo, M.H.; Kim, J.S.; Bae, Y.S. [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of)

2015-06-15

Highlights: • Real time control system for poloidal beta has been designed in KSTAR. • Poloidal beta has been calculated based on the diamagnetic loop signals. • The neutral beam Injector plays a role as the actuator. • The control system has been validated in the KSTAR experiments. - Abstract: Sustaining the plasma in a stable and a high performance condition is one of the important control issues for future steady state tokamaks. In the 2014 KSTAR campaign, we have developed a real-time poloidal beta (β{sub p}) control technique and carried out preliminary experiments to identify its feasibility. In the control system, the β{sub p} is calculated in real time using the measured diamagnetic loop signal, and compared with the target value leading to the change of the neutral beam (NB) heating power using a feedback PID control algorithm. To match the requested power of NB which is operated with constant voltage, the on-time periods of the intervals were adjusted as the ratio of the required power to the maximum achievable one. This paper will present the overall procedures of the β{sub p} control, the β{sub p} estimation process and NB operation scheme implemented in the plasma control system (PCS), and the analysis on the preliminary experimental results.

Demonstration of real-time control for poloidal beta in KSTAR

International Nuclear Information System (INIS)

Han, Hyunsun; Hahn, S.H.; Bak, J.G.; Hyatt, A.; Johnson, R.; Woo, M.H.; Kim, J.S.; Bae, Y.S.

2015-01-01

Highlights: • Real time control system for poloidal beta has been designed in KSTAR. • Poloidal beta has been calculated based on the diamagnetic loop signals. • The neutral beam Injector plays a role as the actuator. • The control system has been validated in the KSTAR experiments. - Abstract: Sustaining the plasma in a stable and a high performance condition is one of the important control issues for future steady state tokamaks. In the 2014 KSTAR campaign, we have developed a real-time poloidal beta (β p ) control technique and carried out preliminary experiments to identify its feasibility. In the control system, the β p is calculated in real time using the measured diamagnetic loop signal, and compared with the target value leading to the change of the neutral beam (NB) heating power using a feedback PID control algorithm. To match the requested power of NB which is operated with constant voltage, the on-time periods of the intervals were adjusted as the ratio of the required power to the maximum achievable one. This paper will present the overall procedures of the β p control, the β p estimation process and NB operation scheme implemented in the plasma control system (PCS), and the analysis on the preliminary experimental results

An enhanced tokamak startup model

Science.gov (United States)

Goswami, Rajiv; Artaud, Jean-François

2017-01-01

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

Electron cyclotron heating studies of the Compact Ignition Tokamak (CIT)

International Nuclear Information System (INIS)

Porkolab, M.; Bonoli, P.T.; Englade, R.; Myer, R.; Smith, G.R.; Kritz, A.H.

1989-01-01

The Compact Ignition Tokamak (CIT) operating scenario calls for ramping the toroidal magnetic field from B/sub T/ = 7.0 (8.0) to 10.0 Tesla in a few seconds, followed by a burn cycle and a ramp-down cycle. Simultaneously, the plasma must be heated from an initial low beta equilibrium (/bar /beta// ≅ 0.44% at 7.0 to 8.0 Tesla) to a final burn equilibrium (/bar /beta// = 2.8%) having 10.0 Tesla on the magnetic axis. Since the toroidal plasma current will be ramped at the same time and since the available time for flat-top magnetic field must be reserved for the burn cycle, it is imperative that densification and heating be carried out as the magnetic field is ramped. Here we examine an approach which is applicable to ECR heating. The frequency remains constant, while the angle of injection is varied by simply rotating a reflecting mirror placed in the path of the incident microwave beam. The rotating mirror permits one to launch waves with sufficiently high N/sub /parallel// so that the Doppler broadened resonance of particles on the magnetic axis with f = 280 GHz and B/sub T/ = 7.0--8.0 Tesla can provide adequate absorption. As the resonance layer moves toward the magnetic axis the beam is swept toward perpendicular to reduce the Doppler width and avoid heating the plasma edge. At B/sub T/ = 10.0 Tesla the beam will be at normal incidence with strong absorption immediately on the high field side of the resonance (relativistic regime). We envisage using the ordinary mode (O-mode, /rvec E//sub RF/ /parallel/ /rvec B/) of polarization which is accessible from the outside (low-field side) of the torus provided the density is such that ω/sub pe/ ≤ ω ∼ ω/sub ce/ (max). 8 refs., 3 figs

High-power heating experiment of spherical tokamaks by use of plasma merging

International Nuclear Information System (INIS)

Ueda, Yoshinobu; Ono, Yasushi

1999-01-01

High-power heating of spherical tokamaks (STs) has been investigated experimentally by use of plasma merging effect. When two STs were coaxially collided, thermal energy of a colliding ST was injected into a target ST during short reconnection time (Alfven time). Though the thermal energy increment increased with decreasing plasma q value, thermal energy loss during the following relaxation, tended to be smaller with increasing q. The produced high-β STs had hallower current profiles and weaker paramagnetic toroidal field than those of single STs. Those heating properties indicate the plasma merging to be a promising initial heating method of ST plasmas. (author)

Parameters of the luminous region surrounding deuterium pellets in the PLT tokamak

International Nuclear Information System (INIS)

McNeill, D.H.; Greene, G.J.; Schuresko, D.D.

1985-08-01

The luminous region of the plasma cloud surrounding deuterium pellets injected into a tokamak is studied spectroscopically. At the time of peak luminosity the average electron density is 2.4 x 10 17 cm -3 to within 30% and the temperature is at most 2.0 eV. The intensity ratio of the Balmer alpha and beta light from the pellets, the total number of emitted photons, and the apparent size of the radiating region are consistent with local thermodynamic equilibrium at this temperature and density

Digital control of plasma position in Damavand tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2002-03-01

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

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

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

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

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

Time - resolved thermography at Tokamak T-10

International Nuclear Information System (INIS)

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

1987-01-01

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

Experience with high heat flux components in large tokamaks

International Nuclear Information System (INIS)

Chappuis, P.; Dietz, K.J.; Ulrickson, M.

1991-01-01

The large present day tokamaks. i.e.JET, TFTR, JT-60, DIII-D and Tore Supra are machines capable of sustaining plasma currents of several million amperes. Pulse durations range from a few seconds up to a minute. These large machines have been in operation for several years and there exists wide experience with materials for plasma facing components. Bare and coated metals, bare and coated graphites and beryllium were used for walls, limiters and divertors. High heat flux components are mainly radiation cooled, but stationary cooling for long pulse duration is also employed. This paper summarizes the experience gained in the large machines with respect to material selection, component design, problem areas, and plasma performance. 2 tabs., 26 figs., 50 refs

Non-inductively driven tokamak plasmas at near-unity βt in the Pegasus toroidal experiment

Science.gov (United States)

Reusch, J. A.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Perry, J. M.; Pierren, C.; Rhodes, A. T.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Weberski, J. D.

2018-05-01

A major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓi, high elongation κ, and high toroidal and normalized beta ( βt and βN) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓi. The low aspect ratio ( R0/a ˜1.2 ) of Pegasus allows access to high κ and high normalized plasma currents ( IN=Ip/a BT>14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high βt plasmas. Equilibrium analyses indicate that βt up to ˜100% is achieved. These high βt discharges disrupt at the ideal no-wall β limit at βN˜7.

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

International Nuclear Information System (INIS)

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

2014-03-01

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

MHD activity and energy loss during beta saturation and collapse at high beta poloidal in PBX

International Nuclear Information System (INIS)

Kugel, H.W.; Sesnic, S.; Bol, K.

1987-10-01

High-β experiments, in medium to high-q tokamak plasmas, exhibit a temporal β saturation and collapse. This behavior has been attributed to ballooning, ideal kink, or tearing modes. In PBX, a unique diagnostic capability allowed studies of the relation between MHD and energy loss for neutral-beam-heated (<6 MW), mildly indented (10 to 15%), nearly steady I/sub p/ discharges that approached the Troyon-Gruber limit. Under these conditions, correlations between MHD activity and energy losses have shown that the latter can be almost fully accounted for by various long wavelength MHD instabilities and that there is no need to invoke high-n ballooning modes in PBX. 6 refs., 4 figs

Development in Diagnostics Application to Control Advanced Tokamak Plasma

International Nuclear Information System (INIS)

Koide, Y.

2008-01-01

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

Regime of very high confinement in the boronized DIII-D tokamak

International Nuclear Information System (INIS)

Jackson, G.L.; Winter, J.; Taylor, T.S.; Burrell, K.H.; DeBoo, J.C.; Greenfield, C.M.; Groebner, R.J.; Hodapp, T.; Holtrop, K.; Lazarus, E.A.; Lao, L.L.; Lippmann, S.I.; Osborne, T.H.; Petrie, T.W.; Phillips, J.; James, R.; Schissel, D.P.; Strait, E.J.; Turnbull, A.D.; West, W.P.; DIII-D Team

1991-01-01

Following boronization, tokamak discharges in DIII-D have been obtained with confinement times up to a factor of 3.5 above the ITER89-P L-mode scaling and 1.8 times greater than the DIII-D/JET H-mode scaling relation. Very high confinement phases are characterized by relatively high central density with n e (0)∼1x10 20 m -3 , and central ion temperatures up to 13.6 keV at moderate plasma currents (1.6 MA) and heating powers (12.5--15.3 MW). These discharges exhibit a low fraction of radiated power, P≤25%, Z eff (0) close to unity, and lower impurity influxes than comparable DIII-D discharges before boronization

Compact toroid fueling of the TdeV tokamak

International Nuclear Information System (INIS)

Martin, F.; Raman, R.; Xiao, C.; Thomas, J.

1993-01-01

Compact toroids have been proposed as a means of centrally fueling tokamak reactors because of the high velocity to which they can be accelerated. These are cold (T e ∼ 10 eV), high density (n e > 10 20 m -3 ) spheromak plasmoids that are accelerated in a magnetized Marshall gun. As a proof of principle experiment, a compact toroid fueler (CTF) has been developed for injection into the TdeV tokamak. The engineering goals of the experiment are to measure and minimize the impurity content of the CT plasma and the neutral gas remaining after CT formation. Also of importance is the effect of CT central fueling on the tokamak density profile and bootstrap current, and the relaxation rate of the density profile providing information on the confinement time of the CT fuel

Tokamak confinement scaling laws

International Nuclear Information System (INIS)

Connor, J.

1998-01-01

The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

Measurement of gross beta radioactivity in high-level liquid waste

International Nuclear Information System (INIS)

Lu Feng; Lin Cansheng; Zhang Xianzi; Chen Guoan; Zhang Chonghai

1992-01-01

Using beta plastic scintillation counter of low level background, gross beta radioactivity of twelve samples for high-level liquid waste is determined directly. Beta efficiency curves of plastic scintillation counter for four mass thickness are calibrated in advance. Determining gross beta radioactivity, gross efficiency of the scintillation counter for various energy beta ray is calculated via weighted mean method with the ratio of radioactivity for each nuclide. The ratio of radioactivity for nuclides which have gamma disintegration is determined in terms of the radioactivity measured by gamma spectrometer. The ratio of the radioactivity for 90 Sr which has purity beta disintegration is calculated in terms of half life time approximation. The ratio of the radioactivity for 147 Pm which also has purity disintegration is calculated by means of apparent cooling-time approximation. The uncertainty of results for the present work is about +-15%

The spheric tokamak programme at Culham

International Nuclear Information System (INIS)

Sykes, A.

1999-01-01

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

On the generation of Alfven wave current drive in low aspect ratio Tokamaks with neoclassical conductivity

Energy Technology Data Exchange (ETDEWEB)

Bruma, C.; Cuperman, S.; Komoshvili, K. [School of Physics and Astronomy, Tel Aviv University, Tel Aviv (Israel)

1998-08-01

Several low aspect ratio (spherical) Tokamaks (ST's) are now in operation or under construction. These devices would permit cost-effective and attractive embodiment of future fusion reactors: they would provide high {beta}, good confinement and steady state operation at modest field values. Now, a steady state reactor has to be sustained by non-inductively driven currents. Recently, the generation of non-inductive current drive by Alfven waves (AWCD) has been investigated theoretically within the framework of ideal (E{sub p}arallel=0) MHD and non-ideal, resistive (E{sub p}arallel{ne}0) MHD; however, in all these cases, the tokamak device consisted of a cylindrical plasma with simulated toroidal effects. Rather encouraging results have been obtained. In this work we further investigate AWCD in ST's as follows: (i) we use consistent equilibrium profiles with neoclassical conductivity corresponding to an ohmic START discharge; (ii) incorporate effects due to neoclassical conductivity in the elements of the resistive MHD dielectric tensor, in the solution of the full (E{sub p}arallel{ne}0) wave equation as well as in the calculation of AWCD; and (iii) carry out a systematic search for antenna parameters optimizing the AWCD. (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)

Production and analysis of thermonuclear plasmas in high beta devices. Progress report

International Nuclear Information System (INIS)

1976-01-01

During the October 1975--July 1976 reporting period, significant progress was reported in all ongoing projects funded by ERDA. Construction of the major new experiment THOR was delayed due to difficulties in bringing the Maxwell Laboratory swinging LC-pulse generators up to specifications. These technical difficulties have now been overcome and the pulsers were accepted late in April. THOR is almost back on schedule and physics results on plasma heating are expected by September of this year. TERP, originally funded as an Exploratory Concept, has been operating successfully for over a year and given confinement physics results which are important for the development of Maximum Beta Tokamaks. The Measurement and Instrumentation efforts have resulted in a number of instrument developments that have been successfully tested on FTP or STP, our fast or small theta pinches. Both FTP and STP were studied intensively to obtain ion and electron heating rates and anomalous post-implosion resistivities. These results are supported by the University of Maryland theory and simulation programs and SAI

Predictions and observations of low-shear beta-induced shear Alfven-acoustic eigenmodes in toroidal plasmas

Energy Technology Data Exchange (ETDEWEB)

Gorelenkov, N.N. [Princeton Plasma Physics Laboratory, Princeton University (United States)], E-mail: ngorelen@pppl.gov; Berk, H.L. [IFS, Austin, Texas (United States); Fredrickson, E. [Princeton Plasma Physics Laboratory, Princeton University (United States); Sharapov, S.E. [Euroatom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire (United States)

2007-10-08

New global MHD eigenmode solutions arising in gaps in the low frequency Alfven-acoustic continuum below the geodesic acoustic mode (GAM) frequency have been found numerically and have been used to explain relatively low frequency experimental signals seen in NSTX and JET tokamaks. These global eigenmodes, referred to here as Beta-induced Alfven-Acoustic Eigenmodes (BAAE), exist in the low magnetic safety factor region near the extrema of the Alfven-acoustic continuum. In accordance to the linear dispersion relations, the frequency of these modes shifts as the safety factor, q, decreases. We show that BAAEs can be responsible for observations in JET plasmas at relatively low beta <2% as well as in NSTX plasmas at relatively high-beta >20%. In contrast to the mostly electrostatic character of GAMs the new global modes also contain an electromagnetic (magnetic field line bending) component due to the Alfven coupling, leading to wave phase velocities along the field line that are large compared to the sonic speed. Qualitative agreement between theoretical predictions and observations are found.

Lower hybrid heating experiments in tokamaks: an overview

International Nuclear Information System (INIS)

Porkolab, M.

1985-10-01

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

The residual zonal dynamics in a toroidally rotating tokamak

International Nuclear Information System (INIS)

Zhou Deng

2015-01-01

Zonal flows, initially driven by ion-temperature-gradient turbulence, may evolve due to the neoclassic polarization in a collisionless tokamak plasma. In this presentation, the form of the residual zonal flow is presented for tokamak plasmas rotating toroidally at arbitrary velocity. The gyro-kinetic equation is analytically solved to give the expression of residual zonal flows with arbitrary rotating velocity. The zonal flow level decreases as the rotating velocity increases. The numerical evaluation is in good agreement with the previous simulation result for high aspect ratio tokamaks. (author)

Local transport in Joint European Tokamak edge-localized, high-confinement mode plasmas with H, D, DT, and T isotopes

International Nuclear Information System (INIS)

Budny, R. V.; Ernst, D. R.; Hahm, T. S.; McCune, D. C.; Christiansen, J. P.; Cordey, J. G.; Gowers, C. G.; Guenther, K.; Hawkes, N.; Jarvis, O. N.

2000-01-01

The edge-localized, high-confinement mode regime is of interest for future Tokamak reactors since high performance has been sustained for long durations. Experiments in the Joint European Tokamak [M. Keilhacker , Nuclear Fusion 39, 209 (1999)] have studied this regime using scans with the toroidal field and plasma current varied together in H, D, DT, and T isotopes. The local energy transport in more than fifty of these plasmas is analyzed, and empirical scaling relations are derived for energy transport coefficients during quasi-steady state conditions using dimensionless parameters. Neither the Bohm nor gyro-Bohm expressions give the shapes of the profiles. The scalings with β and ν * are in qualitative agreement with Ion Temperature Gradient theory

Impedance of an intense plasma-cathode electron source for tokamak startup

Science.gov (United States)

Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Perry, J. M.

2016-05-01

An impedance model is formulated and tested for the ˜1 kV , 1 kA/cm2 , arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma ( narc≈1021 m-3 ) within the electron source, and the less dense external tokamak edge plasma ( nedge≈1018 m-3 ) into which current is injected at the applied injector voltage, Vinj . Experiments on the Pegasus spherical tokamak show that the injected current, Iinj , increases with Vinj according to the standard double layer scaling Iinj˜Vinj3 /2 at low current and transitions to Iinj˜Vinj1 /2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb˜Iinj/Vinj1 /2 . For low tokamak edge density nedge and high Iinj , the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb˜narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.

Activation analysis of the compact ignition tokamak

International Nuclear Information System (INIS)

Selcow, E.C.

1986-01-01

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

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

Nonneutralized charge effects on tokamak edge magnetohydrodynamic stability

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Axisymmetric tokamak scapeoff transport

International Nuclear Information System (INIS)

Singer, C.E.; Langer, W.D.

1982-08-01

We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed

ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

International Nuclear Information System (INIS)

HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

2002-01-01

A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

Resistive internal kink modes in a tokamak with high-pressure plasma

International Nuclear Information System (INIS)

Kuvshinov, B.N.; Mikhajlovskij, A.B.; Tatarinov, E.G.

1988-01-01

Theory of resistive internal kink modes in a tokamak with high-pressure plasma is developed. Equation for Fourie-image of disturbed displacment in a resistive layer ie derived with regard to effects of the fourth order by plasma pressure within the framework of single-liquid approach. In its structure this equation coincides with a similar equation for resistive balloon modes and has an exact solution expressed by degenerated hypergeometric function. A general dispersion equation for resistive kink modes is derived with regard to the effects indicated. It is shown that plasma pressure finiteness leads to the reduction of reconnection and tyring-mode increments

Electron thermal transport in tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Konings, J A

1994-11-30

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

Study on possibility of plasma current profile determination using an analytical model of tokamak equilibrium

International Nuclear Information System (INIS)

Moriyama, Shin-ichi; Hiraki, Naoji

1996-01-01

The possibility of determining the current profile of tokamak plasma from the external magnetic measurements alone is investigated using an analytical model of tokamak equilibrium. The model, which is based on an approximate solution of the Grad-Shafranov equation, can set a plasma current profile expressed with four free parameters of the total plasma current, the poloidal beta, the plasma internal inductance and the axial safety factor. The analysis done with this model indicates that, for a D-shaped plasma, the boundary poloidal magnetic field prescribing the external magnetic field distribution is dependent on the axial safety factor in spite of keeping the boundary safety factor and the plasma internal inductance constant. This suggests that the plasma current profile is reversely determined from the external magnetic analysis. The possibility and the limitation of current profile determination are discussed through this analytical result. (author)

Models for impurity effects in tokamaks

International Nuclear Information System (INIS)

Hogan, J.T.

1980-03-01

Models for impurity effects in tokamaks are described with an emphasis on the relationship between attainment of high β and impurity problems. We briefly describe the status of attempts to employ neutral beam heating to achieve high β in tokamaks and propose a qualitative model for the mechanism by which heavy metal impurities may be produced in the startup phase of the discharge. We then describe paradoxes in impurity diffusion theory and discuss possible resolutions in terms of the effects of large-scale islands and sawtooth oscillations. Finally, we examine the prospects for the Zakharov-Shafranov catastrophe (long time scale disintegration of FCT equilibria) in the context of present and near-term experimental capability

Can better modelling improve tokamak control?

International Nuclear Information System (INIS)

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

1997-01-01

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

Advanced Tokamak Stability Theory

Science.gov (United States)

Zheng, Linjin

2015-03-01

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

Disruption mitigation with high-pressure helium gas injection on EAST tokamak

Science.gov (United States)

Chen, D. L.; Shen, B.; Granetz, R. S.; Qian, J. P.; Zhuang, H. D.; Zeng, L.; Duan, Y.; Shi, T.; Wang, H.; Sun, Y.; Xiao, B. J.

2018-03-01

High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time  ⩽150 μs, capable of injecting up to 7 × 1022 particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas (N_He≃ N_plasma ) can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity (N_He≃ 40× N_plasma ) can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (˜7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas.

Predictions and Observations of Low-shear Beta-induced Alfvén-acoustic Eigenmodes in Toroidal Plasmas

Energy Technology Data Exchange (ETDEWEB)

Gorelenkov, N. N.; Berk, H. L.; Fredrickson, E.; Sharapov, S. E.

2007-07-02

New global MHD eigenmode solutions arising in gaps in the low frequency Alfvén -acoustic continuum below the geodesic acoustic mode (GAM) frequency have been found numerically and have been used to explain relatively low frequency experimental signals seen in NSTX and JET tokamaks. These global eigenmodes, referred to here as Beta-induced Alfvén-Acoustic Eigenmodes (BAAE), exist in the low magnetic safety factor region near the extrema of the Alfvén-acoustic continuum. In accordance to the linear dispersion relations, the frequency of these modes shifts as the safety factor, q, decreases. We show that BAAEs can be responsible for observations in JET plasmas at relatively low beta < 2% as well as in NSTX plasmas at relatively high beta > 20%. In contrast to the mostly electrostatic character of GAMs the new global modes also contain an electromagnetic (magnetic field line bending) component due to the Alfvén coupling, leading to wave phase velocities along the field line that are large compared to the sonic speed. Qualitative agreement between theoretical predictions and observations are found.

Investigation of MHD instabilities and control in KSTAR preparing for high beta operation

Science.gov (United States)

Park, Y. S.; Sabbagh, S. A.; Bialek, J. M.; Berkery, J. W.; Lee, S. G.; Ko, W. H.; Bak, J. G.; Jeon, Y. M.; Park, J. K.; Kim, J.; Hahn, S. H.; Ahn, J.-W.; Yoon, S. W.; Lee, K. D.; Choi, M. J.; Yun, G. S.; Park, H. K.; You, K.-I.; Bae, Y. S.; Oh, Y. K.; Kim, W.-C.; Kwak, J. G.

2013-08-01

Initial H-mode operation of the Korea Superconducting Tokamak Advanced Research (KSTAR) is expanded to higher normalized beta and lower plasma internal inductance moving towards design target operation. As a key supporting device for ITER, an important goal for KSTAR is to produce physics understanding of MHD instabilities at long pulse with steady-state profiles, at high normalized beta, and over a wide range of plasma rotation profiles. An advance from initial plasma operation is a significant increase in plasma stored energy and normalized beta, with Wtot = 340 kJ, βN = 1.9, which is 75% of the level required to reach the computed ideal n = 1 no-wall stability limit. The internal inductance was lowered to 0.9 at sustained H-mode duration up to 5 s. In ohmically heated plasmas, the plasma current reached 1 MA with prolonged pulse length up to 12 s. Rotating MHD modes are observed in the device with perturbations having tearing rather than ideal parity. Modes with m/n = 3/2 are triggered during the H-mode phase but are relatively weak and do not substantially reduce Wtot. In contrast, 2/1 modes to date only appear when the plasma rotation profiles are lowered after H-L back-transition. Subsequent 2/1 mode locking creates a repetitive collapse of βN by more than 50%. Onset behaviour suggests the 3/2 mode is close to being neoclassically unstable. A correlation between the 2/1 mode amplitude and local rotation shear from an x-ray imaging crystal spectrometer suggests that the rotation shear at the mode rational surface is stabilizing. As a method to access the ITER-relevant low plasma rotation regime, plasma rotation alteration by n = 1, 2 applied fields and associated neoclassical toroidal viscosity (NTV) induced torque is presently investigated. The net rotation profile change measured by a charge exchange recombination diagnostic with proper compensation of plasma boundary movement shows initial evidence of non-resonant rotation damping by the n = 1, 2 applied

Investigation of MHD instabilities and control in KSTAR preparing for high beta operation

International Nuclear Information System (INIS)

Park, Y.S.; Sabbagh, S.A.; Bialek, J.M.; Berkery, J.W.; Lee, S.G.; Ko, W.H.; Bak, J.G.; Jeon, Y.M.; Kim, J.; Hahn, S.H.; Yoon, S.W.; Lee, K.D.; You, K.-I.; Bae, Y.S.; Oh, Y.K.; Park, J.K.; Ahn, J.-W.; Choi, M.J.; Yun, G.S.; Park, H.K.

2013-01-01

Initial H-mode operation of the Korea Superconducting Tokamak Advanced Research (KSTAR) is expanded to higher normalized beta and lower plasma internal inductance moving towards design target operation. As a key supporting device for ITER, an important goal for KSTAR is to produce physics understanding of MHD instabilities at long pulse with steady-state profiles, at high normalized beta, and over a wide range of plasma rotation profiles. An advance from initial plasma operation is a significant increase in plasma stored energy and normalized beta, with W tot = 340 kJ, β N = 1.9, which is 75% of the level required to reach the computed ideal n = 1 no-wall stability limit. The internal inductance was lowered to 0.9 at sustained H-mode duration up to 5 s. In ohmically heated plasmas, the plasma current reached 1 MA with prolonged pulse length up to 12 s. Rotating MHD modes are observed in the device with perturbations having tearing rather than ideal parity. Modes with m/n = 3/2 are triggered during the H-mode phase but are relatively weak and do not substantially reduce W tot . In contrast, 2/1 modes to date only appear when the plasma rotation profiles are lowered after H–L back-transition. Subsequent 2/1 mode locking creates a repetitive collapse of β N by more than 50%. Onset behaviour suggests the 3/2 mode is close to being neoclassically unstable. A correlation between the 2/1 mode amplitude and local rotation shear from an x-ray imaging crystal spectrometer suggests that the rotation shear at the mode rational surface is stabilizing. As a method to access the ITER-relevant low plasma rotation regime, plasma rotation alteration by n = 1, 2 applied fields and associated neoclassical toroidal viscosity (NTV) induced torque is presently investigated. The net rotation profile change measured by a charge exchange recombination diagnostic with proper compensation of plasma boundary movement shows initial evidence of non-resonant rotation damping by the n = 1, 2

Ultra-long pulse operation using lower hybrid waves on the superconducting high field tokamak TRIAM-1M

International Nuclear Information System (INIS)

Moriyama, S.; Nakamura, Y.; Nagao, A.; Jotaki, E.; Nakamura, K.; Hiraki, N.; Itoh, S.

1990-01-01

Ultra-long pulse operation (>3 min) was achieved on the superconducting high field tokamak TRIAM-1M. In this operation, the plasma current was maintained with a relatively peaked current distribution by the 2.45 GHz radiofrequency power (P RF ≤ 35 kW) alone. A stationary plasma with a driven current of up to 35 kA and a line averaged electron density of up to 3x10 12 cm -3 was produced by precise plasma position and gas feed control. The extremely long discharge showed the interesting characteristics that the high temperatures of about 1 keV for the electrons and about 0.5 keV for the ions were kept almost constant during steady state current drive and that there was no impurity accumulation which could have a fatally adverse effect on steady state tokamak operation. (author). 16 refs, 17 figs

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

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.)

High precision wavelength measurements of X-ray lines emitted from TS-Tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Platz, P. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Cornille, M.; Dubau, J. [Observatoire de Paris, 92 - Meudon (France)

1996-01-01

X-ray line spectra from highly charged impurity ions have been taken with a high-resolution Bragg-crystal spectrometer on the Tore Supra (TS) tokamak. By cross-checking the wavelengths of reference lines from the heliumlike ions Ti20 + (2.6 Angstroms) and Ar16 + (3.95 Angstroms) we first demonstrate that it is possible to measure wavelengths with a precision, {lambda}/{delta}{lambda}, of better than 50000. We than determine the wavelengths of n=3 to n=2 transitions of neonlike Ag37+ in the 4 Angstroms spectral range. (authors). 16 refs., 7 figs., 3 tabs.

Control of the Resistive Wall Mode in Advanced Tokamak Plasmas on DIII-D

International Nuclear Information System (INIS)

Garofalo, A.M.; Strait, E.J.; Bialek, J.; Frederickson, E.; Gryaznevich, M.; Jensen, T.H.; Johnson, L.C.; La Haye, R.J.; Navratil, G.A.; Lazarus, E.A.; Luce, T.C.; Makowski, M.; Okabayashi, M.; Rice, B.W.; Scoville, J.T.; Turnbull, A.D.; Walker, M.L.

1999-01-01

Resistive wall mode (RWM) instabilities are found to be a limiting factor in advanced tokamak (AT) regimes with low internal inductance. Even small amplitude modes can affect the rotation profile and the performance of these ELMing H-mode discharges. Although complete stabilization of the RWM by plasma rotation has not yet been observed, several discharges with increased beam momentum and power injection sustained good steady-state performance for record time extents. The first investigation of active feedback control of the RWM has shown promising results: the leakage of the radial magnetic flux through the resistive wall can be successfully controlled, and the duration of the high beta phase can be prolonged. The results provide a comparative test of several approaches to active feedback control, and are being used to benchmark the analysis and computational models of active control

Comparative studies of stellarator and tokamak transport

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-01

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

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

Speculative Betas

OpenAIRE

Harrison Hong; David Sraer

2012-01-01

We provide a model for why high beta assets are more prone to speculative overpricing than low beta ones. When investors disagree about the common factor of cash-flows, high beta assets are more sensitive to this macro-disagreement and experience a greater divergence-of-opinion about their payoffs. Short-sales constraints for some investors such as retail mutual funds result in high beta assets being over-priced. When aggregate disagreement is low, expected return increases with beta due to r...

Single Null Negative Triangularity Tokamak for Power Handling

Science.gov (United States)

Kikuchi, Mitsuru; Medvedev, S.; Takizuka, T.; Sauter, O.; Merle, A.; Coda, S.; Chen, D.; Li, J. X.

2017-10-01

Power and particle control in fusion reactor is challenge and we proposed the negative triangularity tokamak (NTT) to eliminate ELM by operating L-mode edge with improved core confinement. The SN configuration has more flexibility in shaping by adopting rectangular-shaped TF coils. The limiting normalized beta is 3.56 with wall stabilization and 3.14 without wall. The vertical stability is assured under a reasonable control system. The wetted area on the divertor plates becomes wider in proportion to the larger major radius at the divertor strike points due to the NT configuration. In addition to the major-radius effect, the ``Flux Tune Expansion (FTE)'' is adopted to further reduce the heat load on the divertor plate by factor of 2.6 with a coil current 3 MA. L-mode edge also allows further increase in wetted area. The fusion power of 3 GW is deliverable only at normalized beta 2.1. Therefore this reactor may be operable stably against the serious MHD activities. The CD power for SS operation is 175 MW at Q = 17. AC operation is also possible option. A required HH factor is relatively modest H = 1.12.

Magnetic sensor for steady state tokamak

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

Energy Technology Data Exchange (ETDEWEB)

Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Capece, A.; Koel, B.; Roszell, J. [Princeton University, Princeton, New Jersey 08544 (United States); Biewer, T. M.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Kubota, S. [University of California at Los Angeles, Los Angeles, California 90095 (United States); Beiersdorfer, P. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

2015-05-15

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

International Nuclear Information System (INIS)

Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G.; Capece, A.; Koel, B.; Roszell, J.; Biewer, T. M.; Gray, T. K.; Kubota, S.; Beiersdorfer, P.

2015-01-01

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started

Edge radial electric field structure in quiescent H-mode plasmas in the DIII-D tokamak

Energy Technology Data Exchange (ETDEWEB)

Burrell, K H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); West, W P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Doyle, E J [University of California, Los Angeles, CA 90095-1597 (United States); Austin, M E [University of Texas at Austin, Austin, TX 78712 (United States); DeGrassie, J S [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Gohil, P [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Greenfield, C M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Groebner, R J [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Jayakumar, R [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); Kaplan, D H [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lao, L L [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Leonard, A W [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M A [Lawrence Livermore National Laboratory, Livermore, CA 94551-9900 (United States); McKee, G R [University of Wisconsin, Madison, WI 53706-1687 (United States); Solomon, W M [Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451 (United States); Thomas, D M [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Rhodes, T L [University of California, Los Angeles, CA 90095-1597 (United States); Wade, M R [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wang, G [University of California, Los Angeles, CA 90095-1597 (United States); Watkins, J G [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Zeng, L [University of California, Los Angeles, CA 90095-1597 (United States)

2004-05-01

H-mode operation is the choice for next step tokamak devices based on either conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the {beta} limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D over the past four years have demonstrated a new operating regime, the quiescent H-mode (QH-mode) regime, that solves these problems. QH-mode plasmas have now been run for over 4 s (>30 energy confinement times). Utilizing the steady-state nature of the QH-mode edge allows us to obtain unprecedented spatial resolution of the edge ion profiles and the edge radial electric field, E{sub r}, by sweeping the edge plasma slowly past the view points of the charge exchange spectroscopy system. We have investigated the effects of direct edge ion orbit loss on the creation and sustainment of the QH-mode. Direct loss of ions injected into the velocity-space loss cone at the plasma edge is not necessary for creation or sustainment of the QH-mode. The direct ion orbit loss has little effect on the edge E{sub r} well. The E{sub r} at the bottom of the well in these cases is about -100 kV m{sup -1} compared with -20 to -30 kV m{sup -1} in the standard H-mode. The well is about 1 cm wide, which is close to the diameter of the deuteron gyro-orbit. We also have investigated the effect of changing edge triangularity by changing the plasma shape from upwardly biased single null to magnetically balanced double null. We have now achieved the QH-mode in these double-null plasmas. The increased triangularity allows us to increase pedestal density in QH-mode plasmas by a factor of about 2.5 and overall pedestal pressure by a factor of 2. Pedestal {beta} and {nu}{sup *} values matching the values desired for ITER have been achieved. In

Predictions and observations of global beta-induced Alfven-acoustic modes in JET and NSTX

Energy Technology Data Exchange (ETDEWEB)

Gorelenkov, N N [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Berk, H L [Institute for Fusion Studies, University of Texas, Austin, TX 78712 (United States); Crocker, N A [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Fredrickson, E D [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Kaye, S [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Kubota, S [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Park, H [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States); Peebles, W [Institute of Plasma and Fusion Research, University of California, Los Angeles, CA 90095-1354 (United States); Sabbagh, S A [Department of Applied Physics, Columbia University, New York, NY 10027-6902 (United States); Sharapov, S E [Euroatom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Stutmat, D [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Tritz, K [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Levinton, F M [Nova Photonics, One Oak Place, Princeton, NJ 08540 (United States); Yuh, H [Nova Photonics, One Oak Place, Princeton, NJ 08540 (United States)

2007-12-15

In this paper we report on observations and interpretations of a new class of global MHD eigenmode solutions arising in gaps in the low frequency Alfven-acoustic continuum below the geodesic acoustic mode frequency. These modes have been just reported (Gorelenkov et al 2007 Phys. Lett. 370 70-7) where preliminary comparisons indicate qualitative agreement between theory and experiment. Here we show a more quantitative comparison emphasizing recent NSTX experiments on the observations of the global eigenmodes, referred to as beta-induced Alfven-acoustic eigenmodes (BAAEs), which exist near the extrema of the Alfven-acoustic continuum. In accordance to the linear dispersion relations, the frequency of these modes may shift as the safety factor, q, profile relaxes. We show that BAAEs can be responsible for observations in JET plasmas at relatively low beta <2% as well as in NSTX plasmas at relatively high beta >20%. In NSTX plasma observed magnetic activity has the same properties as predicted by theory for the mode structure and the frequency. Found numerically in NOVA simulations BAAEs are used to explain the observed properties of relatively low frequency experimental signals seen in NSTX and JET tokamaks.

Dependence of {beta} {center_dot} {tau} on plasma shape in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Lazarus, E.A. [Oak Ridge National Lab., TN (United States)

1993-12-31

In this paper we discuss the observed variation in plasma performance with plasma shape, in particular, we shall compare single and double null diverted plasmas. The product {beta} {center_dot} {tau} has been used as a figure-of-merit for comparing different toroidal magnetic configurations. Here we shall use it as the figure-of-merit for comparing differing configurations within the DIII-D tokamak. (author) 5 refs., 5 figs.

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)

Tokamak building-design considerations for a large tokamak device

International Nuclear Information System (INIS)

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

1981-01-01

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

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

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.

Development of hard X-ray spectrometer with high time resolution on the J-TEXT tokamak

Energy Technology Data Exchange (ETDEWEB)

Ma, T.K.; Chen, Z.Y., E-mail: zychen@hust.edu.cn; Huang, D.W.; Tong, R.H.; Yan, W.; Wang, S.Y.; Dai, A.J.; Wang, X.L.

2017-06-01

A hard X-ray (HXR) spectrometer has been developed to study the runaway electrons during the sawtooth activities and during the runaway current plateau phase on the J-TEXT tokamak. The spectrometer system contains four NaI scintillator detectors and a multi-channel analyzer (MCA) with 0.5 ms time resolution. The dedicated peak detection circuit embedded in the MCA provides a pulse height analysis at count rate up to 1.2 million counts per second (Mcps), which is the key to reach the high time resolution. The accuracy and reliability of the system have been verified by comparing with the hardware integrator of HXR flux. The temporal evolution of HXR flux in different energy ranges can be obtained with high time resolution by this dedicated HXR spectrometer. The response of runaway electron transport with different energy during the sawtooth activities can be studied. The energy evolution of runaway electrons during the plateau phase of runaway current can be obtained. - Highlights: • A HXR spectrometer with high time resolution has been developed on J-TEXT tokamak. • The response of REs transport during the sawtooth activities can be investigated. • The energy evolution of REs following the disruptions can be monitored.

Discharge cleaning for a tokamak

International Nuclear Information System (INIS)

Ishii, Shigeyuki

1983-01-01

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

Electron cyclotron heating/current-drive system using high power tubes for QUEST spherical tokamak

Science.gov (United States)

Onchi, Takumi; Idei, H.; Hasegawa, M.; Nagata, T.; Kuroda, K.; Hanada, K.; Kariya, T.; Kubo, S.; Tsujimura, T. I.; Kobayashi, S.; Quest Team

2017-10-01

Electron cyclotron heating (ECH) is the primary method to ramp up plasma current non-inductively in QUEST spherical tokamak. A 28 GHz gyrotron is employed for short pulses, where the radio frequency (RF) power is about 300 kW. Current ramp-up efficiency of 0.5 A/W has been obtained with focused beam of the second harmonic X-mode. A quasi-optical polarizer unit has been newly installed to avoid arcing events. For steady-state tokamak operation, 8.56 GHz klystron with power of 200 kW is used as the CW-RF source. The high voltage power supply (54 kV/13 A) for the klystron has been built recently, and initial bench test of the CW-ECH system is starting. The array of insulated-gate bipolar transistor works to quickly cut off the input power for protecting the klystron. This work is supported by JSPS KAKENHI (15H04231), NIFS Collaboration Research program (NIFS13KUTR085, NIFS17KUTR128), and through MEXT funding for young scientists associated with active promotion of national university reforms.

High beta and second stability region transport and stability analysis

International Nuclear Information System (INIS)

1991-01-01

This document describes ideal and resistive MHD studies of high-beta plasmas and of the second stability region. Significant progress is reported on the resistive stability properties of high beta poloidal ''supershot'' discharges. For these studies initial profiles were taken from the TRANSP code which is used extensively to analyze experimental data. When an ad hoc method of removing the finite pressure stabilization of tearing modes is implemented it is shown that there is substantial agreement between MHD stability computation and experiment. In particular, the mode structures observed experimentally are consistent with the predictions of the resistive MHD model. We also report on resistive stability near the transition to the second region in TFTR. Tearing modes associated with a nearby infernal mode may explain the increase in MHD activity seen in high beta supershots and which impede the realization of Q∼1. We also report on a collaborative study with PPPL involving sawtooth stabilization with ICRF

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

International Nuclear Information System (INIS)

Ide, S.

2005-01-01

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

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

Magnetohydrodynamic stability of tokamak edge plasmas

International Nuclear Information System (INIS)

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

1998-01-01

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

On the Demand for High-beta Stocks

DEFF Research Database (Denmark)

Christoffersen, Susan E. K.; Simutin, Mikhail

2017-01-01

Prior studies have documented that pension plan sponsors often monitor a fund’s performance relative to a benchmark. We use a first-difference approach to show that in an effort to beat benchmarks, fund managers controlling large pension assets tend to increase their exposure to high-beta stocks...

Tokamak start-up with electron-cyclotron heating

International Nuclear Information System (INIS)

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

1981-01-01

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

Tokamak start-up with electron-cyclotron heating

Energy Technology Data Exchange (ETDEWEB)

Holly, D J; Prager, S C; Shepard, D A; Sprott, J C [Wisconsin Univ., Madison (USA)

1981-11-01

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

Spreading of wave-driven currents in a tokamak

International Nuclear Information System (INIS)

Ignat, D.W.; Kaita, R.; Jardin, S.C.; Okabayashi, M.

1996-01-01

Lower hybrid current drive (LHCD) in the tokamak Princeton Beta Experiment-Modification (PBX-M) is computed with a dynamic model in order to understand an actual discharge aimed at raising the central q above unity. Such configurations offer advantages for steady-state operation and plasma stability. For the particular parameters of this PBX-M experiment, the calculation found singular profiles of plasma current density J and safety factor q developing soon after LHCD begins. Smoothing the lower hybrid-driven current and power using a diffusion-Eke equation and a velocity-independent diffusivity for fast-electron current brought the model into reasonable agreement with the measurements if D fast ∼ 1.0 m 2 /s. Such a value for D fast is in the range suggested by other work

Behaviour of metallic droplets in a tokamak plasma

International Nuclear Information System (INIS)

Hildebrandt, D.; Juettner, B.; Pursch, H.; Jakubka, K.; Stoeckel, J.; Zacek, F.

1989-01-01

Micrometre sized tantalum droplets were injected into a tokamak plasma by a controllable arcing gun located behind the wall. The trajectories of the ablating particles were photographed by a high speed camera. Various possible mechanisms which may explain the observed curvature of the particle paths are discussed. The migration of the ablated material in the tokamak was studied by post-mortem analysis of collector probes and limiters. (author). Letter-to-the-editor. 12 refs, 9 figs

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.

Theory of nonaxisymmetric vertical displacement events in tokamaks

International Nuclear Information System (INIS)

Fitzpatrick, R.

2011-01-01

A semi-analytic sharp-boundary model of a nonaxisymmetric vertical displacement event (VDE) in a large aspect-ratio, high-beta (i.e. β ∼ ε), vertically elongated tokamak plasma is developed. The model is used to simulate nonaxisymmetric VDEs with a wide range of different plasma equilibrium and vacuum vessel parameters. These simulations yield poloidal halo current fractions and toroidal peaking factors whose magnitudes are similar to those seen in experiments, and also reproduce the characteristic inverse scaling between the halo current fraction and the toroidal peaking factor. Moreover, the peak poloidal halo current density in the vacuum vessel is found to correlate strongly with the reciprocal of the minimum edge safety factor attained during the VDE. In addition, under certain circumstances, the ratio of the net sideways force acting on the vacuum vessel to the net vertical force is observed to approach unity. Finally, the peak vertical force per unit area acting on the vessel is found to have a strong correlation with the equilibrium toroidal plasma current at the start of the VDE, but is also found to increase with increasing vacuum vessel resistivity relative to the scrape-off layer plasma.

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

International Nuclear Information System (INIS)

Lopez C, R.

1986-07-01

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

Design of a microwave calorimeter for the microwave tokamak experiment

International Nuclear Information System (INIS)

Marinak, M.

1988-01-01

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

Covalent labeling of the beta-adrenergic ligand-binding site with para-(bromoacetamidyl)benzylcarazolol. A highly potent beta-adrenergic affinity label

International Nuclear Information System (INIS)

Dickinson, K.E.; Heald, S.L.; Jeffs, P.W.; Lefkowitz, R.J.; Caron, M.G.

1985-01-01

Para-(Bromoacetamidyl)benzylcarazolol (pBABC) was synthesized and found to be an extremely potent affinity label for beta-adrenergic receptors. Its interaction with mammalian (rabbit and hamster lung) and nonmammalian (turkey and frog erythrocyte) beta-adrenergic receptors was similar, displaying EC 50 values of 400-900 pM for inhibiting 125 I-cyanopindolol binding to these receptors. pBABC reduced the number of beta-adrenergic receptors in frog erythrocyte membranes, without any change in the affinity of the remaining sites for [ 125 I]iodocyanopindolol. pBABC has been radioiodinated. As assessed by sodium dodecyl sulfate-polyacrylamide gel electrophoresis, this affinity probe specifically labeled the beta-adrenergic peptide of a purified preparation of hamster lung, with high efficiency (approximately 40%) and with a pharmacological specificity characteristic of an interaction at the beta 2-adrenergic receptor ligand-binding site. Comparison of the proteolyzed products derived from purified receptor labeled with [ 125 I]pBABC and with the photoaffinity agent [ 125 I]p-azidobenzylcarazolol suggested that covalent labeling of the beta-adrenergic receptor by these probes occurs at similar domains of the beta-adrenergic receptor

Active feedback stabilization of axisymmetric modes in highly elongated tokamak plasmas

International Nuclear Information System (INIS)

Ward, D.J.; Hofmann, F.

1993-07-01

Active feedback stabilization of the vertical instability is studied for highly elongated tokamak plasmas (1≤κ≤3), and evaluated in particular for the TCV configuration. It is shown that the feedback can strongly affect the form of the eigenfunction for these highly elongated equilibria, and this can have detrimental effects on the ability of the feedback system to properly detect and stabilize the plasma. A calculation of the vertical displacement that uses poloidal flux measurements, poloidal magnetic field measurements, and corrections for the vessel eddy currents and active feedback currents was found to be effective even in the cases with the worst deformations of the eigenfunction. We also examine how these deformations affect differently shaped equilibria, and it is seen that the magnitude of the deformation of the eigenfunction is strongly function of the plasma elongation. (author) 15 figs., 13 refs

Fusion-product transport in axisymmetric tokamaks: losses and thermalization

International Nuclear Information System (INIS)

Hively, L.M.

1980-01-01

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

A flexible software architecture for tokamak discharge control systems

International Nuclear Information System (INIS)

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

1995-01-01

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

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

International Nuclear Information System (INIS)

Wan Baonian

2002-01-01

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

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

International Nuclear Information System (INIS)

Isayama, Akihiko; Tanaka, Hitoshi

2009-01-01

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

High temperature outgassing tests on materials used in the DIII-D tokamak

International Nuclear Information System (INIS)

Holtrop, K.L.; Hansink, M.J.

2006-01-01

This article is a continuation of previous work on determining the outgassing characteristics of materials used in the DIII-D magnetic fusion tokamak [K. L. Holtrop, J. Vac. Sci. Technol. A 17, 2064 (1999)]. Achievement of high performance plasma discharges in the DIII-D tokamak requires careful control of impurity levels. Among the techniques used to control impurities are routine bakes of the vacuum vessel to an average temperature of 350 deg. C. Materials used in DIII-D must release only very small amounts of impurities (below 2x10 -6 mole) at this temperature that could be transferred to the first wall materials and later contaminate plasma discharges. To better study the behavior of materials proposed for use in DIII-D at elevated temperatures, the initial outgassing test chamber was improved to include an independent heating control of the sample and a simple load lock chamber. The goal was to determine not only the total degassing rate of the material during baking, but to also determine the gas species composition and to obtain a quantitative estimate of the degassing rate of each species by the use of a residual gas analyzer. Initial results for aluminum anodized using three different processes, stainless steel plated with black oxide and black chrome, and a commercially available fiber optic feedthrough will be presented

Preliminary Design of Alborz Tokamak

Science.gov (United States)

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

2012-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-10-07

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Highly effective portable beta spectrometer for precise depth selective electron Moessbauer spectroscopy

International Nuclear Information System (INIS)

Aldiyarov, N.U.; Kadyrzhanov, K.K.; Seytimbetov, A.M.; Zhdanov, V.S.

2007-01-01

Full text: More broad application of the nuclear-physical method of precise Depth Selective Electron Moessbauer Spectroscopy (DS EMS) is limited by insufficient accessibility of highly-effective beta spectrometers with acceptable resolution. It should be mentioned that the method DS EMS is realized at a combined installation that consists of a highly-effective beta spectrometer and a conventional portable nuclear gamma-resonance spectrometer. Yet few available beta spectrometers have sophisticated design and controlling; in most cases they are cumbersome. All the attempts to simplify beta spectrometers resulted in noticeable worsening of depth resolution for the DS EMS method making the measurements non precise. There is currently an obvious need in a highly-effective portable easily controlled beta spectrometer. While developing such portable beta spectrometer, it is more promising to use as basis a simpler spectrometer, which has ratio of sample size to spectrometer size of about five times. The paper presents an equal-arm version of a highly-effective portable beta spectrometer with transverse heterogeneous sector magnetic field that assures double focusing. The spectrometer is equipped with a large-area non-equipotential source (a sample under investigation) and a position-sensitive detector. This portable spectrometer meets all requirements for achievement of the DS EMS depth resolution close to the physical limit and demonstrates the following main characteristics: equilibrium orbit radius ρ 0 = 80 mm, instrumental energy resolution 0.6 % at solid angle 1 % of 4π steradian, area of non-equipotential source ∼ 80 mm 2 , registration by position-sensitive detector of ∼ 10 % of the energy interval. Highly-effective portable beta spectrometer assures obtaining Moessbauer data with depth resolution close to physical limit of the DS EMS method. So in measurements at conversion and Auger electrons with energies of about units of keV and above, the achieved

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

Moving Divertor Plates in a Tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Zhang, H.

2009-01-01

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

Moving Divertor Plates in a Tokamak

Energy Technology Data Exchange (ETDEWEB)

S.J. Zweben, H. Zhang

2009-02-12

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

Oat beta-glucan ameliorates insulin resistance in mice fed on high-fat and high-fructose diet

Directory of Open Access Journals (Sweden)

Jie Zheng

2013-12-01

Full Text Available Methods: This study sought to evaluate the impact of oat beta-glucan on insulin resistance in mice fed on high-fat and high-fructose diet with fructose (10%, w/v added in drinking water for 10 weeks. Results: The results showed that supplementation with oat beta-glucan could significantly reduce the insulin resistance both in low-dose (200 mg/kg−1 body weight and high-dose (500 mg/kg−1 body weight groups, but the high-dose group showed a more significant improvement in insulin resistance (P<0.01 compared with model control (MC group along with significant improvement in hepatic glycogen level, oral glucose, and insulin tolerance. Moreover, hepatic glucokinase activity was markedly enhanced both in low-dose and high-dose groups compared with that of MC group (P<0.05. Conclusion: These results suggested that supplementation of oat beta-glucan alleviated insulin resistance and the effect was dose dependent.

Comparison of particle confinement in the high confinement mode plasmas with the edge localized mode of the Japan Atomic Energy Research Institute Tokamak-60 Upgrade and the DIII-D tokamak

International Nuclear Information System (INIS)

Takenaga, H.; Mahdavi, M.A.; Baker, D.R.

2001-01-01

Particle confinement was compared for the high confinement mode plasmas with the edge localized mode in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [S. Ishida, JT-60 Team, Nucl. Fusion 39, 1211 (1999)] and the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] considering separate confinement times for particles supplied by neutral beam injection (NBI) (center fueling) and by recycling and gas-puffing (edge fueling). Similar dependence on the NBI power was obtained in JT-60U and DIII-D. The particle confinement time for center fueling in DIII-D was smaller by a factor of 4 in the low density discharges and by a factor of 1.8 in the high density discharges than JT-60U scaling, respectively, suggesting the stronger dependence on the density in DIII-D. The particle confinement time for edge fueling in DIII-D was comparable with JT-60U scaling in the low density discharges. However, it decreased to a much smaller value in the high density discharges

Progress Towards High Performance, Steady-state Spherical Torus

International Nuclear Information System (INIS)

Ono, M.; Bell, M.G.; Bell, R.E.; Bigelow, T.; Bitter, M.; Blanchard, W.; Boedo, J.; Bourdelle, C.; Bush, C.; Choe, W.; Chrzanowski, J.; Darrow, D.S.; Diem, S.J.; Doerner, R.; Efthimion, P.C.; Ferron, J.R.; Fonck, R.J.; Fredrickson, E.D.; Garstka, G.D.; Gates, D.A.; Gray, T.; Grisham, L.R.; Heidbrink, W.; Hill, K.W.; Hoffman, D.; Jarboe, T.R.; Johnson, D.W.; Kaita, R.; Kaye, S.M.; Kessel, C.; Kim, J.H.; Kissick, M.W.; Kubota, S.; Kugel, H.W.; LeBlanc, B.P.; Lee, K.; Lee, S.G.; Lewicki, B.T.; Luckhardt, S.; Maingi, R.; Majeski, R.; Manickam, J.; Maqueda, R.; Mau, T.K.; Mazzucato, E.; Medley, S.S.; Menard, J.; Mueller, D.; Nelson, B.A.; Neumeyer, C.; Nishino, N.; Ostrander, C.N.; Pacella, D.; Paoletti, F.; Park, H.K.; Park, W.; Paul, S.F.; Peng, Y.-K. M.; Phillips, C.K.; Pinsker, R.; Probert, P.H.; Ramakrishnan, S.; Raman, R.; Redi, M.; Roquemore, A.L.; Rosenberg, A.; Ryan, P.M.; Sabbagh, S.A.; Schaffer, M.; Schooff, R.J.; Seraydarian, R.; Skinner, C.H.; Sontag, A.C.; Soukhanovskii, V.; Spaleta, J.; Stevenson, T.; Stutman, D.; Swain, D.W.; Synakowski, E.; Takase, Y.; Tang, X.; Taylor, G.; Timberlake, J.; Tritz, K.L.; Unterberg, E.A.; Von Halle, A.; Wilgen, J.; Williams, M.; Wilson, J.R.; Xu, X.; Zweben, S.J.; Akers, R.; Barry, R.E.; Beiersdorfer, P.; Bialek, J.M.; Blagojevic, B.; Bonoli, P.T.; Carter, M.D.; Davis, W.; Deng, B.; Dudek, L.; Egedal, J.; Ellis, R.; Finkenthal, M.; Foley, J.; Fredd, E.; Glasser, A.; Gibney, T.; Gilmore, M.; Goldston, R.J.; Hatcher, R.E.; Hawryluk, R.J.; Houlberg, W.; Harvey, R.; Jardin, S.C.; Hosea, J.C.; Ji, H.; Kalish, M.; Lowrance, J.; Lao, L.L.; Levinton, F.M.; Luhmann, N.C.; Marsala, R.; Mastravito, D.; Menon, M.M.; Mitarai, O.; Nagata, M.; Oliaro, G.; Parsells, R.; Peebles, T.; Peneflor, B.; Piglowski, D.; Porter, G.D.; Ram, A.K.; Rensink, M.; Rewoldt, G.; Roney, P.; Shaing, K.; Shiraiwa, S.; Sichta, P.; Stotler, D.; Stratton, B.C.; Vero, R.; Wampler, W.R.; Wurden, G.A.

2003-01-01

Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction (∼60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted

Effects of enhanced elongation and paramagnetism on the parameter space of the ignition spherical torus

International Nuclear Information System (INIS)

Strickler, D.J.; Peng, Y-K.M.; Borowski, S.K.; Selcow, E.C.; Miller, J.B.

1985-01-01

The Ignition Spherical Torus (IST) is a small aspect ratio device retaining only indispensable components along the major axis of a tokamak plasma, such as a cooled, normal conductor producing a toroidal magnetic field. The IST is expected to be a cost-effective approach to ignition by taking advantage of low field, large natural plasma elongation, high plasma current, high beta, and tokamak confinement. These result in compact, high-performance devices with relatively simple magnetic systems as compared with ignition tokamaks of larger aspect ratio. The plasma enhancement of the toroidal field on axis, or plasma paramagnetism, is significant in the IST. The use of this plasma-enhanced field in conventional tokamak beta and density limits leads to increased plasma pressure and performance and therefore smaller device size for a given ignition margin

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

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

New results from Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.K.

2002-01-01

New results from Globus-M spherical tokamak (ST) are presented. Reported are the achievements of high plasma current of 0.36 MA and high toroidal magnetic field of 0.55 T. Plasma column stability in Globus-M is conserved at low edge safety factors and high plasma densities. Achieved lowest safety factor was q(cyl) 19 m -3 . New methods of density increase are discussed. Low-density boarder of operational space is investigated. Runaway electrons properties and conditions of their generation are investigated. Results look promising for STs. Plasma-wall interaction study was performed. Silicon probes were installed into vacuum vessel. They were exposed to boronization, first, and then deposited film interacted with plasma. Discussed are film properties. Briefly described are new diagnostic tools installed on tokamak. Status and preliminary results obtained with auxiliary heating systems are shown. (author)

Generalized MHD for numerical stability analysis of high-performance plasmas in tokamaks

International Nuclear Information System (INIS)

Mikhailovskii, A.B.

1998-01-01

A set of generalized magnetohydrodynamic (MHD) equations is formulated to accommodate the effects associated with high ion and electron temperatures in high-performance plasmas in tokamaks. The effects of neoclassical bootstrap current, neoclassical ion viscosity, the ion finite Larmor radius effect and electron and ion drift effects are taken into account in two-fluid MHD equations together with gyroviscosity, parallel viscosity, electron parallel inertia and collisionless ion heat flux. The ion velocity is identified as the plasma velocity, while the electron velocity is expressed in terms of the plasma velocity and electric current. Ion and electron momentum equations are combined to give the plasma momentum equation. The perpendicular (with respect to the equilibrium magnetic field) ion momentum equation is used as perpendicular Ohm's law and the parallel electron momentum equation - as parallel Ohm's law. Perpendicular Ohm's law allows for the Hall and ion drift effects. Parallel Ohm's law includes the electron drift effect, collisionless skin effect and bootstrap current. In addition, both perpendicular and parallel Ohm's laws contain the resistivity. Due to the quasineutrality condition, the ions and electrons are characterized by the same number density which is described by the ion continuity equation. On the other hand, the ion and electron temperatures are allowed to be different. The ion temperature is described by the ion energy equation allowing for the oblique heat flux, in addition to the perpendicular ion heat flux. The electron temperature is determined by the condition of high parallel electron heat conductivity. The ion and electron parallel viscosities are represented in a form valid for all the collisionality regimes (Pfirsch-Schluter, plateau, and banana). An optimized form of the generalized MHD equations is then represented in terms of the toroidal coordinate system used in the JET equilibrium and stability codes. The derived equations

Scaling for scrape-off layer plasma in tokamak

International Nuclear Information System (INIS)

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

1977-12-01

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

LIDAR Thomson scattering for advanced tokamaks. Final report

International Nuclear Information System (INIS)

Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

1996-01-01

The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured

High beta plasma operation in a toroidal plasma producing device

International Nuclear Information System (INIS)

Clarke, J.F.

1978-01-01

A high beta plasma is produced in a plasma producing device of toroidal configuration by ohmic heating and auxiliary heating. The plasma pressure is continuously monitored and used in a control system to program the current in the poloidal field windings. Throughout the heating process, magnetic flux is conserved inside the plasma and the distortion of the flux surfaces drives a current in the plasma. As a consequence, the total current increases and the poloidal field windings are driven with an equal and opposing increasing current. The spatial distribution of the current in the poloidal field windings is determined by the plasma pressure. Plasma equilibrium is maintained thereby, and high temperature, high beta operation results

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

Electron cyclotron current drive efficiency in an axisymmetric tokamak

Energy Technology Data Exchange (ETDEWEB)

Gutierrez-Tapia, C.; Beltran-Plata, M. [Instituto Nacional de Investigaciones Nucleares, Dept. de Fisica, Mexico D.F. (Mexico)

2004-07-01

The neoclassical transport theory is applied to calculate electron cyclotron current drive (ECCD) efficiency in an axisymmetric tokamak in the low-collisionality regime. The tokamak ordering is used to obtain a system of equations that describe the dynamics of the plasma where the nonlinear ponderomotive (PM) force due to high-power radio-frequency (RF) waves is included. The PM force is produced around an electron cyclotron resonant surface at a specific poloidal location. The ECCD efficiency is analyzed in the cases of first and second harmonics (for different impinging angles of the RF waves) and it is validated using experimental parameter values from TCV and T-10 tokamaks. The results are in agreement with those obtained by means of Green's function techniques. (authors)

Inward particle transport at high collisionality in the Experimental Advanced Superconducting Tokamak

International Nuclear Information System (INIS)

Wang, G. Q.; Ma, J.; Weiland, J.; Zang, Q.

2013-01-01

We have made the first drift wave study of particle transport in the Experimental Advanced Superconducting Tokamak (Wan et al., Nucl. Fusion 49, 104011 (2009)). The results reveal that collisions make the particle flux more inward in the high collisionality regime. This can be traced back to effects that are quadratic in the collision frequency. The particle pinch is due to electron trapping which is not very efficient in the high collisionality regime so the approach to equilibrium is slow. We have included also the electron temperature gradient (ETG) mode to give the right electron temperature gradient, since the Trapped Electron Mode (TE mode) is weak in this regime. However, at the ETG mode number ions are Boltzmann distributed so the ETG mode does not give particle transport

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

NARCIS (Netherlands)

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

2015-01-01

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

ACHIEVING AND SUSTAINING STEADY-STATE ADVANCED TOKAMAK CONDITIONS ON DIII-D

International Nuclear Information System (INIS)

WADE, MR; MURAKAMI, M; BRENNAN, DP; CASPER, TA; FERRON, JR; GAROFALO, AM; GREENFIELD, CM; HYATT, AW; JAYAKUMAR, R; KINSEY, JE; LAHAYE, RJ; LAO, LL; LAZARUS, EA; LOHR, J; LUCE, TC; PETTY, CC; POLITZER, PA; PRATER, R; STRAIT, EJ; TURNBULL, AD; WATKINS, JG; WEST, WP

2002-01-01

Recent experiments on the DIII-D tokamak have demonstrated the feasibility of sustaining advanced tokamak conditions that combine high fusion power density (β > 4%), high bootstrap current fraction (f BS ∼ 65%), and high non-inductive current fractions (f NI ∼ 85%) for several energy confinement times. The duration of such conditions is limited only by resistive relaxation of the current density profile. Modeling studies indicate that the application of off-axis ECCD will be able to maintain a favorable current density profile for several seconds

Achieving and sustaining steady-state advanced tokamak conditions on DIII-D

International Nuclear Information System (INIS)

Wade, M.R.; Murakami, M.; Brennan, D.P.

2003-01-01

Recent experiments on the DIII-D tokamak have demonstrated the feasibility of sustaining advanced tokamak conditions that combine high fusion power density (β > 4%), high bootstrap current fraction (f BS ∼ 65%), and high non-inductive current fractions (f NI ∼85%) for several energy confinement times. The duration of such conditions is limited only by resistive relaxation of the current density profile. Modeling studies indicate that the application of off-axis ECCD will be able to maintain a favorable current density profile for several seconds. (author)

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1993-04-01

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

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)

Non-axisymmetric simulation of the vertical displacement event in tokamaks

International Nuclear Information System (INIS)

Lim, Y.Y.; Lee, J.K.; Shin, K.J.; Hur, M.S.

1999-01-01

Tokamak plasmas with highly elongated cross sections are subject to a vertical displacement event (VDE). The nonlinear magnetohydrodynamic (MHD) evolutions of tokamak plasmas during the VDE are simulated by a three-dimensional MHD code as a combination of N=0 and N=1 components. The nonlinear evolution during the VDE is strongly affected by the relative amplitude of the N=1 to the N=0 modes. (author)

Tokamaks (Second Edition)

Energy Technology Data Exchange (ETDEWEB)

Stott, Peter [JET, UK (United Kingdom)

1998-10-01

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

Tokamaks (Second Edition)

International Nuclear Information System (INIS)

Stott, Peter

1998-01-01

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

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

Theory of high-n toroidicity-induced shear Alfven eigenmode in tokamaks

International Nuclear Information System (INIS)

Fu, G.Y.; Cheng, C.Z.; Princeton Univ., NJ

1989-07-01

High-n WKB-ballooning mode equation is employed to study toroidicity-induced shear Alfven eigenmodes (TAE) in the δ - α space, where δ = (r/q)(dq/dr) is the magnetic shear, and α = -(2Rq 2 /B 2 )(dp/dr) is the normalized pressure gradient for tokamak plasmas. In the ballooning mode first stability region, TAE modes are found to exist only for α less than some critical value α c . We also find that these TAE modes reappear in the ballooning mode second stability region for bands of α values. The global envelope structures of these TAE modes are studied by WKB method and are found to be bounded radially if the local mode frequency has a maximum in radius. 15 refs., 14 figs

Modelling dust transport in tokamaks

International Nuclear Information System (INIS)

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

2008-01-01

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

Trade studies of plasma elongation for next-step tokamaks

International Nuclear Information System (INIS)

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

1988-09-01

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

New results from the Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.K.; Ananiev, A.S.; Amoskov, V.M.

2003-01-01

New results from the Globus-M spherical tokamak are presented. High plasma current of 0.36 MA, high toroidal magnetic field of 0.55 T and other important plasma characteristics were achieved. Described are the operational space and plasma stability limits in the OH regime. The factors limiting operational space (MHD instabilities, runaway electrons, etc.) are discussed. New experiments on plasma fuelling are described. First results of experiments with a coaxial plasma gun injector are presented. Initial results of a plasma - wall interaction study are outlined. First results obtained with new diagnostic tools installed on the tokamak are presented. An auxiliary heating system test was performed. Preliminary results of simulations and experiments are given. (author)

Stability analysis of internal ideal modes in low-shear tokamaks

International Nuclear Information System (INIS)

Wahlberg, C.; Graves, J. P.

2007-01-01

The stability of internal, ideal modes in tokamaks with low magnetic shear in the plasma core is analyzed. For equilibria with large aspect ratio, a parabolic pressure profile and a flat q profile in the core, an exact solution of the ideal magnetohydrodynamic (MHD) stability equations is found. The solution includes the eigenfunctions and the complete spectra of two distinctly different MHD phenomena: A family of fast-growing, Mercier-unstable global eigenmodes localized in a low-shear region with q 1 in the core. In the latter case the solution in addition includes one unstable eigenmode, if beta is larger than a critical value depending on the width of the low-shear region and on the q-profile in the edge region

Stability of negative central magnetic shear discharges in the DIII-D tokamak

International Nuclear Information System (INIS)

Strait, E.J.; Chu, M.S.; Ferron, J.R.

1996-12-01

Discharges with negative central magnetic shear (NCS) hold the promise of enhanced fusion performance in advanced tokamaks. However, stability to long wavelength magnetohydrodynamic modes is needed to take advantage of the improved confinement found in NCS discharges. The stability limits seen in DIII-D experiments depend on the pressure and current density profiles and are in good agreement with stability calculations. Discharges with a strongly peaked pressure profile reach a disruptive limit at low beta, β N = β (I/aB) -1 ≤ 2.5 (% m T/MA), caused by an n = 1 ideal internal kink mode or a global resistive instability close to the ideal stability limit. Discharges with a broad pressure profile reach a soft beta limit at significantly higher beta, β N = 4 to 5, usually caused by instabilities with n > 1 and usually driven near the edge of the plasma. With broad pressure profiles, the experimental stability limit is independent of the magnitude of negative shear but improves with the internal inductance, corresponding to lower current density near the edge of the plasma. Understanding of the stability limits in NCS discharges has led to record DIII-D fusion performance in discharges with a broad pressure profile and low edge current density

Fast wave current drive experiment on the DIII-D tokamak

International Nuclear Information System (INIS)

Petty, C.C.; Pinsker, R.I.; Chiu, S.C.; deGrassie, J.S.; Harvey, R.W.; Lohr, J.; Luce, T.C.; Mayberry, M.J.; Prater, R.; Porkolab, M.; Baity, F.W.; Goulding, R.H.; Hoffman, J.D.; James, R.A.; Kawashima, H.

1992-06-01

One method of radio-frequency heating which shows theoretical promise for both heating and current drive in tokamak plasmas is the direct absorption by electrons of the fast Alfven wave (FW). Electrons can directly absorb fast waves via electron Landau damping and transit-time magnetic pumping when the resonance condition ω - κ parallele υ parallele = O is satisfied. Since the FW accelerates electrons traveling the same toroidal direction as the wave, plasma current can be generated non-inductively by launching FW which propagate in one toroidal direction. Fast wave current drive (FWCD) is considered an attractive means of sustaining the plasma current in reactor-grade tokamaks due to teh potentially high current drive efficiency achievable and excellent penetration of the wave power to the high temperature plasma core. Ongoing experiments on the DIII-D tokamak are aimed at a demonstration of FWCD in the ion cyclotron range of frequencies (ICRF). Using frequencies in the ICRF avoids the possibility of mode conversion between the fast and slow wave branches which characterized early tokamak FWCD experiments in the lower hybrid range of frequencies. Previously on DIII-D, efficient direct electron heating by FW was found using symmetric (non-current drive) antenna phasing. However, high FWCD efficiencies are not expected due to the relatively low electron temperatures (compared to a reactor) in DIII-D

Magnetic confinement experiment -- 1: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1994-01-01

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

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

Extremely fast vertical displacement event induced by a plasma βp collapse in high βp tokamak disruptions

International Nuclear Information System (INIS)

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

1996-05-01

In a vertically elongated (κ ∼ 1.5), high β p (β p ∼ 1.7) tokamak with a resistive shell, extremely fast vertical displacement events (VDE's) induced by a model of strong β p collapse were found through computer simulations using the Tokamak Simulation Code. Although the plasma current quench, which had been shown to be the prime cause of VDE's in a relatively low β p tokamak (β p ∼ 0.2), was not observed during the VDE evolution, the observed growth rate of VDE's was almost five times (γ ∼ 655 sec -1 ) faster than the growth rate of the usual positional instability (γ ∼ 149 sec -1 ). The essential mechanism of the β p collapse-induced VDE was clarified to be the significant destabilization of positional instability due to a large and sudden degradation of the decay n-index in addition to a reduction of the stability index n s . It is pointed out that the shell-geometry characterizes the VDE dynamics, and that the VDE rate depends strongly both on the magnitude of the β p collapse and the n-index of the equilibria just before the β p collapse occurs. A new guide line for designing the fusion reactor is proposed with considering the impact of disruptions. (author)

Progress in application of high temperature superconductor in tokamak magnets

Czech Academy of Sciences Publication Activity Database

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

2013-01-01

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

Theory of energetic/alpha particle effects on magnetohydrodynamic modes in tokamaks

International Nuclear Information System (INIS)

Chen, L.; White, R.B.; Rewoldt, G.; Colestock, P.; Rutherford, P.H.; Chen, Y.P.; Ke, F.J.; Tsai, S.T.; Bussac, M.N.

1989-01-01

The presence of energetic particles is shown to qualitatively modify the stability properties of ideal as well as resistive magnetohydrodynamic (MHD) modes in tokamaks. Specifically, we demonstrate that, consistent with highpower ICRF heating experiments in JET, high energy trapped particles can effectively stabilize the sawtooth mode, providing a possible route to stable high current tokamak operation. An alternative stabilization scheme employing barely circulating energetic particles is also proposed. Finally, we present analytical and numerical studies on the excitations of high-n MHD modes via transit resonances with circulating alpha particles. 14 refs., 3 figs

Design and construction of Alborz tokamak vacuum vessel system

International Nuclear Information System (INIS)

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

2012-01-01

Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

Microwave Tokamak Experiment: Overview and status

International Nuclear Information System (INIS)

1990-05-01

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

Recent results from the DIII-D tokamak and implications for future devices

International Nuclear Information System (INIS)

Luxon, J.L.

1995-02-01

Improvements to the DIII-D tokamak have led to significant new research results and enhanced performance. These results provide important inputs to the design of next generation divertor systems including the upgrade of the DIII-D divertor. The use of graphite for the plasma facing components and careful wall preparation has enabled the routine achievement of regimes of enhanced energy confinement. In elongated discharges, triangularity has been found to be important in attaining good discharge performance as measured by the product of the normalized plasma pressure and the energy confinement time, βτ E This constrains the design of the divertor configuration (X-point location). Active pumping of the divertor region using an in-situ toroidal cryogenic pump has demonstrated control of the plasma density in H-mode discharges and allowed the dependence of confinement on plasma density and current to be separately determined. Helium removal from the plasma edge sufficient to achieve effective ash removal in reactor discharges has also been demonstrated using this pumping configuration. The reduction of the heat flux to the divertor plates has been demonstrated using two different techniques to increase the radiation in the boundary regions of the plasma and thus reduce the heat flux to the divertor plates; deuterium gas injection has been used to create a strongly radiating localized zone near the X-point, and impurity (neon) injection to enhance the radiation from the plasma mantle. Precise shaping of the plasma current profile has been found to be important in achieving enhanced tokamak performance. Transiently shaped current profiles have been used to demonstrate regimes of plasmas with high beta and good confinement. Control of the current profile also is important to sustaining the plasma in the Very High (VH)-mode of energy confinement

The basics of spherical tokamaks and progress in European research

International Nuclear Information System (INIS)

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

2003-01-01

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

Betting Against Beta

DEFF Research Database (Denmark)

Frazzini, Andrea; Heje Pedersen, Lasse

We present a model with leverage and margin constraints that vary across investors and time. We find evidence consistent with each of the model’s five central predictions: (1) Since constrained investors bid up high-beta assets, high beta is associated with low alpha, as we find empirically for U...... of the BAB factor is low; (4) Increased funding liquidity risk compresses betas toward one; (5) More constrained investors hold riskier assets........S. equities, 20 international equity markets, Treasury bonds, corporate bonds, and futures; (2) A betting-against-beta (BAB) factor, which is long leveraged low beta assets and short high-beta assets, produces significant positive risk-adjusted returns; (3) When funding constraints tighten, the return...

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation