FIRST EXPERIMENTS IN JET. Results obtained from JET since June 1983 are described which show that this large tokamak behaves in a similar manner to smaller tokamaks, but with correspondingly improved plasma parameters. Long-duration hydrogen and deuterium plasmas (>10 s) have been obtained with electron temperatures reaching > 4 keV for power dissipations < 3 MW and with * Euratom-IPP Association, Institut fur Plasmaphysik, Garching, Federal Republic of Germany. ** Euratom-ENEA Association, Centro di Frascati, Italy. *** Euratom-UKAEA Association, Culham Laboratory, Abingdon, Oxfordshire, United Kingdom. **** University of Dusseldorf, Dusseldorf, Federal Republic of Germany. + Euratom-Ris0 Association, Ris<f> National Laboratory, Roskilde, Denmark. ++ Euratom-CNR Association, Istituto di Física del Plasma, Milan, Italy. +++ Imperial College of Science and Technology, University of London, London, United Kingdom. ++++ Euratom-FOM Association, FOM Instituut voor Plasmafysica,. Nieuwegein, Netherlands. ® Euratom-Suisse Association, Centre de Recherches en Physique des Plasmas, Lausanne, Switzerland.

Plasma Physics and Controlled Nuclear Fusion Research

An efficient method is given to reconstruct the current profile parameters, the plasma shape, and a current profile consistent with the magnetohydrodynamic equilibrium constraint from external magnetic measurements, based on a Picard iteration approach which approximately conserves the measurements. Computational efforts are reduced by parametrizing the current profile linearly in terms of a number of physical parameters. Results of detailed comparative calculations and a sensitivity study are described. Illustrative calculations to reconstruct the current profiles and plasma shapes in ohmically and auxiliarily heated Doublet III plasmas are given which show many interesting features of the current profiles.

https://iopscience.iop.org/article/10.1088/0029-5515/25/11/007/pdf

Reconstruction of current profile parameters and plasma shapes in tokamaks

An efficient method is given for self-consistent reconstruction of the tokamak current profiles and their associated magnetic topology using the magnetohydrodynamic (MHD) equilibrium constraint from external magnetic measurements, kinetic profile measurements, internal poloidal magnetic field measurements, and topological information from soft X-ray (SXR) measurements. Illustrative examples for beam heated H-mode divertor discharges in the DIII-D tokamak are presented, using the experimentally measured kinetic profile information and external magnetic data from the existing diagnostics. Comparative reconstructions of the current profile using various combinations of diagnostics are given. Also presented is an alternative magnetic analysis method in which the MHD equilibrium is reconstructed using external magnetic data and a constraint on the edge pressure gradient. The results of a sensitivity study are given which show that the axial safety factor q(0) can be more accurately determined when additional information from internal poloidal magnetic measurements is used in conjunction with the external magnetic, kinetic and SXR topological data.

https://iopscience.iop.org/article/10.1088/0029-5515/30/6/006/pdf

Equilibrium analysis of current profiles in tokamaks

This is a review of what is known about fluctuations and anomalous transport processes in tokamaks. It mostly considers experimental results obtained after, and not included in, the reviews of Liewer [Nucl. Fusion 25, 543 (1985)], Robinson [in Turbulence and Anomalous Transport in Magnetized Plasmas (Ecole Polytechnique, Palaiseau, France, 1986), p. 21], and Surko [in Turbulence and Anomalous Transport in Magnetized Plasmas (Ecole Polytechnique, Palaiseau, France, 1986), p. 93]. Therefore much of the pioneering work in the field is not covered. Emphasis is placed on results where comparisons between fluctuations and transport properties have been attempted, particularly from the tokamak TEXT [Nucl. Technol./Fusion 1, 479 (1981)]. A brief comparison of experimentally measured total fluxes with the predictions of neoclassical theory demonstrates that transport is often anomalous; fluctuations are thought to be the cause.The measurements necessary to determine any such fluctuation‐driven fluxes are described. The diagnostics used to measure these quantities, together with some of the statistical techniques employed to analyze the data, are outlined. In the plasma edge detailed measurements of the quantities required to directly determine the fluctuation‐driven fluxes are available. The total and fluctuation‐driven fluxes are compared: the result emphasizes the importance of edge turbulence. No model adequately describes all the measured properties. In the confinement region experimental observations are presently restricted to measurements of density and potential fluctuations and their correlations. Various distinct turbulence features that have been observed are described, and their characteristics compared with the predictions of various models. Correlations observed between these fluctuations and plasma transport properties are summarized. A separate section on magnetic fluctuations shows there is very little information available inside the plasma, generally prohibiting quantified comparisons between fluctuation levels and transport. Both coherent and turbulent magnetic fluctuations are addressed, and the differences between low and high plasma pressure (low and high beta) are noted. The contributions of alternate confinement devices, such as stellarators and reversed field pinches, to understanding tokamak anomalous transport are discussed. Finally, future directions are proposed.

Fluctuations and anomalous transport in tokamaks

Early operation of the Alcator‐C‐MOD tokamak [I.H. Hutchinson, Proceedings of IEEE 13th Symposium on Fusion Engineering, Knoxville, TN, edited by M. Lubell, M. Nestor, and S. Vaughan (Institute of Electrical and Electronic Engineers, New York, 1990), Vol. 1, p. 13] is surveyed. Reliable operation, with plasma current up to 1 MA, has been obtained, despite the massive conducting superstructure and the associated error fields. However, vertical disruptions are not slowed by the long vessel time constant. With pellet fueling, peak densities up to 9×1020 m−3 have been attained and ‘‘snakes’’ are often seen. Initial characterization of divertor and scrape‐off layer is presented and indicates approximately Bohm diffusion. The edge plasma shows a wealth of marfe‐like phenomena, including a transition to detachment from the divertor plates with accompanying radiative divertor regions. Energy confinement generally appears to exceed the expectations of neo‐Alcator scaling. A transition to Ohmic H mode has been observed. Ion cyclotron heating experiments have demonstrated good power coupling, in agreement with theory.

/pdf/first-results-from-alcator-c-mod-2x43pp5oly.pdf

First results from Alcator‐C‐MOD*

Integral relations for the average poloidal beta p and the plasma internal inductance li are derived from the magnetohydrodynamic (MHD) equilibrium equation for an axisymmetric torus. The volume-dependent parameters that appear depend only weakly on the actual current density distribution inside the plasma and can be evaluated approximately, given the plasma shape and boundary poloidal magnetic field. In practice, these can be accurately and efficiently obtained for both diverted and limited plasmas from measured external poloidal magnetic field and flux values by approximating the plasma current distribution using a few filaments or distributed sources. For a tokamak plasma with a non-circular cross-section of sufficient elongation, p and li can then be approximately determined separately. This is demonstrated for analytic equilibria of known shape as well as for actual Doublet III (D-III) plasmas for which p and li have been determined by using other methods. Results of a sensitivity study are described.

https://iopscience.iop.org/article/10.1088/0029-5515/25/10/004/pdf

Separation of β̄p and ℓi in tokamaks of non-circular cross-section

Measurements of ion temperature profiles, obtained on a single-shot basis, have been coupled with transport analysis to obtain spatial profiles of the ion thermal diffusivity ?i for a series of 18 neutral-beam-heated discharges in the Doublet III tokamak. The expanded boundary divertor discharges examined in this study had the following range of parameters: Bt = 2.53 T, Ip = 0.35?0.89 MA, vertical elongation = 1.4?1.9, n?e = (3.7?7.9) ? 1013 cm?3, PT = 3.5?6.6 MW, and ?i* = 0.03?0.22 , for r/a = 0.2?0.7. The experimentally inferred values of ?i have been compared with the values of the ion thermal diffusivity predicted by neoclassical theory. On the basis of a detailed error analysis, the results for all discharges examined are the same: ?i is approximately equal to near the plasma centre, but for r/a between about 0.2 and 0.8, ?i is both larger than and almost certainly has a spatial variation different from that of . The differences between ?i and are larger than can be explained by errors in the experimental measurements and indicate that there are processes operative for the loss of ion heat which are not included in the usual transport models for ion confinement.

Experimentally inferred ion thermal diffusivity profiles in the Doublet III tokamak: Comparison with neoclassical theory

Neutral-beam-heated expanded-boundary (XB) divertor discharges have been obtained in Doublet III with high heating efficiency for wide ranges of plasma parameters (Ip: 300–800 kA; BT: 8–24kG; e :(2–10) × 1013cm−3; Pb 3 MW), a mild deterioration of the energy confinement time has been observed.

http://iopscience.iop.org/0029-5515/25/1/005/pdf/0029-5515_25_1_005.pdf

W. Pfeiffer

Papers

Reconstruction of current profile parameters and plasma shapes in tokamaks

Separation of β̄p and ℓi in tokamaks of non-circular cross-section

Experimentally inferred ion thermal diffusivity profiles in the Doublet III tokamak: Comparison with neoclassical theory

A regime of improved energy confinement in beam-heated expanded-boundary discharges in Doublet III