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Author

R.D. Stambaugh

Other affiliations: University of Texas at Austin
Bio: R.D. Stambaugh is an academic researcher from General Atomics. The author has contributed to research in topics: Divertor & Tokamak. The author has an hindex of 33, co-authored 96 publications receiving 4707 citations. Previous affiliations of R.D. Stambaugh include University of Texas at Austin.
Topics: Divertor, Tokamak, DIII-D, Fusion power, Plasma


Papers
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Journal ArticleDOI
TL;DR: In this paper, 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.
Abstract: 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.

1,587 citations

Journal ArticleDOI
TL;DR: In this article, the ITER design requirements were reviewed and as appropriate updated and the focus of this paper will be on recent work affecting ITER with special emphasis on topics affecting near-term procurement arrangements.
Abstract: As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

218 citations

Journal ArticleDOI
TL;DR: In this paper, 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.
Abstract: 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.

214 citations

Journal ArticleDOI
TL;DR: In this article, the basic paradigm of electric field shear stabilization has been discussed in the light of the most recent data, and the experimental results with various theories have been compared.
Abstract: Combined theoretical and experimental work has resulted in the creation of a paradigm which has allowed semi-quantitative understanding of the edge confinement improvement that occurs in the H-mode. Shear in the E*B flow of the fluctuations in the plasma edge can lead to decorrelation of the fluctuations, decreased radial correlation lengths and reduced turbulent transport. Changes in the radial electric field, the density fluctuations and the edge transport consistent with shear stabilization of turbulence have been seen in several tokamaks. The purpose of this paper is to discuss the most recent data in the light of the basic paradigm of electric field shear stabilization and to critically compare the experimental results with various theories.

199 citations

Journal ArticleDOI
TL;DR: In this paper, a simple theoretical model of the nonlinear stability of a tearing mode under the effect of externally applied resonant field perturbations yields a critical island width for the onset of a locked mode.
Abstract: Locked n = 1 tearing modes are observed over a wide range of parameter space in DIII-D and other tokamaks. Much of the difficulty with low density operation is attributed to locked modes and they are also observed as precursors of density limit disruptions. From observations of a consistent locked mode toroidal phase, it appears that the modes are locking to a small field perturbation caused by slight irregularities in the location of one or more of the vertical field coils with respect to the toroidal field coil. By intentionally producing an n = 1 field with an external coil, it was possible to influence the onset of locked modes in low q, low density plasmas. The result is a significantly expanded or reduced stable operating parameter space, depending on the polarity and magnitude of the external perturbation applied and whether or not intrinsic field errors are reduced or increased. A simple theoretical model of the non-linear stability of a tearing mode under the effect of externally applied resonant field perturbations yields a critical island width for the onset of a locked mode. The island widths, computed by field line tracing, for combinations of intrinsic field errors and 'n = 1 coil' fields are in qualitative agreement with the critical island width for instability.

152 citations


Cited by
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Journal Article
A. Gibson, Tadashi Sekiguchi, K. Lackner1, S. Bodner, R. Hancox 
TL;DR: In this paper, the first experiments in JET have been described, which show that this large tokamak behaves in a similar manner to smaller tokak, but with correspondingly improved plasma parameters.
Abstract: 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 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.

3,647 citations

Journal ArticleDOI
TL;DR: In this paper, 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.
Abstract: 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.

1,587 citations

Journal ArticleDOI
TL;DR: The ExB shear stabilization model was originally developed to explain the transport barrier formed at the plasma edge in tokamaks after the L (low) to H (high) transition as mentioned in this paper.
Abstract: One of the scientific success stories of fusion research over the past decade is the development of the ExB shear stabilization model to explain the formation of transport barriers in magnetic confinement devices. This model was originally developed to explain the transport barrier formed at the plasma edge in tokamaks after the L (low) to H (high) transition. This concept has the universality needed to explain the edge transport barriers seen in limiter and divertor tokamaks, stellarators, and mirror machines. More recently, this model has been applied to explain the further confinement improvement from H (high)-mode to VH (very high)-mode seen in some tokamaks, where the edge transport barrier becomes wider. Most recently, this paradigm has been applied to the core transport barriers formed in plasmas with negative or low magnetic shear in the plasma core. These examples of confinement improvement are of considerable physical interest; it is not often that a system self-organizes to a higher energy state with reduced turbulence and transport when an additional source of free energy is applied to it. The transport decrease that is associated with ExB velocity shear effects also has significant practical consequences for fusion research. The fundamental physics involved in transport reduction is the effect of ExB shear on the growth, radial extent and phase correlation of turbulent eddies in the plasma. The same fundamental transport reduction process can be operational in various portions of the plasma because there are a number ways to change the radial electric field Er. An important theme in this area is the synergistic effect of ExB velocity shear and magnetic shear. Although the ExB velocity shear appears to have an effect on broader classes of microturbulence, magnetic shear can mitigate some potentially harmful effects of ExB velocity shear and facilitate turbulence stabilization.

1,251 citations

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

1,187 citations

Journal ArticleDOI
TL;DR: A review of recent advances in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed in this paper.
Abstract: Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

1,051 citations