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Recent progress on ATF

Description: The ATF experiment will test improvements to high-beta, steady-state toroidal confinement using external helical fields. The device design has been optimized to (1) provide direct access to the high-beta second-stability regime, (2) have sufficient flexibility to study a large range of toroidal configurations both with and without plasma current, (3) test the reduction of low-collisionality transport by EXB drifts induced by the self-consistent radial electric field, and (4) permit steady-state, high-beta operation without disruptions. Continued physics studied at ORNL and recent results from foreign stellarator experiments have increased confidence in ATF performance.
Date: January 1, 1984
Creator: Neilson, G.H.
Partner: UNT Libraries Government Documents Department

Experimental results on tokamak beta limits from ISX-B

Description: It is observed that the expression A<..beta..>q/sub psi//kappa=18% defines one of the operational boundaries for ISX-B and other tokamaks. Although this expression closely coincides with the simplest expressions for ideal MHD instability thresholds, it is not obvious that it constitutes a real tokamak beta limit. Confinement degradation is observed well within this operating boundary, and there is neither experimental evidence nor a satisfactory theoretical model connecting this phenomenon with ideal MHD activity.
Date: January 1, 1984
Creator: Neilson, G.H.
Partner: UNT Libraries Government Documents Department

Finite order polynomial moment solutions of the homogeneous Grad-Shafranov equation

Description: A method for generating the finite positive order polynomial moment solutions of the homogeneous Grad-Shafranov equation to arbitrary order and the explicit form of the first few moments are given. A criticism of the method is discussed, and several practical examples are given.
Date: February 1, 1984
Creator: Reusch, M.F. & Neilson, G.H.
Partner: UNT Libraries Government Documents Department

Feedback control modeling of plasma position and current during intense heating in ISX-B

Description: The ISX-B Tokamak at ORNL is designed to have 1.8 MW (and eventually 3 MW) of neutral beam power injected to heat the plasma. This power may raise the anti ..beta.. of the plasma to over 5% in less than 50 msec if the plasma is MHD stable. The results of a numerical simulation of the feedback control system and poloidal coil power supplies necessary to control the resulting noncircular (D-shaped or elliptical) plasma are presented. The resulting feedback control system is shown to be straightforward, although nonlinear voltage-current dependence is assumed in the power supplies. The required power supplied to the poloidal coils in order to contain the plasma under the high heating rates is estimated.
Date: August 1, 1979
Creator: Charlton, L.A.; Swain, D.W. & Neilson, G.H.
Partner: UNT Libraries Government Documents Department

Injection-dominated tokamak experiments at ORNL

Description: Experiments on the Oak Ridge Tokamak (ORMAK) have demonstrated ion and electron heating and improvements in anti ..beta../sub T/, anti n/sub e/, and q(a/sub l/) with neutral beam injection. They have also emphasized the need for low impurity levels in injected plasmas and the advantages of co- as opposed to counterinjection. These results, together with the favorable confinement and impurity results obtained in the Impurity Study Experiment (ISX-A) are encouraging in terms of injection-dominated, high beta experiments planned for ISX-B. This device will use 3.0 MW of injection power to study beta limits, confinement, heating, and impurity control in noncircular cross-section plasmas.
Date: January 1, 1978
Creator: Neilson, G H; Lyon, J F & Murakami, M
Partner: UNT Libraries Government Documents Department

Determination of plasma shape from poloidal field measurements on ISX-B

Description: The ISX-B tokamak has a poloidal coil system designed to produce circular, elliptical, and D-shaped plasmas. Plasma shape and low-order multipole moments of the plasma current distribution are determined from experimental measurements of B/sub Z/, B/sub R/, and/or psi around the periphery of the vacuum chamber. The experimental arrangement and method of analysis of results, using a least squares method to fit the data points to a finite current filament model, are described in this report. Plasma shape results for circular and D-shaped plasmas with b/a less than or equal to 1.5 and an analysis of the sensitivity of the technique to measurement errors are presented. The results indicate that this method gives accurate measurements of the plasma boundary and is relatively insensitivie to errors.
Date: March 1, 1980
Creator: Swain, D.W.; Bates, S.; Neilson, G.H. & Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Alternative poloidal field configurations for ITER

Description: The US Home Team has investigated the physics and engineering issues for two alternate poloidal field coil configurations for ITER. The first is called the Segmented CS configuration, where all of the solenoid modules are pancake-wound. The second option, termed the Hybrid CS configuration, utilizes a layer-wound central module and pancake-wound end modules. Performance comparisons are presented for the baseline design and the two alternate PF configurations, characterizing the 21 MA reference scenario. Alternate operating modes such as reverse-shear operation and a 17 MA driven mode were evaluated, but are not reported here.
Date: September 2, 1997
Creator: Bulmer, R.H. & Neilson, G.H.
Partner: UNT Libraries Government Documents Department

Plasma vertical stability and feedback control for TPX

Description: The n=0 axisymmetric vertical stability and vertical position control have been examined for the Tokamak Physics Experiment. The passive stabilization is accomplished by using stabilizer plates close to the plasma. The present configuration is found to provide robust stability over wide range of plasma parameters. The active feed-back control of the plasma vertical position is done using coils located inside the vacuum vessel. These are required to control random disturbances leading to {le} 1.0 cm RMS displacements from the midplane, and acceptable coil currents and voltages are found.
Date: November 1, 1993
Creator: Kessel, C. E; Jardin, S. C. & Neilson, G. H.
Partner: UNT Libraries Government Documents Department

Physics Basis for High-Beta, Low-Aspect-Ratio Stellarator Experiments

Description: High-beta, low-aspect-ratio (compact) stellarators are promising solutions to the problem of developing a magnetic plasma configuration for magnetic fusion power plants that can be sustained in steady-state without disrupting. These concepts combine features of stellarators and advanced tokamaks and have aspect ratios similar to those of tokamaks (2-4). They are based on computed plasma configurations that are shaped in three dimensions to provide desired stability and transport properties. Experiments are planned as part of a program to develop this concept. A beta = 4% quasi-axisymmetric plasma configuration has been evaluated for the National Compact Stellarator Experiment (NCSX). It has a substantial bootstrap current and is shaped to stabilize ballooning, external kink, vertical, and neoclassical tearing modes without feedback or close-fitting conductors. Quasi-omnigeneous plasma configurations stable to ballooning modes at beta = 4% have been evaluated for the Quasi-Omnigeneous Stellarator (QOS) experiment. These equilibria have relatively low bootstrap currents and are insensitive to changes in beta. Coil configurations have been calculated that reconstruct these plasma configurations, preserving their important physics properties. Theory- and experiment-based confinement analyses are used to evaluate the technical capabilities needed to reach target plasma conditions. The physics basis for these complementary experiments is described.
Date: November 1, 1999
Creator: Brooks, A.; Reiman, A.H.; Neilson, G.H.; Zarnstorff, M.C. & al, et
Partner: UNT Libraries Government Documents Department

Physics Design of the National Compact Stellarator Experiment

Description: Compact quasi-axisymmetric stellarators offer the possibility of combining the steady-state low-recirculating power, external control, and disruption resilience of previous stellarators with the low-aspect ratio, high beta-limit, and good confinement of advanced tokamaks. Quasi-axisymmetric equilibria have been developed for the proposed National Compact Stellarator Experiment (NCSX) with average aspect ratio approximately 4.4 and average elongation approximately 1.8. Even with bootstrap-current consistent profiles, they are passively stable to the ballooning, kink, vertical, Mercier, and neoclassical-tearing modes for b &gt; 4%, without the need for external feedback or conducting walls. The bootstrap current generates only 1/4 of the magnetic rotational transform at b = 4% (the rest is from the coils). Transport simulations show adequate fast-ion confinement and thermal neoclassical transport similar to equivalent tokamaks. Modular coils have been designed which reproduce the physics properties, provide good flux surfaces, and allow flexible variation of the plasma shape to control the predicted MHD stability and transport properties.
Date: February 21, 2002
Creator: Neilson, G.H.; Zarnstorff, M.C.; Lyon, J.F. & Team, the NCSX
Partner: UNT Libraries Government Documents Department

Comparison of Options for a Pilot Plant Fusion Nuclear Mission

Description: 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 pilot plant mission encompassed component test and fusion nuclear science missions plus the requirement to produce net electricity with high availability in a device designed to be prototypical of the commercial device. Three magnetic configuration options were developed around this mission: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS). With the completion of the study and separate documentation of each design option a question can now be posed; how do the different designs compare with each other as candidates for meeting the pilot plant mission? In a pro/con format this paper will examine the key arguments for and against the AT, ST and CS magnetic configurations. Key topics addressed include: plasma parameters, device configurations, size and weight comparisons, diagnostic issues, maintenance schemes, availability influences and possible test cell arrangement schemes.
Date: August 27, 2012
Creator: Brown, T; Goldston, R J; El-Guebaly, L; Kessel, C; Neilson, G H; Malang, S et al.
Partner: UNT Libraries Government Documents Department

Disruptions, Disruptivity, and Safer Operating Windows in the High-β Spherical Torus NSTX

Description: 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. 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. This paper reviews the AT pilot plant design, detailing the selected maintenance approach, the device arrangement and sizing of the in-vessel components. Details of interfacing auxiliary systems and services that impact the ability to achieve high availability operations will also be discussed.
Date: September 26, 2012
Creator: Brown, T; Goldston, R J; El-Guebaly, L; Kessel, C; Neilson, G H; Malang, S et al.
Partner: UNT Libraries Government Documents Department

Stabilization of the external kink and other MHD issues. Summary report

Description: An MHD workshop entitled ``Stabilization of the External Kink and Other MHD Issues`` was held June 1993. This is a summary report of activities at that workshop, structured to respond to the three questions in the charge (letter from J. Willis). The experimental and theoretical status of these issues, and the R&D needs in each area, are addressed. We discuss the potential impact on the TPX and ITER programs of these issues. The workshop participants came from a broad and diverse range of institutions in the fusion program, including international participants. As a result, we believe the summary here reflects some consensus of the community on these very important program issues, and that the TPX and ITER programs will benefit from these discussions. The title of the workshop was chosen to indicate both our knowledge and our uncertainty of MHD phenomena limiting {beta} and causing disruptions in tokamaks. The purpose was to bring together theorists and experimentalist in order to assess our current understanding of the external kink instability at high {beta}, and to assess the potential for passive or active stabilization of the dominant modes. We also outlined the R&D needed for TPX and other future devices. Not only was the preworkshop theory clearly presented, but significant new theoretical results were described for the first time, emphasizing the roles of the resistivity of the cold edge plasma and of the plasma toroidicity in the stability criteria. Excellent reviews of the effects of the vessel walls on plasma stability were given as related to the DIII-D, TFTR, JET, PBX-M, and HBT-EP experiments. These results are generally consistent with the more complete theory.
Date: August 13, 1993
Creator: Freidberg, J. P.; Goldston, R. J.; Jardin, S. C.; Neilson, G. H.; Rosenbluth, M. N.; Taylor, T. S. et al.
Partner: UNT Libraries Government Documents Department

Physics design requirements for the Tokamak Physics Experiment (TPX)

Description: 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.
Date: November 1, 1993
Creator: Neilson, G. H.; Goldston, R. J.; Jardin, S. C.; Reiersen, W. T.; Nevins, W. M.; Porkolab, M. et al.
Partner: UNT Libraries Government Documents Department

Progress Toward Attractive Stellarators

Description: The quasi-axisymmetric stellarator (QAS) concept offers a promising path to a more compact stellarator reactor, closer in linear dimensions to tokamak reactors than previous stellarator designs. Concept improvements are needed, however, to make it more maintainable and more compatible with high plant availability. Using the ARIES-CS design as a starting point, compact stellarator designs with improved maintenance characteristics have been developed. While the ARIES-CS features a through-the-port maintenance scheme, we have investigated configuration changes to enable a sector-maintenance approach, as envisioned for example in ARIES AT. Three approaches are reported. The first is to make tradeoffs within the QAS design space, giving greater emphasis to maintainability criteria. The second approach is to improve the optimization tools to more accurately and efficiently target the physics properties of importance. The third is to employ a hybrid coil topology, so that the plasma shaping functions of the main coils are shared more optimally, either with passive conductors made of high-temperature superconductor or with local compensation coils, allowing the main coils to become simpler. Optimization tools are being improved to test these approaches.
Date: January 5, 2011
Creator: Neilson, G. H.; Brown, T. G.; Gates, D. A.; Lu, K. P.; Zarnstorff, M. C.; Boozer, A. H. et al.
Partner: UNT Libraries Government Documents Department

Mission and Readiness Assessment for Fusion Nuclear Facilities

Description: Magnetic fusion development toward DEMO will most likely require a number of fusion nuclear facilities (FNF), intermediate between ITER and DEMO, to test and validate plasma and nuclear technologies and to advance the level of system integration. The FNF mission space is wide, ranging from basic materials research to net electricity demonstration, so there is correspondingly a choice among machine options, scope, and risk in planning such a step. Readiness requirements to proceed with a DEMO are examined, and two FNF options are assessed in terms of the contributions they would make to closing DEMO readiness gaps, and their readiness to themselves proceed with engineering design about ten years from now. An advanced tokamak (AT) pilot plant with superconducting coils and a mission to demonstrate net electricity generation would go a long way toward DEMO. As a next step, however, a pilot plant would entail greater risk than a copper-coil FNSF-AT with its more focussed mission and technology requirements. The stellarator path to DEMO is briefly discussed. Regardless of the choice of FNF option, an accompanying science and technology development program, also aimed at DEMO readiness, is absolutely essential.
Date: December 12, 2012
Creator: Neilson, G. H.; Brown, T. G.; Gates, D. A.; Kessel, C. E.; Menard, J. E.; Prager, S. C. et al.
Partner: UNT Libraries Government Documents Department

Progress In NCSX and QPS Design and Construction

Description: The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The stellarator core is designed to produce a compact 3-D plasma that combines stellarator and tokamak physics advantages. The engineering challenges of NCSX stem from its complex geometry. From the project&#x27;s start in April, 2003 to September, 2004, the fabrication specifications for the project&#x27;s two long-lead components, the modular coil winding forms and the vacuum vessel, were developed. An industrial manufacturing R&amp;D program refined the processes for their fabrication as well as production cost and schedule estimates. The project passed a series of reviews and established its performance baseline with the Department of Energy. In September 2004, fabrication was approved and contracts for these components were awarded. The suppliers have completed the engineering and tooling preparations and are in production. Meanwhile, the project completed preparations for winding the coils at PPPL by installing a coil manufacturing facility and developing all necessary processes through R&amp;D. The main activities for the next two years will be component manufacture, coil winding, and sub-assembly of the vacuum vessel and coil subsets. Machine sector sub-assembly, machine assembly, and testing will follow, leading to First Plasma in July 2009.
Date: October 20, 2005
Creator: Reiersen, W.; Heitzenroeder, P.; Neilson, G. H.; Nelson, B.; Zarnstorff, M.; Brown, T. et al.
Partner: UNT Libraries Government Documents Department

Design Of JET ELM Control Coils For Operation At 350 C

Description: A study has confirmed the feasibility of designing, fabricating and installing resonant magnetic field perturbation (RMP) coils in JET1 with the objective of controlling edge localized modes (ELM). A system of two rows of in-vessel coils, above the machine midplane, has been chosen as it not only can investigate the physics of and achieve the empirical criteria for ELM suppression, but also permits variation of the spectra allowing for comparison with other experiments. These coils present several engineering challenges. Conditions in JET necessitate the installation of these coils via remote handling, which will impose weight, dimensional and logistical limitations. And while the encased coils are designed to be conventionally wound and bonded, they will not have the usual benefit of active cooling. Accordingly, coil temperatures are expected to reach 350 C during bakeout as well as during plasma operations. These elevated temperatures are beyond the safe operating limits of conventional OFHC copper and the epoxies that bond and insulate the turns of typical coils. This has necessitated the use of an alternative copper alloy conductor C18150 (CuCrZr). More importantly, an alternative to epoxy had to be found. An R&amp;D program was initiated to find the best available insulating and bonding material. The search included polyimides and ceramic polymers. The scope and status of this R&amp;D program, as well as the critical engineering issues encountered to date are reviewed and discussed.
Date: September 20, 2010
Creator: Zatz, I. J.; Brooks, A.; Cole, M.; Neilson, G. H.; Lowry, C.; Mardenfeld, M. et al.
Partner: UNT Libraries Government Documents Department

Plasma shape control calculations for BPX divertor design

Description: The Burning Plasma Experiment (BPX) divertor is to be capable of withstanding heat loads corresponding to ignited operation and 500 MW of fusion power for a current rise time and flattop lasting several seconds. The poloidal field (PF), diagnostic, and feedback equilibrium control systems must provide precise X-point position control in order to sweep the separatrices across the divertor target surface and optimally distribute the heat loads. A control matrix MHD equilibrium code, BEQ, and the Tokamak Simulation Code (TSC) are used to compute preprogrammed double-null (DN) divertor sweep trajectories that maximize sweep distance while simultaneously satisfying a set of strict constraints: minimum lengths of the field lines between the X-point and strike points, minimum spacing between the inboard plasma edge and the limiter, maximum spacing between the outboard plasma edge and the ICRF antennas, minimum safety factor, and linked poloidal flux. A sequence of DN diverted equilibria and a consistent TSC fiducial discharge simulation are used in evaluating the performance of the BPX divertor shape and possible modifications. 5 refs., 10 figs.
Date: January 1, 1991
Creator: Strickler, D.J.; Neilson, G.H. (Oak Ridge National Lab., TN (United States)); Jardin, S.C. & Pomphrey, N. (Princeton Univ., NJ (United States). Plasma Physics Lab.)
Partner: UNT Libraries Government Documents Department

Flexibility and Robustness Calculations for NCSX

Description: The National Compact Stellarator Experiment (NCSX) will study the physics of low aspect ratio, high beta quasi-axisymmetric stellarators. In order to achieve the scientific goals of the NCSX mission, the device must be capable of supporting a wide range of variations in plasma configuration about a reference equilibrium. Numerical experiments are presented which demonstrate this capability.
Date: June 6, 2002
Creator: Pomphrey, N.; Hatcher, R.; Hirshman, S.P.; Hudson, S.; Ku, L-P; Lazarus, E.A. et al.
Partner: UNT Libraries Government Documents Department

Component Manufacturing Development for the National Compact Stellarator Experiment (NCSX)

Description: NCSX [National Compact Stellarator Experiment] is the first of a new class of stellarators called compact stellarators which hold the promise of retaining the steady state feature of the stellarator but at a much lower aspect ratio and using a quasi-axisymmetric magnetic field to obtain tokamak-like performance. Although much of NCSX is conventional in design and construction, the vacuum vessel and modular coils provide significant engineering challenges due to their complex shapes, need for high dimensional accuracy, and the high current density required in the modular coils due space constraints. Consequently, a three-phase development program has been undertaken. In the first phase, laboratory/industrial studies were performed during the development of the conceptual design to permit advances in manufacturing technology to be incorporated into NCSX's plans. In the second phase, full-scale prototype modular coil winding forms, compacted cable conductors, and 20 degree sectors of the vacuum vessel were fabricated in industry. In parallel, the NCSX project team undertook R&amp;D studies that focused on the windings. The third (production) phase began in September 2004. First plasma is scheduled for January 2008.
Date: October 28, 2004
Creator: Heitzenroeder, P.J.; Brown, T.G.; Chrzanowski, J.H.; Cole, M.J.; Goranson, P.L.; Neilson, G.H. et al.
Partner: UNT Libraries Government Documents Department