skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Progress towards steady state on NSTX

Authors:
;
Publication Date:
Research Org.:
Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
898023
Report Number(s):
UCRL-JRNL-221039
DOE Contract Number:
W-7405-ENG-48
Resource Type:
Journal Article
Resource Relation:
Journal Name: Nuclear Fusion, vol. 46, n/a, May 3, 2006, S22-S28
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION

Citation Formats

Gates, D A, and al., e. Progress towards steady state on NSTX. United States: N. p., 2006. Web. doi:10.1088/0029-5515/46/3/S04.
Gates, D A, & al., e. Progress towards steady state on NSTX. United States. doi:10.1088/0029-5515/46/3/S04.
Gates, D A, and al., e. Mon . "Progress towards steady state on NSTX". United States. doi:10.1088/0029-5515/46/3/S04. https://www.osti.gov/servlets/purl/898023.
@article{osti_898023,
title = {Progress towards steady state on NSTX},
author = {Gates, D A and al., e},
abstractNote = {},
doi = {10.1088/0029-5515/46/3/S04},
journal = {Nuclear Fusion, vol. 46, n/a, May 3, 2006, S22-S28},
number = ,
volume = ,
place = {United States},
year = {Mon May 01 00:00:00 EDT 2006},
month = {Mon May 01 00:00:00 EDT 2006}
}
  • In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal beta and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from kappa similar to 2.1 to kappa similar to 2.6-approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal beta for long pulse discharges. The increase in beta associated with an increase in plasma current at nearly fixed poloidal beta, which enablesmore » higher beta(1), with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of I MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (delta similar to 0.8) at elevated elongation (kappa similar to 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented.« less
  • 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(BS) >= 60%, q(95) similar to 4-5 and G = beta H-N(scaling)/q(95)(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,more » 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 x B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at beta N >= 4 with q(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(N) = 4, H-89 = 2.5, with f(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.« less
  • In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal {beta} and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on the National Spherical Torus Experiment (NSTX) has been raised from {kappa} {approx} 2.1 to {kappa} {approx} 2.6--approximately a 25% increase. This increase in elongation has lead to a doubling increase in the toroidal {beta} for long-pulse discharges. The increase in {beta} is associated with an increase in plasma current at nearly fixed poloidalmore » {beta}, which enables higher {beta}{sub t} with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 second. Data is presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption during and to delay the onset of MHD instabilities. A modeled integrated scenario, which has 100% non-inductive current drive with very high toroidal {beta}, will also be presented. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity ({delta} {approx} 0.8) at elevated elongation ({kappa} {approx} 2.5). The other main requirement for steady state on NSTX is the ability to drive a fraction of the total plasma current with radio-frequency waves. The results of High Harmonic Fast Wave heating and current drive studies as well as electron Bernstein Wave emission studies will be presented.« less
  • Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S ≡ q95 (Iρ/αΒτ)) of ~41 (MA m-1 Τ-1) simultaneous with a record plasma elongation of κ ≡ β /α ~ 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as wellmore » as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance.« less
  • Research on the spherical torus (or spherical tokamak) (ST) 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 ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized 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 US, an ST-based Component Test Facility (CTF) and, ultimatelymore » 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 ({beta}), non-inductive 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 {beta}{sub T} of up to 35% with a near unity central {beta}{sub T} have been obtained. NSTX will be exploring advanced regimes where {beta}{sub T} up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction ({approx}60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to I{sub p} {approx} 500 kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.« less