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B.E. Nelson,
R.D. Benson,
L.A. Berry,
A.B. Brooks,
M.J. Cole,
P.J. Fogrty,
P.L. Goranson,
P. Heitzenroeder,
S.P. Hirschman, G.H. Jones,
J.F. Lyon,
P.K. Mioduszewski,
D.A. Monticello,
D.A. Spong,
D.J. Strickler,
A.S. Ware,
D.E. Williamson
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ABSTRACT: The engineering design status of the Quasi-Poloidal Stellarator Experiment (QPS) is presented. The purpose, configuration, and possible manufacturing and assembly techniques of the various components of the core are described.
Fusion Engineering, 2002. 19th Symposium on; 02/2002
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ABSTRACT: The design of the lower hybrid launcher is configured to couple
1.5 MW of steady state lower hybrid heating/lower hybrid current drive
(LHH/LHCD) power at 3.7 GHz to the Tokamak Physics Experiment (TPX)
plasma. The launcher utilizes a unique laminated, water-cooled
construction to achieve a compact array of 128 waveguides at the grille
surface and includes real-time position control for plasma coupling. The
design features of this launcher are described with details of the
principal components and the possible upgrade of the launcher to 3 MW.
Results of steady state thermal analysis of the launcher grille septa
are included
Fusion Engineering, 1993., 15th IEEE/NPSS Symposium on; 11/1993
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ABSTRACT: The tokamak radiation shielding includes the neutron and gamma
shielding around the torus and penetrations required to (1) limit
activation of components outside the shield to levels that permit
hands-on maintenance and (2) limit the nuclear heating of the
superconducting coils and cold structure. The primary design drivers are
space, the 350°C bakeout temperature, and cost; therefore, different
shield materials were used for different shield components and
locations. The shielding is divided into three areas: (1) torus
shielding around the vacuum vessel, (2) duct shielding around the vacuum
pumping ducts and vertical diagnostic ducts, and (3) penetration
shielding in and around the radial ports. The major shield components
include water between the walls of the vacuum vessel, lead
monosilicate/boron carbide tiles that are attached to the exterior of
the vacuum vessel, shield plugs that fill the openings of the large
radial ports, and polyethylene/lead/boron shield blocks for duct
shielding. A description of the shielding configuration and the
performance and operational requirements will be discussed
Fusion Engineering, 1993., 15th IEEE/NPSS Symposium on; 11/1993
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ABSTRACT: The conceptual design of the TPX vacuum pumping system is presented. The baseline concept includes a high vacuum pumping system, a roughing and backing system, volume pumping ducts, a leak detection system, a diagnostic pumping system, and a cryostat pumping system. The high vacuum pumping system will initially evacuate the torus, provide pumping of the diverters during operation, and provide pumping for glow discharge cleaning. The high vacuum pumping system has high throughput and variable conductance capabilities and includes cryocondensation pumps for pumping deuterium during normal operation as well as turbomolecular pumps for pumping helium and for glow discharge cleaning. Sixteen vacuum ducts extend from the vacuum vessel through the cryostat to the pumping system; each duct contains a torus isolation valve and an electrical break. Butterfly valves at the cryopump inlets will be used for throttling the pumps and for pump regeneration. In this way, half of the pumps can be regenerated while the others are operating. The specific design parameters and predicted performance of the vacuum pumping system are discussed, as are the upgrade options for steady state and DT operation
Fusion Engineering, 1993., 15th IEEE/NPSS Symposium on; 11/1993
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ABSTRACT: A proposal has been made to improve the performance of ITER
(International Thermonuclear Experimental Reactor) by modifying the
reference design point. The high-aspect-ratio design (HARD) proposal
increases the aspect ratio to four, increases the central toroidal field
to 7 T, and decreases the peak plasma current to 14.8 MA. Systems
studies indicate that the HARD parameters provide improved technology
testing capability by increasing the neutron wall loading while
maintaining the CDA (conceptual design activity) physics constraints.
Changes in configuration, loading, stresses, and other factors for the
containment structure components, based on the HARD parameters, have
been investigated. In general, the HARD configuration has only minor
effects on the containment structures. Advantages include better
distribution of resistance and lower loads and stresses for the vacuum
vessel. Disadvantages include much higher electromagnetic loads on the
inboard blankets, less margin in the stability parameter, and the
requirement to put passive loops on the inboard blankets. No significant
change in cost would be expected
Fusion Engineering, 1991. Proceedings., 14th IEEE/NPSS Symposium on; 11/1991
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B.E. Nelson,
L.A. Berry,
A.B. Brooks,
M.J. Cole,
J.C. Chrzanowski,
H.-M. Fan,
P.J. Fogarty,
P.L. Goranson,
P.J. Heitzenroeder,
S.P. Hirshman, G.H. Jones,
J.F. Lyon,
G.H. Neilson,
W.T. Reiersen,
D.J. Strickler,
D.E. Williamson
[show abstract]
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ABSTRACT: The National Compact Stellarator Experiment (NCSX) [http://www.pppl.gov/ncsx/Meetings/CDR/CDRFinal/EngineeringOverview_R2.pdf] is being designed as a proof of principal test of a quasi-axisymmetric compact stellarator. This concept combines the high beta and good confinement features of an advanced tokamak with the low current, disruption-free characteristics of a stellarator. NCSX has a three-field-period plasma configuration with an average major radius of 1.4 m, an average minor radius of 0.33 m and a toroidal magnetic field on axis of up to 2 T. The stellarator core is a complex assembly of four coil systems that surround the highly shaped plasma and vacuum vessel. Heating is provided by up to four, 1.5 MW neutral beam injectors and provision is made to add 6 MW of ICRH. The experiment will be built at the Princeton Plasma Physics Laboratory, with first plasma expected in 2007.
Fusion Engineering and Design.
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ABSTRACT: Several designs that address plasma-facing plate configurations
and thermal-hydraulic design issues have been developed for the ITER
(International Thermonuclear Experimental Reactor) divertor. Design
criteria growing out of physics requirements, physical constraints, and
remote handling requirements impose severe mechanical requirements on
the support structure and its attachments. The authors discuss the
design criteria for the divertor mounting structure and identify the
mechanical design issues that need to be addressed. Achieving the
criteria may require the development of new components and innovative
configurations, specifically a new class of remote fasteners and
electrically resistant material for mounts. The possible design of such
components and an R&D program to develop them are described, and
issues specific to the high-aspect-ratio design (HARD) configuration are
summarized. Analysis and experiments that will resolve these issues and
concerns and lead to a final ITER design are identified
Fusion Engineering, 1991. Proceedings., 14th IEEE/NPSS Symposium on;
