Engineering design of the national spherical tokamak experiment

J. Spitzer; M. Ono; M. Peng; D. Bashore; T. Bigelow; A. Brooks; J. Chrzanowski; H. M. Fan; P. Heitzenroeder; T. Jarboe; R. Kaita; S. Kaye; H. Kugel; R. Majeski; C. Neumeyer; al et al

doi:10.13182/fst96-a11963134

Engineering design of the national spherical tokamak experiment

J. Spitzer
, M. Ono
, M. Peng
, D. Bashore
, T. Bigelow
, A. Brooks
, J. Chrzanowski
, H. M. Fan
, P. Heitzenroeder
, T. Jarboe
, R. Kaita
, S. Kaye
, H. Kugel
, R. Majeski
, C. Neumeyer
, al et al

Research output: Contribution to journal › Conference article › peer-review

33 Scopus citations

Abstract

The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio device with a plasma current of 1 MA. The tokamak features auxiliary heating and current drive with a close-fitting conducting shell to maximize the plasma pressure. NSTX is designed for an experimental pulse length that will demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The design also takes maximum advantage of existing facilities and components from previous Princeton devices to reduce the overall program costs.

Original language	English
Pages (from-to)	1337-1341
Number of pages	5
Journal	Unknown Journal
Volume	30
Issue number	3
DOIs	https://doi.org/10.13182/fst96-a11963134
State	Published - 1996
Externally published	Yes
Event	Proceedings of the 1996 12th Topical Meeting on the Technology of Fusion Energy. Part 2 (of 2) - Reno, NV, USA Duration: Jun 16 1996 → Jun 20 1996

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10.13182/fst96-a11963134

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title = "Engineering design of the national spherical tokamak experiment",

abstract = "The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio device with a plasma current of 1 MA. The tokamak features auxiliary heating and current drive with a close-fitting conducting shell to maximize the plasma pressure. NSTX is designed for an experimental pulse length that will demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The design also takes maximum advantage of existing facilities and components from previous Princeton devices to reduce the overall program costs.",

author = "J. Spitzer and M. Ono and M. Peng and D. Bashore and T. Bigelow and A. Brooks and J. Chrzanowski and Fan, \{H. M.\} and P. Heitzenroeder and T. Jarboe and R. Kaita and S. Kaye and H. Kugel and R. Majeski and C. Neumeyer and \{et al\}, al",

year = "1996",

doi = "10.13182/fst96-a11963134",

language = "English",

volume = "30",

pages = "1337--1341",

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note = "Proceedings of the 1996 12th Topical Meeting on the Technology of Fusion Energy. Part 2 (of 2) ; Conference date: 16-06-1996 Through 20-06-1996",

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TY - JOUR

T1 - Engineering design of the national spherical tokamak experiment

AU - Spitzer, J.

AU - Ono, M.

AU - Peng, M.

AU - Bashore, D.

AU - Bigelow, T.

AU - Brooks, A.

AU - Chrzanowski, J.

AU - Fan, H. M.

AU - Heitzenroeder, P.

AU - Jarboe, T.

AU - Kaita, R.

AU - Kaye, S.

AU - Kugel, H.

AU - Majeski, R.

AU - Neumeyer, C.

AU - et al, al

PY - 1996

Y1 - 1996

N2 - The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio device with a plasma current of 1 MA. The tokamak features auxiliary heating and current drive with a close-fitting conducting shell to maximize the plasma pressure. NSTX is designed for an experimental pulse length that will demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The design also takes maximum advantage of existing facilities and components from previous Princeton devices to reduce the overall program costs.

AB - The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio device with a plasma current of 1 MA. The tokamak features auxiliary heating and current drive with a close-fitting conducting shell to maximize the plasma pressure. NSTX is designed for an experimental pulse length that will demonstrate quasi-steady state non-inductively driven advanced tokamak operation. The design also takes maximum advantage of existing facilities and components from previous Princeton devices to reduce the overall program costs.

UR - https://www.scopus.com/pages/publications/0030074127

U2 - 10.13182/fst96-a11963134

DO - 10.13182/fst96-a11963134

M3 - Conference article

AN - SCOPUS:0030074127

VL - 30

SP - 1337

EP - 1341

JO - Unknown Journal

JF - Unknown Journal

IS - 3

T2 - Proceedings of the 1996 12th Topical Meeting on the Technology of Fusion Energy. Part 2 (of 2)

Y2 - 16 June 1996 through 20 June 1996

ER -

Engineering design of the national spherical tokamak experiment

Abstract

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