Overview of the physics and engineering design of NSTX upgrade

J. Menard, J. Caniky, J. Chrzanowski, M. Denault, L. Dudek, S. Gerhardt, S. Kaye, C. Kessel, E. Kolemen, R. Maingi, C. Neumeyer, M. Ono, E. Perry, R. Raman, S. Sabbagh, M. Smith, V. Soukhanovskii, T. Stevenson, R. Strykowsky, P. TitusK. Tresemer, M. Viola, M. Williams

Research output: Chapter in Book/Report/Conference proceedingConference contribution

7 Scopus citations

Abstract

The spherical tokamak (ST) is a leading candidate for a fusion nuclear science facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the U.S. actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3-6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1-1.5s to 5s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up (from IP = 0.4MA up to 0.8-1MA) as needed for an ST-based FNSF. In boundary physics, NSTX and higher-A tokamaks measure an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavorably impact next-step devices. Recently, NSTX has successfully demonstrated very high flux expansion and substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade are described.

Original languageEnglish (US)
Title of host publication2011 IEEE/NPSS 24th Symposium on Fusion Engineering, SOFE 2011
DOIs
StatePublished - Nov 16 2011
Event2011 IEEE/NPSS 24th Symposium on Fusion Engineering, SOFE 2011 - Chicago, IL, United States
Duration: Jun 26 2011Jun 30 2011

Publication series

NameProceedings - Symposium on Fusion Engineering

Other

Other2011 IEEE/NPSS 24th Symposium on Fusion Engineering, SOFE 2011
CountryUnited States
CityChicago, IL
Period6/26/116/30/11

All Science Journal Classification (ASJC) codes

  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering

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  • Cite this

    Menard, J., Caniky, J., Chrzanowski, J., Denault, M., Dudek, L., Gerhardt, S., Kaye, S., Kessel, C., Kolemen, E., Maingi, R., Neumeyer, C., Ono, M., Perry, E., Raman, R., Sabbagh, S., Smith, M., Soukhanovskii, V., Stevenson, T., Strykowsky, R., ... Williams, M. (2011). Overview of the physics and engineering design of NSTX upgrade. In 2011 IEEE/NPSS 24th Symposium on Fusion Engineering, SOFE 2011 [6052355] (Proceedings - Symposium on Fusion Engineering). https://doi.org/10.1109/SOFE.2011.6052355