Power and particle exhaust in an ST-FNSF

A spherical tokamak (ST) configuration is favorable for a Fusion Nuclear Science Facility (FNSF), due to its small size and relatively low cost. However, the compactness of the ST also exacerbates the power and particle handling problems anticipated in next-step devices, since local fluxes are higher and less space is available for optimizing plasma-facing components. Here we present an analysis of the power and particle handling requirements of a candidate ST-FNSF, based on 0-D exhaust projections as well as 2-D edge plasma modeling using the SOLPS code. These show that, for reasonable assumptions on cross-field transport, operating points can be identified that are consistent with both core plasma operation and power and particle exhaust requirements.

[1]  R. Maingi,et al.  Neutral pressure dynamics in the upper plenums in the DIII-D tokamak , 2004 .

[2]  Robert J. Goldston,et al.  Heuristic drift-based model of the power scrape-off width in low-gas-puff H-mode tokamaks , 2011 .

[3]  D. Ryutov.  Geometrical Properties of a "Snow-Flake" Divertor , 2007 .

[4]  J.M. Park,et al.  Fusion Nuclear Science Facility (FNSF) before Upgrade to Component Test Facility (CTF) , 2010 .

[5]  D. Reiter,et al.  The EIRENE and B2-EIRENE Codes , 2005 .

[6]  Brian Labombard,et al.  Analysis of a multi-machine database on divertor heat fluxesa) , 2011 .

[7]  A. Loarte,et al.  Basic divertor operation in ITER-FEAT , 2002 .

[8]  S. Ide,et al.  Compact DEMO, SlimCS: design progress and issues , 2009 .

[9]  A Continuous Cryogenic Diffusion Pump For Fusion Reactors , 2005, 21st IEEE/NPS Symposium on Fusion Engineering SOFE 05.

[10]  P. Valanju,et al.  Super-X divertors and high power density fusion devices , 2009 .

[11]  T. Eich,et al.  Inter-ELM power decay length for JET and ASDEX upgrade: measurement and comparison with heuristic drift-based model. , 2011, Physical review letters.

[12]  Philippe Ghendrih,et al.  The Plasma Boundary of Magnetic Fusion Devices , 2001 .

[13]  A. Loarte,et al.  Scaling laws for edge plasma parameters in ITER from two-dimensional edge modelling , 2003 .

[14]  Rajesh Maingi,et al.  Pump plenum pressure dependence on divertor plasma parameters and magnetic geometry in the DIII-D tokamak , 1999 .

[15]  V. Soukhanovskii,et al.  Dependence of divertor heat flux widths on heating power, flux expansion, and plasma current in the NSTX , 2011 .

[16]  Nstx Team,et al.  Divertor heat flux mitigation in the National Spherical Torus Experimenta) , 2008 .

[17]  S. Wolfe,et al.  A new look at density limits in tokamaks , 1988 .

[18]  B. Braams,et al.  Plasma Edge Physics with B2‐Eirene , 2006 .

[19]  D. Ryutov.  Geometrical properties of a “snowflake” divertor , 2007 .

[20]  A. Kameari,et al.  Ceramic turbomolecular pumping system in reactor structure of FER , 1989 .

[21]  P. C. Stangeby,et al.  Obtaining reactor-relevant divertor conditions in tokamaks , 2011 .

[22]  Tomonori Takizuka,et al.  Power requirement for accessing the H-mode in ITER , 2008 .

[23]  B. LaBombard.  Scaling of the power exhaust channel in Alcator C-Mod , 2010 .

[24]  B. Braams.  Radiative Divertor Modelling for ITER and TPX , 1996 .