The advanced tokamak path to a compact net electric fusion pilot plant

Physics-based simulations project a compact net electric fusion pilot plant with a nuclear testing mission is possible at modest scale based on the advanced tokamak concept, and identify key parameters for its optimization. These utilize a new integrated 1.5D core-edge approach for whole device modeling to predict performance by self-consistently applying transport, pedestal and current drive models to converge fully non-inductive stationary solutions, predicting profiles and energy confinement for a given density. This physics-based approach leads to new insights and understanding of reactor optimization. In particular, the levering role of high plasma density is identified, which raises fusion performance and self-driven ‘bootstrap currents’, to reduce current drive demands and enable high pressure with net electricity at a compact scale. Solutions at 6–7 T, ∼4 m radius and 200 MW net electricity are identified with margins and trade-offs possible between parameters. Current drive comes from neutral beam and ultra-high harmonic (helicon) fast wave, though other advanced approaches are not ruled out. The resulting low recirculating power in a double null configuration leads to a divertor heat flux challenge that is comparable to ITER, though reactor solutions may require more dissipation. Strong H-mode access (x2 margin over L–H transition scalings) and ITER-like heat fluxes are maintained with ∼20%–60% core radiation, though effects on confinement need further analysis. Neutron wall loadings appear tolerable. The approach would benefit from high temperature superconductors, as higher fields would increase performance margins while potential for demountability may facilitate nuclear testing. However, solutions are possible with conventional superconductors. An advanced load sharing and reactive bucking approach in the device centerpost region provides improved mechanical stress handling. The prospect of an affordable test device which could close the loop on net-electric production and conduct essential nuclear materials and breeding research is compelling, motivating research to validate the techniques and models employed here.

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

[2]  H. Zohm,et al.  On the Minimum Size of DEMO , 2010 .

[3]  E. Doyle,et al.  Comparison of turbulence measurements from DIII-D low-mode and high-performance plasmas to turbulence simulations and models , 2001 .

[4]  J. Manickam,et al.  Advanced tokamak physics-status and prospects , 1994 .

[5]  L. L. Lao,et al.  Advanced tokamak research in DIII-D , 2004 .

[6]  S. Jardin,et al.  Plasma Profile and Shape Optimization for the Advanced Tokamak Power Plant, ARIES-AT , 2001 .

[7]  T. Petrie,et al.  High fusion performance in Super H-mode experiments on Alcator C-Mod and DIII-D , 2019, Nuclear Fusion.

[8]  J. Contributors,et al.  Scaling of the tokamak near the scrape-off layer H-mode power width and implications for ITER , 2013 .

[9]  T. C. Luce,et al.  Development of Steady-State Advanced Tokamak Research in the DIII-D Tokamak , 2005 .

[10]  Jun Ho Yeom,et al.  Design concept of K-DEMO for near-term implementation , 2015 .

[11]  M Kikuchi,et al.  Prospects of a stationary tokamak reactor , 1993 .

[12]  J. A. Leuer,et al.  Fusion Nuclear Science Facility Candidates , 2011 .

[13]  A. M. Garofalo,et al.  DIII-D Research to Prepare for Steady State Advanced Tokamak Power Plants , 2018, Journal of Fusion Energy.

[14]  J. Kinsey,et al.  Non-dimensional scaling of turbulence characteristics and turbulent diffusivity , 2001 .

[15]  T. C. Luce,et al.  Realizing Steady State Tokamak Operation for Fusion Energy , 2009 .

[16]  Y. Sakamoto,et al.  Technological assessment between vertical and horizontal remote maintenance schemes for DEMO reactor , 2017 .

[17]  T. Osborne,et al.  Access to a new plasma edge state with high density and pressures using the quiescent H mode. , 2014, Physical review letters.

[18]  L. Lao,et al.  Reconstruction of current profile parameters and plasma shapes in tokamaks , 1985 .

[19]  L. L. Lao,et al.  Beta limits in long-pulse tokamak discharges , 1997 .

[20]  John D Galambos,et al.  Commercial tokamak reactor potential with advanced tokamak operation , 1995 .

[21]  Nathaniel J. Fisch,et al.  Theory of current-drive in plasmas , 1987 .

[22]  G. Staebler,et al.  Progress in extending high poloidal beta scenarios on DIII-D towards a steady-state fusion reactor and impact of energetic particles , 2020, Nuclear Fusion.

[23]  M. Gadomska,et al.  Overview of EU DEMO design and R&D activities , 2014 .

[24]  C. C. Petty,et al.  Sizing up plasmas using dimensionless parametersa) , 2006 .

[25]  C. E. Kessel,et al.  Physics Basis for a Conservative Physics and Conservative Technology Tokamak Power Plant: ARIES-ACT2 , 2015 .

[26]  A. Glasser.  The direct criterion of Newcomb for the ideal MHD stability of an axisymmetric toroidal plasma , 2016 .

[27]  J. Kinsey,et al.  A theory-based transport model with comprehensive physicsa) , 2006 .

[28]  G. Bateman,et al.  The tokamak Monte Carlo fast ion module NUBEAM in the National Transport Code Collaboration library , 2004 .

[29]  J. Schweinzer,et al.  Impurity seeding for tokamak power exhaust: from present devices via ITER to DEMO , 2013 .

[30]  Minami Yoda,et al.  The ARIES Advanced and Conservative Tokamak Power Plant Study , 2014 .

[31]  H. R. Wilson,et al.  A first-principles predictive model of the pedestal height and width: development, testing and ITER optimization with the EPED model , 2011 .

[32]  T. Petrie,et al.  Advances in the steady-state hybrid regime in DIII-D—a fully non-inductive, ELM-suppressed scenario for ITER , 2017 .

[33]  T. Luce,et al.  Stability of DIII-D high-performance, negative central shear discharges , 2017 .

[34]  T S Taylor,et al.  Sustained stabilization of the resistive-wall mode by plasma rotation in the DIII-D tokamak. , 2001, Physical review letters.

[35]  Laila A. El-Guebaly,et al.  Fusion nuclear science facilities and pilot plants based on the spherical tokamak , 2016 .

[36]  L. C. Bernard,et al.  GATO: An MHD stability code for axisymmetric plasmas with internal separatrices , 1981 .

[37]  Nermin A. Uckan,et al.  Chapter 9: ITER contributions for Demo plasma development , 2007 .

[38]  Robert L. Miller,et al.  Synergism between cross-section and profile shaping in beta optimization of tokamak equilibria with negative central shear , 1998 .

[39]  H. Zohm,et al.  Where to locate DEMO in a one-step-to-an-FPP strategy , 2016 .

[40]  L. L. Lao,et al.  The ARIES-AT advanced tokamak, Advanced technology fusion power plant , 2006 .

[41]  D. H. Na,et al.  Development of a systematic, self-consistent algorithm for the K-DEMO steady-state operation scenario , 2017 .

[42]  J. Bialek,et al.  H2 optimal control techniques for resistive wall mode feedback in tokamaks , 2018 .

[43]  E. Surrey,et al.  Optimal design of a toroidal field magnet system and cost of electricity implications for a tokamak using high temperature superconductors , 2015 .

[44]  P. Bonoli,et al.  ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets , 2014, 1409.3540.

[45]  H. St. John,et al.  An efficient transport solver for tokamak plasmas , 2017, Comput. Phys. Commun..

[46]  O. Meneghini,et al.  Integrated modeling of high βN steady state scenario on DIII-D , 2018 .

[47]  L. L. Lao,et al.  Optimized profiles for improved confinement and stability in the DIII-D tokamak , 1994 .

[48]  M. Greenwald.  Density limits in toroidal plasmas , 2002 .

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

[50]  T. C. Luce,et al.  Application of dimensionless parameter scaling techniques to the design and interpretation of magnetic fusion experiments , 2008 .

[51]  K. Matsuda,et al.  Ray tracing study of the electron cyclotron current drive in DIII-D using 60 GHz , 1989 .