The Spherical Tokamak Components Test Facility Rapporteured by Howard Wilson representing Plasma Science and Fusion Engineering Conditions of Spherical.

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Presentation transcript:

The Spherical Tokamak Components Test Facility Rapporteured by Howard Wilson representing Plasma Science and Fusion Engineering Conditions of Spherical Torus Component Test Facility FT/3-1Rb, Y.-K.M. Peng, et al A Steady State Spherical Tokamak for Components Testing FT/3-1Ra, H.R. Wilson, et al 20th IAEA Fusion Energy Conference, Vilamoura, 2004

Plasma Science and Fusion Engineering Conditions of Spherical Torus Component Test Facility Y.-K. M. Peng, C.C. NeuMeyer, P.J. Fogarty, C. Kessel, D.J. Strickler, P. Rutherford, D. Mikkelsen, T.W. Burgess, R. Bell, J. Menard, D. Gates, S. Sabbagh, O. Mitarai, J. Schmidt, E, Synakowski, J. Tsai, L. Grisham, B.E. Nelson, E.T. Cheng, L. El-Guebaly Oak Ridge National Laboratory Princeton Plasma Physics Laboratory Columbia University, New York A Steady State Spherical Tokamak for Components Testing H.R. Wilson, G.M. Voss, R.J. Akers, L. Appel, I. Chapman, J.P. Christiansen, G. Cunningham, A. Dnestrovskij, O. Keating, M.J. Hole, G. Huysmans, A. Kirk, P.J. Knight, M. Loughlin, K.G. McClements, A.W. Morris, M.R. O’Brien, D.Yu. Sychugov, M. Valovic UKAEA Culham Science Centre I V Kurchatov Institute Imperial College, London Kyushu Tokai University, Japan TSI Research, Solano Beach, CA USA University of Wisconsin University of Sydney CEA Cadarache Moscow State University

Outline Motivation Why a spherical tokamak Design features Plasma physics studies Conclusions

Motivation The present strategy for addressing plasma physics, fusion technology and materials is through ITER, IFMIF and DEMO: ITER IFMIF DEMO Plasma physics and technology Testing small material samples Full testing of components and structures A dedicated, small scale Components Testing Facility would provide support for DEMO: – Greater flexibility – More rapid blanket-testing capability CTF

The Goals of a CTF The conditions necessary for testing were identified by an international committee [1]: – A fusion neutron wall loading in the range 1-2MWm  2 – Steady state operation – A total neutron fluence of ~6MW-yrm  2 within ~12 years – Total test area exceeding 10m 2 – Magnetic field strength exceeding 2T [1] Abdou, et al, Fusion Technology 29 (1996)1 We explore the possibility that a driven burning plasma spherical tokamak can meet these requirements

The ORNL/PPPL study explores a larger device: – Benefit: able to test tritium generation towards self-sufficiency, including material composition of chamber systems – Cost: restricted range of suitable materials ORNL/PPPL and Culham strategies are complementary The Culham strategy is for a compact device: – Benefit: does not need to generate tritium – Cost: access to optimised regimes limited by tritium availability The size of the device has implications for tritium management: Culham and ORNL/PPPL adopt complementary approaches

Main advantages of an ST as CTF Tight aspect ratio means few neutrons absorbed on inner wall High volume-to-surface area ratio: high fusion power density Easy to maintain: centre column and divertor coils drop out through bottom Culham design (peaked current profile) All coils are normal- conducting copper

High availability requirement drives the design ORNL/PPPL design Hands-on connect and disconnect service lines outside of shielding and vacuum boundaries Divertor, cylindrical blanket, TF center leg, and shield assembly removed/installed vertically Machine Assembly/Disassembly Sequence Are Made Manageable Remove top hatch Remove PF coils and divertors Extract NBI liner, test modules and blanket assemblies Remove centerstack assembly Remove shield assembly

Parameters of the designs Key drivers for both designs: – High TF (~2.5T) and elongation to allow high plasma current at kink limit – Low density for efficient current drive – High beta and good confinement

Based on modest assumptions regarding MHD and confinement Culham design exploits high normalised current Range of ORNL/PPPL designs sits within NSTX data-set MAST Toroidal rotation ORNL/ PPPL Culham Assumed confinement enhancement factors are consistent with MAST high rotation plasmas Consistent with NSTX data, separating electron and ion confinement NSTX

Current Drive Neutral beam injection provides the main CD for both designs ORNL/PPPL design exploring EBW for off-axis CD (~10MW, 140GHz) Culham design exploring ECCD for on-axis CD (20MW, 160GHz) Culham 150keV NBI system Based on LOCUST calculations Off-axis CD required for stability Could provide all fuelling ORNL/PPPL 110keV NBI system (from TSC, PEST2) appropriate for MHD-stable profiles ( l i = )

Layout of ORNL/PPPL design, showing NBI injectors and scheme for removing test modules Top view

Handling the exhaust Two approaches: ORNL/PPPL exploits “inboard-limited” configuration to spread heat load DND also possible Culham adopts DND configuration: Up to ~95% heat to outer target (MAST) High heat loads require novel scheme Exploring “pebble-divertor” (outboard) – reduces heat loads to ~10MWm -2 Resulting heat loads are  15MWm  2 Both designs rely on radiated power

Neutronics study Meets Abdou et al requirements: Equatorial test modules: 1.63MWm  2 Polar test modules: 1.40 MWm  2 6MW-yrm  2 achieved in ~12 yrs with ~30% availability Tritium consumption is ~0.9kgyr  1, so not reliant on ability to generate T Tolerable radiation damage, except for divertor coils (need improved design, using cyanate ester resin) Study of Culham design using MCNP code shows:

Conclusions ORNL/PPPL and Culham have independently developed designs for CTF based on a spherical tokamak – The designs are complementary, with some similarities and some differences – It is encouraging that a range of solutions exist – Modest assumptions for the plasma performance have been made for the “baseline” regimes A CTF based on an ST looks feasible, but requires further research in a number of key areas: – Exhaust and divertor design – Influence of high momentum injection/fast particle content – Off-axis current drive – Start-up – First wall material

Please visit posters FT/3-1Ra and FT/3-1Rb for much more detail Acknowledgements: Work funded by: United Kingdom Engineering and Physical Sciences Research Council; EURATOM; program development of ORNL UT-Battelle, and US DoE Contract Nos DE-AC02-76CH03073 and DE-AC05-96OR22464.