Physics Design of the National High-power Advanced Torus Experiment
Authors: J.E. Menard, R.J. Goldston, G.-Y. Fu, N. Gorelenkov, S.M. Kaye, G. Kramer, R. Maingi, C.L. Neumeyer, S.A. Sabbagh, V.A. Soukhanovskii, and R. Woolley
Abstract:
Moving beyond ITER toward a demonstration power reactor (Demo) will require the integration of stable high fusion gain in steady-state, advanced methods for dissipating very high divertor heat-fluxes, and adherence to strict limits on in-vessel tritium retention. While ITER will clearly address the issue of high fusion gain, and new and planned long-pulse experiments (EAST, JT60-SA, KSTAR, SST-1) will collectively address stable steady-state highperformance operation, none of these devices will adequately address the integrated heat-flux, tritium retention, and plasma performance requirements needed for extrapolation to Demo. Expressing power exhaust requirements in terms of Pheat/R, future ARIES reactors are projected to operate with 60-200MW/m, a Component Test Facility (CTF) or Fusion Development Facility (FDF) for nuclear component testing (NCT) with 40-50MW/m, and ITER 20-25MW/m. However, new and planned long-pulse experiments are currently projected to operate at values of Pheat/R no more than 16MW/m. Furthermore, none of the existing or planned experiments are capable of operating with very high temperature first-wall (Twall = 600-1000C) which may be critical for understanding and ultimately minimizing tritium retention with a reactor-relevant metallic first-wall. The considerable gap between present and near-term experiments and the performance needed for NCT and Demo motivates the development of the concept for a new experiment — the National High-power advanced-Torus eXperiment (NHTX) — whose mission is to study the integration of a fusion-relevant plasma-material interface with stable steady-state high-performance plasma operation.
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Presented at the Thirty-fourth EPS Conference on Plasma Physics, 2–6 July, 2007, Warsaw, Poland.
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Download PPPL-4252 July 2007 (pdf 660 kb)
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