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Meeting Report MEETING REPORT SUMMARY OF WORKSHOP ON 3. potential for stronger vertical stability and fewer or ESTABLISHING THE PHYSICS BASIS no disruptions, easing design of first wall, blanket, NEEDED TO ACCESS THE POTENTIAL and shield structure OF COMPACT TOROIDAL REACTORS, 4. possible cost advantages due to smaller core size and OAK RIDGE, TENNESSEE, lower magnetic fields JULY 19-22, 1994 5. physics that permits, if desirable, higher wall load- ing in smaller unit size INTRODUCTION 6. possibly the only configuration that permits viable advanced fuel reactors. A workshop on establishing the physics basis needed to access the potential of compact toroidal reactors was held Major issues for this configuration are July 19-21, 1994, at Oak Ridge National Laboratory, Oak 1. unshielded single-turn normal-conducting center leg Ridge, Tennessee. The presentations and discussions at the for the toroidal field coil (TFC), which requires reg- workshop covered the low-aspect-ratio (A) tokamak and ular replacement (at fluences <10 MW-yr/m2) and the spheromak. The findings of the workshop, summarized generates added radioactive waste here, covered 2. divertor design and power handling, which are yet to 1. reactor and pathway be developed for any A and may be more challeng- 2. plasma confinement ing for low A 3. magnetohydrodynamic (MHD) equilibrium and 3. recirculating power associated with normal-conduct- global stability ing center leg and current drive, which must be minimized. 4. MHD beta-limit studies Systems code calculations for deuterium-tritium and 5. plasma heating and current drive advanced-fuel low-^4 reactors using the presented physics as- 6. divertor power and particle handling sumptions were recommended by the workshop. The \ow-A tokamak offers the possibility of a steady- 7. present device experimental plans state high-volume plasma neutron source (HVPNS) that 2 8. action items. produces neutron wall loading of 1 to 2 MW/m in small, driven, high-current plasmas (R0 « 0.8 to 1.0 m, Ip « 6 to The workshop benefited from advice from R. A. Blanken 9 MA). Such a device provides a reduced-cost approach for and S. A. Eckstrand of the Office of Fusion Energy of the testing full-function power blankets if high neutron fluence U.S. Department of Energy (DOE). («6 MW-yr/m2) can be obtained. It would complement the capability of the International Thermonuclear Experi- REACTOR AND PATHWAY mental Reactor (ITER) in testing full-size blankets to a lim- ited fluence (~1 MW-yr/m2). Reliable power blankets are Tokamak needed to achieve high duty factors in a fusion demonstra- tion power plant. The low-^4 tokamak provides the following reactor ad- vantages relative to the standard and high-^4 tokamaks: Spheromak 1. better access for remote maintenance of blanket, shield, and divertor The potential advantages for the spheromak are 2. simpler system requiring no inboard blanket, no 1. no need for TFCs, leading to a simplified torus ohmic, and possibly no poloidal divertor coils geometry 210 FUSION TECHNOLOGY VOL. 29 MAR. 1996 210 2. helicity injection current drive, which is projected to Spheromak be efficient in reactor level devices, assuming devel- For the spheromak, the most important need is to de- opment of steady-state helicity injection with small termine core energy confinement in a steadily sustained, energy losses due to the induced magnetic turbulence stable discharge. Spheromak experimental physics made sig- 3. the possibility of combining the divertor, helicity in- nificant progress during the decade in which the United jector for current drive, and fueling system and po- States conducted the research; an electron temperature of sitioning them near the machine axis (small R) for 400 eV was reached in the final tests on the Compact Torus vertical access. Experiment (CTX) (Los Alamos National Laboratory). A strong relationship was recently observed in the The issues facing the spheromak are reversed-field pinch (RFP) experiment of the Madison Sym- 1. lack of data on core energy confinement and high metric Torus (University of Wisconsin) between magnetic beta limit during sustainment fluctuations and plasma heat and particle fluxes Qr and Tr near the plasma edge. 2. unresolved mechanisms and scaling for confinement Magnetic fluctuations and radial transport losses are ob- and beta limits served to decrease as the profile of X = j)\(r)/B(r) is flat- 3. feedback or other stabilization required for global in- tened in spheromaks and in RFPs. As the helicity lifetime stabilities for discharges lasting beyond the flux con- increases with electron temperature, this observation leads server timescales. to the prediction that energy losses will decrease as the spher- omak core becomes hotter and, in some simple modeling, The normal-conductor TFC center leg introduces an that the confinement would be adequate when extrapolated added complexity to the low-^1 tokamak relative to the to a reactor. A model of spheromak core confinement based spheromak. The external toroidal field, however, produces on this mechanism has been developed and indicates favor- the high-<7 tokamak magnetic field geometry with the at- able extrapolation to a reactor. tending promises of tokamak-like confinement, high beta The timescale for plasma buildup in this experiment can stability, current drive, and impurity control. The sphero- be shorter than the soak-through time of the flux conserver, mak progress in these areas has been relatively modest in so that no externally applied vertical field is required to comparison. maintain plasma edge position. No auxiliary heating power Paux would be required for this purpose. PLASMA CONFINEMENT Wall conditioning will be important as in tokamaks; progress in the CTX experiments will be obtained only when Tokamak discharge cleaning and wall conditioning are done. The Small Tight Aspect Ratio Tokamak (START) (Cul- ham Laboratory, United Kingdom) experiment and initial MHD EQUILIBRIUM AND GLOBAL STABILITY kinetic instability analysis using modeled equilibria and pro- files have provided enticing results on plasma confinement Tokamak in the low-y4 tokamak. The following properties characterize the low-,4 Confinement results from a mega-ampere, low-/! toka- tokamak: mak are needed to develop a more broadly based empirical confinement scaling. Only very limited data are available 1. Given a simple vertical field (decay index >0), the from START for comparison with scalings; additional in- limiter plasma elongation (K = b/a) increases to >2 and tri- formation is hoped for from the Current Drive Experiment- angularity to >0.6 naturally without using plasma-shaping Upgrade (CDX-U) [Princeton Plasma Physics Laboratory coils as A approaches the ultralow A range of 1.1 to 1.2. (PPPL)], the Helicity Injected Tokamak (HIT) (University 2. This plasma is vertically stable without external of Washington), and START with neutral beam injection control. (NBI) heating. Auxiliary heating with Paux » Poh will be needed to develop scalings with confidence. 3. Strong elongation, triangularity, and toroidicity lead The experiment will investigate the suppression of to strong plasma shaping, producing (Ipq^/aBt0), which trapped particle modes and related turbulence and transport. approaches 200 MA/m • T at ultralow A. START has pro- The initial analysis indicated the following: duced (Ipq^/aBt0) « 20 MA/m-T for A « 1.4 and K « 1.6. 1. Increased good curvature at low A reduces the drive 4. The START plasma has not experienced major dis- for this instability. ruptions for A < 1.8, whereas CDX-U plasmas did recently. The causes for the difference need to be investigated. 2. Strong dependence on A is expected in the range of 5. The scrape-off layer (SOL) flux tube connected to the 1.2 to 1.6 and on v*e at low collisionality. plasma inboard diminishes as A decreases, leaving the SOL 3. Fluctuation diagnostics in such plasmas will be nec- mostly diverted without using divertor coils. essary to evaluate these effects directly. 6. The PPPL calculations and Tokyo Spheromak-3 The experiment will permit study of H-mode physics at (TS-3) (University of Tokyo, Japan) experimental results low A. The present ohmic plasma data provide no informa- suggest that plasma tilt can be prevented for the low-,4 tion on the L or H mode. Effects of divertor configuration, tokamak when Ip/Itf ^ 2 and/or q^ > 1. The dependence wall conditioning, and threshold powers on L- and H-mode of this limit on plasma shaping and profiles needs to be transitions need to be measured. clarified. FUSION TECHNOLOGY VOL. 29 MAR. 1996 211 Spheromak PLASMA HEATING AND CURRENT DRIVE Major spheromak plasma stability properties are (a) Tokamak tilt, shift, and \ow-n external kinks that can be stabilized by a close-fitting flux-conserver over a resistive wall time and In low-,4 tokamaks, NBI, ion cyclotron resonance fre- (b) feedback stabilization of these modes on longer time- quency wave, ion Bernstein wave, Alfven wave, electron cy- scales. Rotation may provide stability against some modes. clotron wave, and lower-hybrid wave are being considered for heating, and NBI and helicity injection are being con- sidered for current drive. MHD BETA LIMIT STUDIES 1. Many radio-frequency schemes are expected to Tokamak present special challenges for high beta studies at low A be- cause of the low field (<1 T) and high density expected. (Ini- Tremendous progress was reported at this workshop in tial calculations of electron Landau damping, performed quantitative theoretical analysis of tokamak beta limits at after the workshop, look very favorable for fast-wave elec- low A. tron heating at the frequency range of the high ion cyclo- 1. For current profiles with q0 ~ 1 and A down to 1.3, tron harmonics.) the Troyon scaling remains roughly unchanged (@Nt rises slightly in some analyses and falls in others), giving </3,> up 2. For the next-step mega-ampere experiments, NBI is to -20% without a conducting wall.
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