High poloidal beta equilibria in the Tokamak Fusion Test Reactor limited by a natural inboard poloidal field null* S. A. Sabbagh,+ R. A. Gross, M. E. Mauel, and G. A. Navratil Department of Applied Physics, Columbia University, New York, New York 10027 M. G. Bell, R. Bell, M. Bitter, N. L. Bretz, R. V. Budny, C. E. Bush, M. S. Chance, P. C. Efthimion, E. D. Fredrickson, Ft. Hatcher, R. J. Hawryluk, S. P. Hirshman,“) A. C. Janos, S. C. Jardin, D. L. Jassby, J. Manickam, D. C. McCune, K. M. McGuire, S. S. Medley, D. Mueller, Y. Nagayama, D. K. Owens, M. Okabayashi, H. K. Park, A. T. Ramsey, B. C. Stratton, E. J. Synakowski, G. Taylor, R. M. Wieland, and M. C. Zarnstortf PIasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 J. Kesner, E. S. Marmar, and J. L. Terry MIT Plasma Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 ( Received 13 December 1990, accepted 1 April 199 1) Recent operation of the Tokamak Fusion Test Reactor (TFIR) [Plasma Phys. Controlled Nucl. Fusion Research 1, 51 ( 1986) ] has produced plasma equilibria with values of A=flpq + Zi/2 as large as 7, ePpdia =2,u,~(P~)/( (B,))’ as large as 1.6, and Troyon normalized diamagnetic beta [Plasma Phys. Controlled Fusion 26, 209 ( 1984); Phys. Lett. llOA, 29 (1985) 1, PNdia G 10”(~,I)aB,/I, as large as 4.7. When ePpdia 2 1.25, a separatrix entered the vacuum chamber, producing a naturally diverted discharge that was sustained for many energy confinement times, rE. The largest values of ED, and plasma stored energy were obtained when the plasma current was ramped down prior to neutral beam injection. The measured peak ion and electron temperatures were as large as 24 and 8.5 keV, respectively. Plasma stored energy in excess of 2.5 MJ and rE greater than 130 msec were obtained. Confinement times of greater than 3 times that expected from L-mode predictions have been achieved. The fusion power gain Q,, reached a value of 1.3 X lo- 3 in a discharge with I, = 1 MA and ePpdia = 0.85. A large, sustained negative loop voltage during the steady-state portion of the discharge indicates that a substantial noninductive component of 1, exists in these plasmas. Transport code analysis indicates that the bootstrap current constitutes up to 65% of rp. Magnetohydrodynamic (MHD) ballooning stability analysis shows that, while these plasmas are near, or at the p,, limit, the pressure gradient in the plasma core is in the first region of stability to high-n modes.
I. INTRODUCTION Tokamak Reactor, SSTR’ designs, have incorporated fu- sion plasma operation at high fi,. However, since experience A significant concern regarding the next generation of tokamak fusion experiments and fusion reactor concepts is from the operation of large tokamaks has generally been at relatively small values of fip (less than the tokamak aspect the large toroidal plasma current I, needed to provide suffi- cient energy confinement to sustain fusion burn conditions ratio), high% reactor designs have been considered specu- lative, regardless of the benefits they may present. The de- in these devices. If improved confinement and adequate sta- monstration of stable, high% plasma operation in present bility can be achieved, operation at high poloidal betap, can alleviate many problems associated with conventional toka- large tokamak devices would begin to provide the experi- mental basis for the design of such a high% reactor. mak reactor designs.‘** Operation at lower I, associated with high @, considerably reduces the adverse consequences This paper presents results from recent experiments in of plasma disruptions. The power requirements for steady- the Tokamak Fusion Test Reactor, TFTR,’ which have pro- duced plasmas with very large values of fl,. While high% state current drive are reduced at lower Ip. In addition, a plasmas have been achieved in other tokamaks,‘-‘* these are significant amount of neoclassical bootstrap current3-5 is predicted to form at high p,, which could further reduce the first experiments that demonstrate stable plasma equili- bria, limited by a natural inboard poloidal field null at high current drive requirements. fl,, that are sustained for many energy confinement times r6 Capitalizing on these advantages, several reactor design in a large neutral beam heated device. The methods em- studies including the ARIES6 and the JAERI Steady-State ployed to produce these plasmas and the plasma parameters *Paper 416, Bull. Am. Phys. Sot. 35, 1992 (1990). achieved are described in the next section. A preliminary ‘Invited speaker.Present address:Princeton Plasma Physics Laboratory, analysis of stability, confinement, noninductive current Princeton University, Princeton, NJ 08543. ” Presentaddress: Oak Ridge National Laboratory, Oak Ridge, Tennessee drive, and neutron production for these high% plasmas is 37830. presented in the subsequent sections.
2277 Phys. Fluids B 3 (8), August 1991 0899-8221 I91 /082277-08$02.00 @ 1991 American Institute of Physics 2277 II. PLASMA PARAMETERS AND OPERATION AT HIGH been fully decreased to the pedestal value. The largest values @P of &,, dia and flNtlia were obtained in discharges in which 1, Stable plasmas with Ppdia~5.9 and efiPdia< 1.6 were was ramped down. Neutral beam heating, utilizing tangen- created in TFTR, surpassing an apparent limiting value of tial injection of deuterium, usually began with coinjection epPdia(0.7 previously observed in TFTR supershots.13*14 only. Subsequently, counter-injected beams were added to provide nearly balanced co- and counter injection at 18-26 Here, Ppdia=2- PO (p,)/((B,))‘, where (PI) is the volume average of the transverse plasma pressure and ((BP)) is the MW during the high+P, steady state. Line-averaged target line average of the poloidal magnetic field taken over the densities were in the range of ( 1-3) x 1019m - 3and typically outer flux surface. The inverse-aspect-ratio is EXZ/&,, with increased to (3-6) x lOI m - 3 during the auxiliary heating the minor radius a defined as one-half of the midplane width phase. and R, defined as the area-averaged major radius. The Waveforms illustrating the 1, ramp-down and separa- Troyon15 normalized diamagnetic beta, fiNdia E 108(,!3,1) trix formation are shown in Fig. 2. An Ohmic plasma was XaB,/I,, reached 4.7. For comparison, TFTR supershots initially formed with IP = 0.85 MA [Fig. 2(a) 1. The plasma had previously reached values of flNdia ~2.7.‘~ Here, B. is current was ramped down and held approximately constant the vacuum toroidal field at R,. Figure 1, which shows at 0.4 MA during the neutral beam heating phase. In this ~/3~dia as a function of q* = 5 ( a2Bo/R,Ip ) ( 1 + $)/2, illus- particular discharge, 1, decreased at a small rate ( - 0.05 trates how the high% plasmas extend the range of super- MA/set) after the fast 1, ramp down had been completed. shot operation in TFTR. Here, K is the plasma elongation. This was caused by the finite ability of the plasma control As eoPdiais increased to a value of approximately 1.25, the system to hold rP at an exactly programmed value while plasma that is initially limited on the inside wall makes a other plasma quantities, particularly & , were changing rap- transition to a diverted plasma, the boundary being defined idly. The voltage associated with this slow decrease in lP is by a naturally occuring separatrix with a poloidal field null approximately - 0.25 V, which is a small fraction of the - on the inboard side. At the larger values of q*, the value of 1.5 V surface voltage measured during the high% phase E&+~ is limited by the separatrix moving further into the of the discharge. plasma minor radius. At the smaller values of q*, the plas- Figure 2(b) shows the evolution of Ppdia and mas encounter a beta limit, but with a value of PN much A+,,, f 1,/2, with ii being the plasma internal induc- larger than the limit of 3.0 originally proposed by Troyon.15 tance. At maximum stored energy, this discharge attained These plasmas were created in deuterium at relatively values of EfiPdia = 1.43 andpNdi, = 3.4 with q* = 8.5. As a small Ip (0.28-l .O MA during neutral beam heating), nomi- result of the tangential beam injection, these discharges are nal B. = 4.8 T, and R, = 2.45 m. Separatrix-limited dis- anisotropic and fipeq E (BP,, +PP, )/2 1s larger than charges were created with two types of lP time histories, one in which Ip was held constant and another in which rP was decreased, or ramped down, before the start of neutral beam t..<-.‘.-.s‘. TFra Sbor4.54652mxl injection. In the latter case, an Ohmic plasma was formed with I,, in the range of 0.85-1.75 MA and held for about 1 sec. The plasma current was then decreased rapidly at - 2.5 MA/set to a pedestal value in the range of 0.3-l .O MA. The maximum ratio of initial to pedestal 1, was about 2.5. In some discharges, beam injection was started before IP had
PoloidalField at R = 1.52m
ot ’ * * ’ . &. *.!. +’ 2,o 2.5 3.0 3.5 Time (s)
5 10 15 20 FIG. 2. Time history waveformsfor a high,Bpdischarge with IP ramp- 9” down, Shown are (a) the plasmacurrent and neutral beam heating, (b) A and &,,a, (c) the poloidal field at the midplane, on the inboard side of the FIG. 1.d&, for high% dischargesin TFTR plotted as a function of 9*. vacuum vessel,and (d) the midplane H, emission.As&,,, rises,a separa- The high+?, dischargeshave extendedTFTR operation to larger valuesof trix enters the vacuum vessel,as indicated by the reversalof the midplane 9*, eflOdln< 1.6,and B,, J,.,~4.7. Contours ofp,,,, appearas straight lines in poloidal field measuredjust outside the inboard side of the vacuum vessel, this figure. and by a sharp drop in the midplane H, emission.
2278 Phys. Fluids 6, Vol. 3, No. 8, August 1991 Sabbagh et al. 2278 J3,d,P =P,, . The calculated ratio of a,,, /B,, varies through- equilibrium is reached at approximately 2.9 set, and, subse- out the beam heating phase. Modeling of the discharge quently, Ha ( 14”) remains at a large value. The chord at 23” shown in Fig. 2 using the TRANSP code” indicates that the being outside of the range of the strike point, exhibits a more ratio p,,, /ppl reaches a maximum value of 1.65 that occurs standard emission trace. This evolution of the inboard X 10 msec after the start of beam injection. This ratio de- point has also been observed using a visible-light camera that creases to 1.15 by the end of the beam injection, tangentially views the inside of the TFTR vacuum vessel. Figure 2(c) shows the evolution of the poloidal field A subset of the separatrix-limited equilibria has been measured at the midplane on the inboard side of the TFTR modeled using a free-boundary equilibrium code*’ based on vacuum vessel. As Ip is ramped down, and fi, increases, the the technique of Lao and co-workers” that uses external midplane poloidal field decreases and eventually becomes magnetic measurements plus measurements of the local in- negative, indicating that the separatrix has crossed the coil ternal poloidal field from polarimetry of emission from in- position and moved into the vacuum vessel. The separatrix- jected Li pellets.*’ Data from 14 similar discharges were limited discharge is sustained until the end of the beam heat- used to calculate the mean and standard deviation of the ing phase. Figure 2 (d) shows the H, emission, viewed along measurements. Note that the equilibrium reconstruction as- the plasma midplane (0”). As the separatrix enters the vacu- sumes that the measured errors are statistically independent um vessel, and the divertor X point moves into the plasma, and that the pressure is isotropic. Figure 4 shows the recon- the Ha emission at 0” drops to a low level and remains low structed poloidal field profile fitted to the measured internal throughout the separatrix-limited phase. poloidal field data and the poloidal flux contours for the least Additional details of the Ha emission signatures charac- squares “best fit” equilibrium. This technique provides a teristic of the separatrix-limited plasmas in TFTR are shown measurement of the q profile. For the 0.3 MA constant 1, in Fig. 3. Here, the evolution of the plasma boundary from discharge shown in Fig. 4, A = 5.1 f 0.3, fl,, = 1.38 the discharge shown in Fig. 2 is reconstructed from a mag- + 0.12 and q. = 1.6 f 0.55. This is in good agreement with netic analysis code that uses external magnetic measure- A = 5.3 f 0.5 obtained from filament code modeling. The ments and models the plasma current distribution as a col- large calculated Shafranov shift of the magnetic axis agrees lection of current filaments. ‘8,19 Included in the figure are with the axis position inferred from soft x-ray measure- the Ha emission time histories from four chords viewing ments. Measurements of the electron temperature T,, den- different poloidal angles as shown. While the Ha emission sity n,, and ion temperature Ti also exhibit this large shows some fluctuation, no large-scale magnetohydrodyna- outward axis shift. Peak T, = 24 keV and T, = 8.5 keV ( T, mic (MHD) modes or significant loss of plasma stored ener- measured by Thomson scattering) have been achieved in gy is observed to correlate with this activity. At 2.5 set, the these high% TFTR plasmas. plasma is essentially circular and limited on the inner belt limiter. At 2.7 set, the separatrix is entering the vacuum III. STABILITY AND fi LIMIT vessel, the plasma becomes more oblate, and H, (0”) drops. Values of EP, dia > 1.5 have been reached in discharges in There is a similar drop in the Ha emission at 7.5” somewhat which I,, was ramped down as well as in plasmas in which I, later in time as the X point moves further into the vessel and the divertor strike points move to greater poloidal angles. (“vi* ‘~~,‘~‘I”’ This progression of the strike points eventually ceasesas the 0.15 E- (a) uw E a Li Pellet 0.05 3 I? I l-k 1 1.0 5 9 3 J ‘6 g 0.5 0 5 2 E a 9 g 0 0 ,m 5 ,a 5 3 25 tj -0.5 0 5 3 a N -1.0 J 0 2.0 2.5 3.0 3.5 2.0 2.5 3.0 Major radius (m) t 6) 1.4 1.8 2.2 2.6 3.0 3.4 3.8 FIG. 3. Evolution of the plasma outer boundary and H, emission for the R Cm> high-,@,discharge of Fig. 2. The plasma makes a transition from a circular discharge limited on the inner belt limiter, to an ablate separatrix-limited FIG. 4. Free-boundaryequilibrium reconstruction ofa0.3 MA, constant 2, discharge.A drop in the H, emissionis observedat 0” and 7.5”as the separa- discharge. (a) The internal poloidal magnetic field measurementsand the trix enters the vesseland the divertor strike points move to greater poloidal least squares“ best fit” poloidal field profile reconstruction. (b) The poloi- angles. da1flux contours for the equilibrium.
2279 Phys. Fluids B, Vol. 3, No. 8, August 1991 Sabbagh eta/ 2279 was held constant. At values of q* 2 10 (/3,,, 5 2.71, the modes (ELM’s) which appear to cause p saturation or deg- maximum value of eflPdia - 1.6 attained in these plasmas ap- radation. A transition to the limiter H mode generally does pears to be due to a limit imposed by the approach of the not occur at large q* ZZ10 and large e,BPdia2 1. separatrix. At these large values of q*, the maximum value of The enhanced plasma stability, indicated by the ,9,,,diaattained is due to the constraint on eflPdia and is not achievement of increased values of PNdia when I,, is ramped associated with disruption or a soft decay in/3 due to plasma down, appears to be associated with the peaking of the plas- instability. ma current profile, as shown by the increased value of ii, The maximum value of ,BNdia reached in the high% during the high-fiP phase of the discharges. Figure 6 shows TFTR discharges exceeds that of previous TFTR supershot the maximum value ofPNdra as a function of 1,/2 for a subset discharges by a factor of about 1.75. The largest values of of the high% equilibria analyzed by the TRANSP code; data fiNdia were obtained in discharges in which IP was ramped for some supershot discharges are also shown. Here, down. This is illustrated in Fig. 5, where for a subset of the Zi~(B~/2~~~)/(B~/2~,),,,,where(B~),, isthedifferen- high% database, PNdia is plotted as a function of the plasma tial volume average of Bi over the outermost flux surface. current ramp ratio FI,, defined as the ratio of I, before the Cases in which IP was ramped up have smaller values of Zi ramp down, to I, during the neutral beam heating phase. while those in which IP was ramped down have larger values For example, the discharge illustrated in Fig. 2 has of li . The plasma current profile peakedness changes in re- FIp = 2.2. The subset excludes some constant IP discharges sponse to the programmed plasma current and high-& equi- for clarity. No disruptions or large-scale MHD activity were librium effects for each discharge. During 1, ramp-down, observed for high fiP, constant IP plasmas that had the surface voltage reverses in order to decrease the plasma 1 of QPdia 5 1, the discharges have been observed to make a 5.0 / ,- _ : --i-’ ,- , / I /-, / / T-T‘ “, transition into the limiter H mode.23 This transition occurs ’ Disyian es\; more easily in discharges in which I, is ramped downz4 The 1 limiter H mode in TFTR is accompanied by edge-localized 4.0I- - 3.0 ’ ‘i=j i ‘i’ ‘~‘~~o’~ I i1 2.0 ’ - t T I- I X Ip Ramped Down PN dia d j jr\, ../‘ -1 IE31Dismption 1: nn ~T.$cyy$g.@ : ./lp~:,1 0.0 0.5 1.0 1.5 2.0 2.5 3.0 -0.0 0.5 I.0 1.5 2.0 2.5 3.0 FIG. 6. Troyon-normalized diamagneticpas a function of plasmainternal Ip Ramp Ratio (Ippre-ramplIp NBI) inductance.The points representthe maximum /3vd,, attained for eachdis- charge.Trajectory (a) pertainsto a dischargein which I, wasramped down FIG. 5. Troyon-normalized diamagneticflas a function of plasma current from OX5 too.4 MA. Trajectory (b) pertains to a dischargein which I,, was ramp ratio. The maximum value of fl,vd,aincreases as the plasma current ramped down from I.0 to 0.7 MA. The maximum valueoffiy,,,, increases ramp ratio increases. for dischargesin which I, is increased. 2280 Phys. Fluids B, Vol. 3, No. 8, August 1991 Sabbagh etal. 2280 tion. For trajectory (b) , Zp was ramped down from 1.O to 0.7 tions. The beam component of the pressure constitutes 58% MA. The BNdia initially increases as lj increases, but PNdia of the total pressure on axis in the equilibrium in Fig. 7. subsequently decreases coincident with a decrease in Ii. This plasma does not terminate in disruption but rather a soft p IV. ENERGY CONFINEMENT decay leads to the reduction in PNdia. Neutral beam heating Energy confinement times of greater than 3 times L- terminates as pNdia and Ii are decreasing. In other high% mode values have been obtained in these high-& plasmas, cases,PNdia saturates and remains large until the end of the similar to the enhancement reached in supershot plasmas. beam pulse. Figure 8 shows rE normalized to that computed from the The observation of increased pNdia associated with an ITER 89-P energy confinement scaling relation26 as a func- increase in li is similar to results of experiments performed tion of ala, dia for a subset of the high% database. There is a on the DIII-D tokamak.25 This observation does not appear general increase in the confinement enhancement as eflpdia to be machine dependent, since the TFTR experiments were increases, and plasmas in which Z, is ramped down have the performed in a circular vacuum vessel with oblate plasmas largest enhancement factors. Note that the confinement (~-0.7) and Z,, in the range O.O8 3- 6.0 1 - TFlx Shot 46470 z/l/90 SecondStable Region 2- l- zE= MaximumStored Energy I BeamPower 0 llll’t’ll’t’ll’l 0.0 0.5 1.0 1.5 0.0 0.2 0.4 0.6 0.8 1.0 (w- wM!b- V-Q) @p dia J FIG. 8. Energy confinement enhancementas a function of l P,,,,., for high- D, discharges.The largest enhancementfactors occur in plasmasin which FIG. 7. High-n ballooning stability for a high% dischargein which I, is 2, was ramped down. The representativeerror bars shown arise from the ramped down. The plasma pressuregradient is in the first region of high-n uncertainty in the stored energy as calculated from magnetics measure- ballooning stability. ments. 2281 Phys. Fluids B, Vol. 3, No. 8, August 1991 Sabbagh eta/. 2281 V. NONINDUCTIVE CURRENTS AND CURRENT largest bootstrap current fractions were calculated to occur PROFILE CONTROL in Z, ramp-down cases.The surface voltage remains negative The high-fiP plasmas in TFTR have low collisionality for the duration of the beam pulse, relaxing from less than - for both ions and electrons. Under these conditions, a signifi- 1 to about - 0.2 V. A total of 0.35 Wb of poloidal flux cant amount of bootstrap current should contribute to Z,. was measured to be removed from the plasma between 3-4 Shown in Fig. 9 is the time evolution of Z,, surface voltage, and 4.0 sec. The classical time constant associated with this and pP diafor a discharge in which ZPwas ramped down from change is about 1 sec. 1.O to 0.6 MA. The parameters e and Y? for this discharge at 3.9 set are 0.1 and 0.02 at one-half of the minor radius. VI. NEUTRON PRODUCTION AND FUSION POWER GAIN Included with the time history waveforms are the noninduc- tive current components of ZPand the surface voltage calcu- Neutron production in the high% discharges at the lated by TRANSP. As the current is decreased before the beam larger values of Z, (0.85-l .O MA) is comparable to TFTR injection, the surface voltage swings negative, removing 0.29 supershot performance. Shown in Fig. 10 is the maximum Wb of poloidal flux from the plasma between 3.0 and 3.25 neutron rate S’,, plotted as a function of maximum beam sec. The plasma current profile is primarily affected in the power Pbeammax for discharges with Z,>O.S MA and outer l/3 of the plasma, and the classical time constant asso- ePpdia)0.7. Also shown is the nominal supershot neutron ciated with this change is about 0.3 sec. Since the coinjected production performance as a function of beam power.29 Dis- beams are switched on first, there is initially a large fraction charges that have S, within 20% of this curve have of codirected current driven by the beams. In this discharge, 0.75+$&3,, ( 1.12. At a plasma current of about 0.85 MA, there is 37% more counter-injected power than coinjected neutron rates of ( 1,5-l .9) X lOI6 set - ’ have been achieved power during the high-flP phase. Since the coinjected beam- during the discharge and were sustained for 0.3-0.4 set until lines drive current more efficiently than the counter beam- neutral beam heating was terminated. For example, lines becauseof orbit effects, this produces no net beam driv- S,, = 1.7 X lOi set- ’ was reached at ZP = 0.82 MA, en current. The bootstrap current is calculated to be 65% of E,,, = 2.3 MJ, eflPdin = 0.89, and P&,, max= 19 MW. The the total ZP in this case. Similar results were obtained for corresponding deuterium-deuterium fusion power gain discharges with balanced neutral beam injection. In these QbD is 1.1 X 10e3.AtZ,, = 0.97MA,S, = 2.63X 10’6sec-’ calculations, any potential contribution to the bootstrap cur- was reached transiently with E,,, = 2.8 MJ, EPpdia= 0.85, rent due to the anisotropic beam-particle pressure gradient and Lam max= 24 MW, yielding Q,, = 1.3 X lo- 3. This has been neglected. The beam-particle pressure constitutes QuD is 70% of the best value achieved in TFTR, obtained in approximately 45% of the total plasma pressure in the plas- a discharge that had ZP= 1.9 MA. ma core at the end of the discharge. At a fixed plasma current, there is a large increase in the In both constant current, and ZPramp-down cases, there neutron production but only a small gain in Q,, in high% is a significant reversal of the surface voltage during the discharges as compared to supershots where the domains of high-/3, portion of the discharge, as shown in Fig. 9. The these discharges overlap. At Z, = 0.8 MA and &,* max= 14 MW, supershot discharges yield S, = 1.0X lot6 see-‘. At Z, = 1.0 MA and Pk,, max = lg MW, S,, = 1.6~ lOI set - ’ has been reached. The high+,, discharge mentioned above produced more neutrons than 3.0 ,,t,-. ?,[I ,/C.//I, / / 3.5 4.0 4.5 5 10 15 20 25 30 Time (s) Maximum Beam Power (MW) FIG. 9. Calculatednoninductivecurrent contributions tol,. (a) I,, and the calculatednoninductive beam-and bootstrap-drivencomponents ofl,, (b) FIG. 10.Maximum neutron rate asa function of maximum beampower for the measuredand calculated plasmasurface voltage, and (c) the evolution dischargeswith 1,20.5 MA and ePpdla20.7. Thecurve describingthe nomi- of&J,, and neutral beam heating. The bootstrapcurrent constitutes up to nal neutron rate for TFTR supershotsis superimposedfor reference.Neu- 65% ofl, for this discharge.The calculatedsurface voltage waveform is in tron production in high-& dischargeswith I,>O.SS MA is comparableto good agreementwith the measuredwaveform. supershotdischarges. 2282 Phys. Fluids B, Vol. 3, No. 8, August 1991 Sabbagh eta/. 2282 that achieved in the I, = 1 MA supershot by greater than a transformeronly providescontrol of theplasma current den- factor of 1.6. This increase in the neutron rate comes about sity in the outer portion of the plasma. A technique such as from the larger beam power that can be tolerated at a given lower-hybrid current drive, which can drive current effi- Ip in the high;B, discharges before a plasma disruption oc- ciently near the plasma edge may therefore be sufficient to curs. To obtain S, equal to that of the high% discharge with provide and sustain a current profile shape leading to a stable I,, = 1 MA, the plasma current in the supershot case needs plasma exhibiting enhanced fiN performance. Since this sys- to be raised to about 1.35 MA. The neutral beam heating tem would be operating in a plasma with reduced I, and only power for this supershot case with equivalent S, would be 22 in a region of relatively small current density, the power MW, slightly less than was used in the I, = 1 MA high% requirements could be kept low. Another key consideration case. Note that Q,, is 1.2X 10e3 in the I, = 1 MA super- is the role ofthe bootstrap current, which will drive codirect- shot discharge, which is essentially equal to the value ed plasma current, and hence may reduce the externally reached in the high pp discharge at the same plasma current driven current required for steady-state operation. However, but at larger beam power. the bootstrap effect will broaden the current profile and re- duce li which, in the present experiments, is associated with VII. SUMMARY AND DISCUSSION a reduction of the fiN at which the tokamak can operate in a The operating range of TFTR plasmas has been expand- stable fashion. The competing nature of these two effects ed to include plasmas with EP, dia up to 1.6 and fl,vdia up to needs to be carefully considered in high-p tokamak designs 4.7. The range of I,, during neutral beam injection was 0.28 that would operate in the first stability region. A more de- MA 2283 Phys. 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