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JET–P(97)47 Profile Control Experiments in JET using Off-axis Lower Hybrid Current Drive A Ekedahl1, Y Baranov, J A Dobbing, B Fischer, C Gormezano, T T C Jones, M Lennholm, V Parail, F Rimini, J A Romero, P Schild, A C C Sips, F X Söldner, B Tubbing. JET Joint Undertaking, Abingdon, Oxfordshire, OX14 3EA, 1 Also: Institute for Electromagnetic Field Theory and Plasma Physics, Chalmers University of Technology, S-412 96 Göteborg, Sweden. Preprint of a Paper to be submitted for publication in Nuclear Fusion December 1997 “This document is intended for publication in the open literature. It is made available on the understanding that it may not be further circulated and extracts may not be published prior to publication of the original, without the consent of the Publications Officer, JET Joint Undertaking, Abingdon, Oxon, OX14 3EA, UK”. “Enquiries about Copyright and reproduction should be addressed to the Publications Officer, JET Joint Undertaking, Abingdon, Oxon, OX14 3EA”. ABSTRACT. In lower hybrid current drive (LHCD) experiments in JET, up to 7.3 MW LH power has been coupled to X-point plasmas, resulting in sawtooth suppression and full current drive up to 3 MA. The current drive efficiency reached ηCD = 0.26×1020 m-2A/W in these experiments. The LH power deposition and driven current profiles can quite well be reproduced with the raytracing + Fokker-Planck code in JET, even when multi-pass absorption of the LH wave is dominant. Profile control with off-axis LHCD has been used for increasing q above 1, thereby suppressing sawteeth before and during the ELM-free hot ion H-mode. In optimised shear plasmas, a broad q-profile with central negative shear was formed with moderate LH power (≤ 2 MW), applied in the low density phase during the current ramp-up. An internal transport barrier with improved electron confinement was produced, which resulted in a peaking of the electron temperature profile and Te0 up to 10 keV at n e ≈ 1.0×1019 m-3. LHCD has assisted in producing the target q-profile required for high fusion yield in the optimised shear experiments. 1. INTRODUCTION Lower hybrid current drive (LHCD) has shown to be the most efficient of the various methods for non-inductive current drive in tokamaks so far and it has been used for current profile control in several experiments. LHCD has the unique property among the known current drive methods to have high efficiency at low plasma pressure β [1] and could therefore be considered for current drive and profile control in the low-β phase during the plasma current ramp-up in ITER. Modelling of ITER operation scenarios have shown [2] that a q-profile with reversed magnetic shear can be produced and that q can be maintained above 1 over the whole plasma profile for several hundred seconds with a proper combination of LHCD and central heating with fast waves or electron cyclotron waves, thereby suppressing sawteeth during the burn phase. LHCD can also be suitable for creating and maintaining the hollow plasma current profile required in the so-called “advanced tokamak” scenario in ITER, where full non-inductive current drive with large bootstrap fraction is envisaged [1, 2]. In recent years, LHCD experiments have aimed at tailoring the plasma current profile in order to decrease the magnetohydrodynamic (MHD) activity and create or a maintain a profile with negative central magnetic shear. For example, a stationary reversed magnetic shear profile was produced with LHCD in Tore Supra [3]. To achieve this the plasma parameters were chosen in such a way as to limit the LH wave penetration to the outer region of the plasma. In JT60-U [4] the reversed magnetic shear configuration, initially produced with neutral beam injection (NBI), was sustained for 7.5 s with LHCD. It was also demonstrated that the region of reversed magnetic shear could be modified by changing the parallel refractive index, n||, of the launched waves. 1 In this paper we present results of the LHCD experiments carried out in the pumped divertor phase of JET, which started in 1994. In the following section we give a brief description of the LHCD system and in Section 3 we analyse the current drive efficiency in experiments with full LH current drive. In Section 4, LH fast electron profiles are described and compared with the results of the ray-tracing + Fokker-Planck code in JET [5], for different plasma parameters. Section 5 and Section 6 are devoted to profile control experiments with LHCD, sawtooth stabilisation and magnetic shear reversal, respectively. The use of profile control with LHCD in conjunction with the high performance experiments, i.e. the “hot ion H-mode” scenario [6-8] and the “optimised shear” scenario [9, 10], is presented. Finally, Section 7 summarises the results. 2. DESCRIPTION OF THE LHCD SYSTEM IN JET The full 3.70 GHz LHCD system [11] has been in operation in JET since 1994. This followed an experimental campaign with the prototype LHCD system [12], which was used in the experiments in 1990-1992 in the limiter configuration of JET. The full LHCD system has capability to launch 10 MW to the plasma for 20 s. The launcher is composed of 48 multijunctions, each splitting into eight narrow waveguides with the dimensions 9 mm × 72 mm, so that the launcher mouth consists of 384 narrow waveguides in 12 rows and 32 columns. Built-in π/2phase shifters in the small waveguides provide a narrow n||-spectrum with high directivity. The peak of the main lobe in the n||-spectrum can be varied between 1.4 and 2.3 by varying the phase difference between adjacent multijunctions. The majority of the LHCD experiments, and all the results presented in this paper, have been carried out with 0° phasing between multijunctions, which gives a n||-spectrum centred at n||0 = 1.84, with full width ∆n|| = 0.46. The power handling of the launcher in the experiments depends on the conditioning of the waveguides [13] and the electric field in the multijunctions [14]. Good coupling of the LH waves has been obtained by using a feedback loop, which adjusts the radial position of the launcher in real time in order to maintain the requested value of the power reflection coefficient, which is usually in the range 3%-5% [15, 16]. In recent experiments, gas injection near the launcher has provided good coupling (R < 5%) of the LH power at a plasma-launcher distance up to 8 cm [16, 17]. By using this method, which was first developed in ASDEX [18] and later used in JT-60U [19], good coupling conditions can be provided even when the launcher is retracted behind the poloidal limiter. This reduces the risk of high heat load on the launcher and direct plasma-launcher interaction. Even with good coupling of the LH waves and with conditioned waveguides the power handling is limited, as discussed in [13, 14]. The maximum LH power coupled to divertor plasmas has reached 7.3 MW, using 8.2 MW generator power. In profile control experiments, 13 s long LH pulses have been applied, delivering a maximum energy of 68 MJ to the plasma [15]. Real time control of the LHCD power has 2 been employed in order to control the plasma surface loop voltage and the second current moment, which is related to the internal inductance [20]. 3. FULL CURRENT DRIVE EXPERIMENTS Full current drive with LHCD alone has been achieved in X-point configuration with plasma current, IP, between 0.7 MA and 3 MA, line average electron density, n e < 2.0×1019 m-3 and with high toroidal magnetic field, BT > 3 T. The resistive part of the surface loop voltage, Vres, is derived from measurements of the poloidal magnetic flux at different poloidal locations. The inductive contributions arising from varying magnetic energy are taken into account as (MA) 3.0 Ip 2.8 6 (MW) PLH 3 0 2 1 ne 1.0 0.9 `i Vres 0.5 JG97.513/1c This is shown in Fig. 1 for a 3 MA discharge. When full current replacement is obtained the current drive efficiency can be defined as η CD = n e I p R 0 PLH , where R0 is the plasma major radius at the magnetic axis. Fig. 2(a) shows ηCD against volume averaged electron temperature, 〈Te〉, for discharges fulfilling the criterion Vres ≤ 0.05 V. The variation in Zeff was small and has not been taken into account in the graph. For line average electron density typically above 1.1×1019 m-3, the current drive efficiency increases with electron temperature Pulse No. 34426 (1019 m–3) where L is the plasma inductance and Vloop the total surface loop voltage. Full replacement of the plasma current is obtained when Vres is zero. 1 d 1 2 LI P , I P dt 2 (V) Vres = Vloop − 0 10 12 14 16 Time (s) FIG. 1. Full current drive with LHCD in a 3 MA discharge. up to 〈Te〉 ≈ 2 keV and reaches ηCD = 0.26×1020 m-2A/W. Above 〈Te〉 ≈ 2 keV the efficiency decreases. At low electron density (n e < 1.1×1019 m-3) the current drive efficiency is generally lower than at higher density. In Fig. 2(a) there are two groups of data points at low density. The points at 1.5 < 〈Te〉 < 2.0 keV correspond to full current drive data at 1.5 MA and the points at 2.5 < 〈Te〉 < 3.0 keV are from full current drive in ≥ 2 MA discharges. Within these two groups there is a tendency of ηCD to decrease with temperature. In Fig. 2(b) we show the theoretical current drive efficiency, obtained from calculations with the ray-tracing + Fokker-Planck code, using the measured plasma parameters. No effect of residual electric field has to be taken into account. At 〈Te〉 ≈ 2 keV, the calculated efficiency saturates and at low density the efficiency drops slightly. This can be explained by the modification of the absorption at high temperature and low density. Once the electron distribution 3 function has developed a saturated plateau, the wave energy will be damped on lower energy electrons, which results in lower current drive efficiency. For n e > 1.1×1019 m-3 the calculated and experimental efficiencies are in good agreement. At low density the experimental efficiency appears lower than what is expected from the calculations. The low current drive efficiency observed experimentally (Fig. 2(a)) at low electron density and especially at high LH power (PLH > 4 MW), can be attributed to local overdriving of the current. In such conditions, a negative electric field inside the plasma is produced which tends to decelerate the LH produced suprathermal electrons, thereby reducing the current drive efficiency. In these discharges it has also been observed that the hard X-ray emission from fast electrons is lower than expected. To analyse these discharges further, the radial profile of the parallel electric field was calculated by applying the technique described in [21]. The method utilises the radial current profile, determined by the equilibrium reconstruction code EFIT, from magnetic measurements and constraints on infrared polarimetry from two vertical chords through the plasma core, at radii R = 2.70 m and R = 3.02 m (R0 = 2.96 m). The parallel electric field profile is obtained by evaluating the time derivative of the poloidal magnetic flux. Fig. 3(a) 2 shows the loop voltage against normalised poloidal flux, ψ/ψa ≈ (r/a) , for a 2 MA discharge with n e = 0.7×1019 m-3, 〈Te〉 = 3.0 keV and PLH = 5.5 MW. A negative loop voltage is obtained inprofile was used as input in the calculation of the discharge with a 2D Fokker-Planck code, JG96.249/3c 0.4 Experimental (a) ne R Ip PLH 0.3 0.1 Pulse No: 35303, t = 11.0s a) Vres (V) 0.1 Vres from magnetics and EFIT analysis 0 Calculated (b) ne R ILH PLH 3.0 b) Vres = 0 ILH > 5MA 0.3 2.0 0.1 ne > 1.1 x 1019 m–3 ne < 1.1 x 1019 m–3 0 0 1.0 2.0 3.0 <Te> (keV) FIG. 2. (a) Experimental current drive efficiency. (b) Current drive efficiency calculated with the ray-tracing + Fokker-Planck code for the measured parameters. Zero DC electric field is assumed in the calculations. 1.0 Vres < 0 ILH = 2.3MA 0 0 0.2 0.4 0.6 JG96.264/3c 0.2 4 0 –0.1 jLH (MA/m2) LHCD efficiency (1020 m–2 A/W) 0.2 0.8 ψ/ψa FIG. 3. (a) Loop voltage profile from magnetics and EFIT analysis. (b) LH driven current profiles from 2D FokkerPlanck calculations, both with Vres < 0 (from the analysis) and Vres = 0. 1.0 the region 0.1 < ψ/ψa < 1.0. This loop voltage based on the model in [22]. The hot conductivity term, which becomes significant here, reached a maximum of 1.6×σSpitzer at ψ/ψa = 0.6. When the negative DC electric field was introduced in the calculations the resulting LH driven current was 2.3 MA. For comparison, when no DC electric field effect was taken into account the LH driven current amounted to > 5 MA (Fig. 3(b)). In summary, for densities typically in the range n e = 1.1-2.0×1019 m-3, the current drive efficiency increases with 〈Te〉 to reach ηCD = 0.26×1020 m-2A/W. In this range the experimental and calculated efficiencies are in good agreement. For n e < 1.1×1019 m-3 and with high LH power, local overdriving of the current appears to take place, which results in a negative toroidal electric field being produced. 4. LH DEPOSITION PROFILES The LH power deposition and driven current profiles have been studied in a wide range of plasma parameters: plasma current IP = 0.7-3 MA, toroidal magnetic field BT = 1.8-3.4 T, line averaged electron density n e = 0.7-4.0×1019 m-3 and volume averaged electron temperature 〈Te〉 = 0.5-3 keV. The fast electron bremsstrahlung (FEB) diagnostic [23] is the tool for determining the LH deposition profile in the LHCD experiments. This diagnostic provides information about the location and energy distribution of the fast electrons, created by Landau damping of the injected LH waves. The FEB diagnostic detects hard X-rays with energy between 133 keV and 400 keV in four energy windows along ten horizontal and nine vertical lines of sight. Abel inversion of the line integrated brightness profiles yields the local emissivity as a function of the normalised poloidal flux, ψ/ψa. The profiles obtained from the horizontal detectors are mainly used in the analysis. One limitation of the diagnostic is that it only gives reliable data in discharges with low neutron yield and can therefore not be used during high power NBI or ICRH. Further, the low energy limit for photon detection is 133 keV, which excludes information about the low energy part of the electron distribution function. The fast electron distribution function and LH power deposition is modelled with the raytracing + Fokker-Planck code in JET. The code has been validated on data from several LHCD experiments, as presented in [5]. The code uses experimental profiles of electron density and temperature. The plasma equilibrium is obtained from the EFIT reconstruction. In the modelling calculations, the trajectories of approximately 600 rays are followed. The rays are launched from different poloidal angles, representing the poloidal extent of the grill, and with a range of initial n|| that covers the main lobe of the launched n||-spectrum, as calculated by the SWAN code [24]. The damping of the waves along the trajectories is calculated with a 1D or 2D FokkerPlanck code. The code also allows to calculate the hard X-ray emission from the suprathermal electrons, which can then be compared with the measured FEB emission to validate the code. The propagation and absorption of the LH waves in JET plasmas are characterised by multi-pass absorption, i.e. the wave encounters many reflections at the plasma boundary before 5 it is absorbed. This process can change the parallel refractive index of the wave considerably. Analysis of FEB profiles from the LHCD experiments have shown that the fast electron profile shifts off-axis with increasing plasma current and decreasing toroidal magnetic field, which can be translated into a dependence on the edge safety factor, qa. This dependence can be explained as follows. As the plasma current, and therefore poloidal magnetic field, increases the launched wave spectrum spreads to higher poloidal and parallel wave numbers, i.e. n|| increases and the Landau damping condition is fulfilled further out in the plasma where the electron temperature is lower. Similar behaviour was noticed earlier in JT-60U [25]. Fig. 4 and Fig. 5 show the measured FEB profiles and calculated LH current profiles in a 2 MA / 2.8 T and a 3 MA / 3.2 T discharge, respectively, where the safety factor at 95% of the poloidal flux (q95) was 4.6 and 3.5. Fig. 4(a) and Fig. 5(a) show the local FEB emissivity normalised to density, which is equivalent to current density profiles. Fig. 4(b) and Fig. 5(b) show the LH driven current profiles calculated with ray-tracing + 1D Fokker-Planck code. Good agreement between the measured and calculated profiles is obtained. Pulse No. 34443, Ip = 3 MA, BT = 3.2T Pulse No. 35309, Ip = 2 MA, BT = 2.8T (10–9 ph/keV/s) (10–9 ph/keV/s) FEB FEB 8 4 20 10 0 1.0 0 0 0.2 0.4 Ψ/Ψa 0.6 0.8 1.0 FIG. 4. (a) FEB emissivity profile normalised to density and (b) LH driven current profile from ray-tracing calculation for a discharge with IP = 2 MA, BT = 2.8 T, ne = 19 -3 1.9×10 m and 〈Te 〉 = 1.7 keV (#35309). Ray-tracing 0.5 JG97.411/7c 0.5 JLH (MA/m2) Ray-tracing JG97.411/6c JLH (MA/m2) 1.0 0 0 0.2 0.4 Ψ/Ψa 0.6 0.8 1.0 FIG. 5. (a) FEB emissivity profile normalised to density and (b) LH driven current profile from ray-tracing calculation for a discharge with IP = 3 MA, BT = 3.2 T, ne = 19 -3 1.9×10 m and 〈Te 〉 = 2.0 keV (#34443). At high electron density (n e > 3.0×1019 m-3) and steep density gradient near the plasma edge, the propagation of the LH waves can become limited to the outer plasma region, as anticipated from the accessibility criterion. Fig. 6 shows the profiles of electron density and temperature at two different time slices in a 2.5 MA discharge with BT = 2.6 T. The results of the raytracing calculations for these two time slices are shown in Fig. 7(a) and Fig. 7(b), respectively. The poloidal projections of several ray trajectories with initial parallel refractive index n|| = 1.85, 6 Pulse No. 30566 4 ne(1019 m–3) ne 3 2 t=16.5s t=18.5s 1 3 Te 1 JG97.513/5c Te(keV) 2 0 0 0.2 0.4 0.6 ψ/ψa 0.8 1.0 FIG. 6. Profiles of electron density (a) and temperature (b) for two different time slices in a 2.5 MA discharge with BT = 2.6 T (#30566). 2.0 starting from different poloidal angles, are shown. In the case with higher edge density and lower edge temperature (Fig. 7(a)), the waves are trapped near the edge and propagate towards the X-point region, where they encounter a large upshift in n||. Similar results have been observed in JT-60U [26]. When the edge density decreased and edge temperature increased, the penetration of the waves to the plasma interior improved, Fig. 7(b). The corresponding measured FEB profiles and calculated LH current profiles for the two cases are shown in Fig. 8. For the first time slice, corresponding to Fig. 7(a), the power was absorbed at n|| ≈ 4-8, which resulted in very low FEB emission. Pulse No: 30566, t = 16.5s 2.0 1.5 Pulse No. 30566, t=18.5s 1.5 1.0 1.0 0.5 0 Z (m) Z (m) 0.5 0.0 –0.5 –0.5 –1.0 2.0 2.5 3.0 R (m) 3.5 4.0 nIIo = 1.85 –1.5 2.0 2.5 3.0 R (m) 3.5 JG97.411/1c nIIO = 1.85 –1.5 JG97.94/2c –1.0 4.0 FIG. 7. Ray-tracing calculations for the two different electron density and temperature profiles shown in Fig. 6. Analysis of the experimental FEB profiles have shown that the LH power deposition and driven current profile are normally peaked off-axis in the JET experiments. Off-axis deposition is promoted by the non-circular, elongated plasma cross section, compared with circular plasmas. The wave spends a long time in the plasma edge region as it propagates along the magnetic field 7 lines and can there experience a large number of reflections and increase in n||. This behaviour can be modelled with the ray-tracing code. Quite good agreement between the measured and calculated shapes of the FEB profiles is obtained in the experiments. Pulse No. 30566 t=16.5 (10–9 ph/keV/s) 6 a) FEB 4 (x5) 2 0 0.3 c) Ray tracing d) Ray tracing 0.2 0.1 JG97.513/4c jLH (MA/m2) t=18.5 b) FEB 0 0 0.5 ψ/ψa 1.0 0 0.5 1.0 ψ/ψa FIG. 8. Measured FEB emissivity profiles normalised to density (a, b) and calculated LH current profiles from ray-tracing (c, d) for the cases shown in Fig. 6 and Fig. 7. 5. SAWTOOTH STABILISATION 19 -3 Sawteeth have been fully suppressed in 2 MA discharges with n e < 2.0×10 m and in 3 MA discharges with n e < 1.3×1019 m-3, i.e. in conditions close to full current drive. The time evolution of q0 and the sawtooth activity is shown in Fig. 9 for a 3 MA discharge. In some 3 MA discharges the suppression of sawteeth is followed by a peaking of the electron temperature profile, with central electron temperature, Te0, reaching above 8 keV [27]. The evolution of the current profile during long (> 10 s) LHCD pulses has been modelled with the JETTO transport code [28], using the LH current profile from ray-tracing calculations as input. The predictive transport code modelling can well reproduce the behaviour of the internal inductance and the surface loop voltage [29]. LHCD has been used to study the effect of sawtooth stabilisation and current profile control of the target profile for ELM-free hot ion H-modes. In the hot ion H-mode scenario [6-8], high power neutral beam injection (NBI) is applied to a low density, high triangularity divertor plasma. An ELM-free H-mode is produced, in which the central ion temperature can reach above 20 keV and exceed the central electron temperature by a factor of ~2. The high fusion performance phase lasts for 1–2 s and is then terminated by a deterioration of the confinement, which is associated with a variety of MHD activity. The most common types of termination are giant ELMs, sawteeth and outer modes [8]. In order to investigate the effect of current profile 8 Pulse No. 35006, NBI+LHCD Pulse No. 35010, NBI only Pulse No. 33166, Ip = 3MA, BT = 3.3T (MW) PICRH (keV) 1.0 (1016 s–1) ECE 5 (a.u.) Teo 4 6 2 2 0.9 16 17 18 19 Time (s) FIG. 9. Sawtooth stabilisation in a 3 MA discharge. Te0(ECE) RDD 1 0 10 Dα (35006) 5 0 10 q0 (a.u.) 1.0 PNBI 0 10 neo 1.5 10 PLH 5 Dα (35010) JG97.513/3c 0 2.0 6 (keV) 10 PLH 2.5 JG97.411/13c (a.u.) (1019m–3) (MW) 5.0 5 0 15 16 17 18 Time (s) FIG. 10. Time evolution of sawtooth activity, neutron rate and Dα-signal in hot ion H-modes with preceding LHCD (#35006) and without (#35010). The full length of the LH pulse is not shown. modification, a 5 s long LHCD pulse at PLH ≈ 5 MW was applied in a 3 MA plasma at low density (n e ≈ 1.0×1019 m-3). During LHCD the internal inductance decreased and the sawteeth were suppressed. The hot ion H-mode was then produced by high power neutral beam injection (PNBI ≥ 10 MW). In the first set of experiments the ELM-free H-mode period was reduced when preceding LH profile control was used. This was linked to the decrease in edge shear and triangularity, which is a consequence of the reduction in li and broadening of the current profile [27]. Since it is known that high triangularity results in longer ELM-free periods [30], the plasma configuration was modified for the further experiments so as to maintain a higher edge shear during the LHCD phase. Following this, the length of the high performance ELM-free period was similar both with and without preceding LHCD. Fig. 10 shows the result of the second experiment. In the discharge with LHCD (#35006) the sawteeth were stabilised during the NBI phase, while in the discharge without LHCD (#35010) large sawteeth were present. In the discharge with NBI only a large ELM terminated the high performance phase and the neutron rate decayed rapidly afterwards. With preceding LHCD profile control the high performance phase ended by weak MHD activity, followed by a saturation in the neutron rate instead of a decay (Fig. 10). Although the neutron rate was not increased significantly when LHCD was used, the avoidance of sawteeth improved the reliability of the discharges. The second experiment was carried out with limited NBI power and studies of the q-profile variation with LHCD before hot ion H-modes with higher NBI power (~18 MW) remain to be done. 9 6. MAGNETIC SHEAR REVERSAL Improved core confinement and high fusion yield have been obtained in plasmas with reversed or weak magnetic shear in several tokamaks [9, 10, 31–33]. In JET, a D-D reaction rate of 1.1×1017 s-1 has been achieved in the optimised shear scenario, with 18 MW NBI + 6 MW ICRH [9, 10]. Crucial for the performance in these discharges is the q-profile, which is tailored early in the discharge by using an initial fast current ramp rate (~1.5 MA/s), followed by a slower ramp (~0.4 MA/s) up to 3 MA. Further modification of the current profile can be done by applying LHCD during the current ramp-up phase. The effect of LHCD on the q-profile has been studied in the current ramp-up phase of optimised shear type plasmas, using LH power between 1 MW and 3 MW. In these particular experiments the launcher was retracted by 5-10 mm in the shadow of the poloidal limiters and good coupling with ≤ 5% reflection coefficient on all rows was provided by using local gas injection. Between 50% and 100% of the total gas flow required for the main plasma density control was supplied by a gas injection valve in the vicinity of the LH launcher, corresponding to a gas flow of 2.5-5×1021 el./s. The time evolution of the plasma parameters is shown in Fig. 11. The LH power followed a waveform of the requested coupled power by using the real time power control system. A broad current profile with negative central shear was produced. Fig. 12 shows the q-profile obtained from the equilibrium reconstruction for the discharge in Fig. 11 (#39537), compared with the q-profile in a reference discharge with ohmic heating alone. IP 2 1 PLH 0 8 7 <R> 4 t = 4.0 s 0 8 6 Total Gas flux q 4 Near grill 0 5 neo 1.5 1.0 ne 0.5 10 4 <Te> 0 1 2 3 4 5 Time (s) FIG. 11. Time evolution of plasma parameters for a discharge with LHCD pre-heat in the current ramp-up of an optimised shear type discharge. The high power NBI + ICRH phase begins at t = 5.0 s. JG97.411/10c Teo 5 0 10 Pulse No. 39537, LHCD Pulse No. 39515, Ohmic 8 JG97.411/9c (keV) (1019 m–3) (1021 el./s) (%) (MA, MW) Pulse No. 39537 3 0 0.2 0.4 0.6 0.8 1.0 r/a FIG. 12. q-profiles at t = 4.0 s in optimised shear type discharges with LHCD (#39537) and ohmic heating only (#39515). Negative central shear is obtained with PLH ≈ 2 MW. An internal transport barrier with improved electron confinement was obtained in these discharges. This resulted in peaking of the electron temperature profile and central electron temperature reaching 10 keV [17]. Similar electron heating has been observed in reversed magnetic shear discharges with LHCD in Tore Supra [34]. The value of Te0 and the time of the formation of the electron transport barrier varied with LH power, as seen in Fig. 13. In addition to the strong peaking of the electron temperature profile, a slight peaking of the electron density profile was observed, as well as a small increase in ion temperature and neutron rate. Pulse No. 39272 Pulse No. 39275 Te (keV) 1 c PLH 4 2 b a) 0 1 a b) Teo c 5 JG97.411/11c b a 0 2 3 4 Time (s) 5 6 FIG. 13. Evolution of central electron temperature, from LIDAR Thomson scattering measurements, in three discharges with different LH power. 6 4 2 JG97.513/2c 0 10 (keV) t=3s t=4s t=5s 6 0 2 (MW) Pulse No. 39274 Ip (10–9 ph/keV/s) (MA) 2 Pulse No. 39274 0 0 0.2 0.4 0.6 0.8 1.0 r/a FIG. 14. Electron temperature profiles from LIDAR (a) and Abel inverted FEB profiles (b) for different time slices of #39274. Transport code calculations have been carried out for discharge #39274, in which a spontaneous transition to electron temperature profile peaking occurred at approximately 5 s in the discharge (Fig. 13). The experimental measurements of the electron temperature profile and fast electron profile for different time slices are shown in Fig. 14. Fig. 14(b) shows the Abel inverted FEB profiles for electrons with energy in the range 133 keV to 200 keV normalised to electron density, which equivalent to a current profile. When the experimental LH current profiles were used in the transport code, the resulting effective heat conductivity, χeff, became negative. In order to reproduce the observed electron heating while maintaining χeff > 0, more centrally peaked LH current profiles were required. Fig. 15(a) shows the resulting χeff-profiles and Fig. 15(b) shows the LH driven current profile required in order to reproduce the electron heating. A possible explanation is that the profile of the LH current contained in the fast electrons with energy below 133 keV, which is not measured with the FEB diagnostic, is more centrally peaked and is contributing to the central electron heating. A heat pinch is another explanation that can be invoked. 11 0.4 0.2 0 0 JG97.145/16c jLH (MA/m2) (b) 0.2 0.4 0.6 0.8 r/a FIG. 15. Result of transport code calculations for #39274. (a) Radial profiles of the effective heat conductivity, χeff,. (b) Profiles of the LH driven current, calculated by the transport code, required in order to reproduce the experimental electron temperature profiles. 1.0 (1016s–1) (keV) (1019m–3)(MW) 0 2 Pulse No. 40554 Ip 0 1 PLH 0 5 PICRH 0 20 10 0 PNBI 5 ne0 ne 0 20 Ti0 Te0 0 10 5 0 JG97.145/13c (MA) (MW) 2 (MW) χeff (m2/s) Pulse No.39274 (a) t = 3s 4 t = 4s t = 5s RDD 0 2 4 Time (s) 6 8 FIG. 16. Scenario for high fusion yield in the optimised shear experiments. The effect of a q-profile modification by LHCD on the formation of the internal transport barrier for ions during the high power NBI + ICRH phase in the optimised shear experiments has not yet been systematically studied. There is experimental evidence, however, that both a broad, hollow q-profile obtained with a 5 s long LH pulse at ~2 MW and a monotonic q-profile obtained with ohmic heating alone caused a delay of the formation of the ion transport barrier [10]. On the contrary, a prompt triggering of the ion internal transport barrier at the application of the high power heating was obtained with a short LHCD phase at 1 MW applied immediately after the current formation, followed by ICRH to freeze the q-profile and high power injection at 5.0 s in the discharge (Fig. 16). This scenario resulted in a D-D reaction rate of 1.1×1017 s-1 and has also been used in the recent D-T experiments in JET. In this scenario LHCD assists in the plasma breakdown and decreases the internal inductance. The ultimate goal of the optimised shear experiments in JET is to demonstrate quasi steadystate operation with improved confinement. Further development includes more use of LHCD, in particular during the high performance phase, in order to try to maintain a flat or slightly hollow q-profile throughout the high power heating phase. 7. SUMMARY Full current drive up to 3 MA has been achieved in X-point plasmas in JET with dominating offaxis LH power deposition. For electron density in the range n e = 1.1-2.0×1019 m-3, the current drive efficiency increases with electron temperature up to ηCD = 0.26×1020 m-2A/W. This value 12 is in accordance with the requirements for advanced tokamak scenarios in ITER, where operation with large bootstrap fraction of the current is envisaged. Below n e < 1.1×1019 m-3 and at high LH power (PLH > 4 MW) local overdriving of the current takes place. The hard X-ray emission profiles show that the LH power deposition and driven current profiles become broader with increasing plasma current and at IP ≥ 2 MA the LH driven current profile is broader than the ohmic profile. The experimental profiles can be quite well reproduced by the ray-tracing + Fokker-Planck code, using the JET magnetic equilibrium and measured plasma parameters, even if multi-pass absorption of the wave is dominant. The generally good agreement between experiment and model calculations shows that the model may also be used for predicting the LH power deposition and current drive efficiency in ITER, which due to higher electron temperature will be characterised by single-pass absorption. LHCD experiments in JET have shown that current drive with lower hybrid waves can be used as a tool for improving performance via current profile control: • • • • Sawteeth have been fully suppressed, following the increase in q above 1, in discharges up to 3 MA, n e < 2.0×1019 m-3 and BT > 3 T. Stabilisation of sawteeth before and during hot ion H-modes have been achieved with LH profile control, which improved the reliability of the high performance discharges. Broadening of the q-profile and central negative magnetic shear have been obtained with low LH power (~2 MW) during the current ramp-up phase at low density (n e ≈ 1.0×1019 m-3). This resulted in improved electron confinement and Te0 reaching 10 keV. LHCD has been used for current profile broadening in the early phase of the optimised shear discharges. The use of LH profile control during the high performance phase still remains to be developed. ACKNOWLEDGMENTS The authors are grateful to the entire JET Team for support in carrying out the experiments presented here. REFERENCES [1] [2] [3] Final Report on the ITER Task IVA-LH: Definition of Heating and Current Drive Systems, “Lower Hybrid Heating and Current Drive in ITER, Operation Scenarios and Outline System Design”, 1994, Report EUR-CEA-FC-1529. Moreau D, et al, “Modelling of ITER Operation Scenarios”, 1997, NET Report EUR FU/ XII-128/110/97. Litaudon X, Arslanbekov R, Hoang G T, Joffrin E, Kazarian-Vibert F, Moreau D, Peysson Y, et al, Plasma Phys. Control. Fusion 38 (1996) 1603. 13 [4] [5] [6] [7] [8] [9] [10] [11] [12] [13] [14] [15] [16] [17] Ide S, Fujita T, Naito O and Seki M, Plasma Phys. Control. Fusion 38 (1996) 1645. Baranov Y F, Ekedahl A, Froissard P, Gormezano C, Lennholm M, Rimini F and Söldner F X, Nucl. Fusion 36 (1996) 1031. Jones T T C and the JET Team, Plasma Phys. Control. Fusion 37, Suppl 11A (1995) A359. Lomas P J and the JET Team, Proc. 16th IAEA Fusion Energy Conference, Montreal, Canada, 1996 (Vienna: International Atomic Energy Agency), IAEA-CN-64/A1-5. Thomas P and the JET Team, Proc. 16th IAEA Fusion Energy Conference, Montreal, Canada, 1996 (Vienna: International Atomic Energy Agency), IAEA-CN-64/A3-2. Gormezano C, Rimini F, Bhatnagar V, Cottrell G, Eriksson L G, et al, Proc. 12th Topical Conf. on Radio Frequency Power in Plasmas, Savannah, GA, USA, 1997 (New York: American Institute of Physics). Söldner F X and the JET Team, Invited paper at 24th EPS Conf. on Controlled Fusion and Plasma Physics, Berchtesgaden, Germany, 1997. To appear in Plasma Phys. Control. Fusion. Pain M, Brinkschulte H, Bosia G, Brusati M, Dobbing J A, et al, Proc. 13th Symp. on Fusion Engineering, Knoxville, Tennessee, USA, 1989 (New York: IEEE). Dobbing J A, Bosia G F, Brinkschulte H W, Brusati M, Ekedahl A, Gormezano C, et al, Proc. 17th Symp. on Fusion Technology, Rome, Italy, 1992 (Amsterdam: Elsevier Science), Vol. 1, p. 472. Ekedahl A, Dobbing J A, Finburg P, Fischer B, Gormezano C, et al, 1997, JET Report JETP(97)19. To appear in Plasma Devices and Operations. Dobbing J A, Ekedahl A, Finburg P, Fischer B, Gormezano C, et al, 1997, JET Report JETR(97)06. Lennholm M, Baranov Y, Dobbing J A, Ekedahl A, Finburg P, et al, Proc. 16th Symp. on Fusion Engineering, Urbana-Champaign, Illinois, USA, 1995 (New York: IEEE). Schild Ph, Dobbing J A, Ekedahl A, Goniche M, Gormezano C, Lennholm M and Söldner F X, Proc. 17th Symp. on Fusion Engineering, San Diego, CA, USA, 1997 (New York: IEEE). Ekedahl A, Baranov Y, Dobbing J A, Fischer B, Gormezano C, et al, Proc. 12th Topical Conf. on Radio Frequency Power in Plasmas, Savannah, GA, USA, 1997 (New York: American Institute of Physics). [18] Leuterer F, Söldner F, Brambilla M, Münich M, Monaco F, Zouhar M, et al, Plasma Phys. Control. Fusion 33 (1991) 169. [19] Kimura H, Naito O, Saigusa M, Ide S, Nemoto M, Ikeda Y, et al, Proc. 11th Topical Conf. on Radio Frequency Power in Plasmas, Palm Springs, CA, USA, 1995 (New York: American Institute of Physics), p. 81. [20] Romero J A, PhD Thesis, “Real Time Current Profile Control at JET”, UNID University, Madrid, Spain, 1997. Submitted for publication in Fusion Eng. Des. 14 [21] Forest C B, Kupfer K, Luce T C, Politzer P A, Lao L L, Wade M R, Whyte D G and Wròblewski, Phys. Rev. Lett. 73 (1994) 2444. [22] Shoucri M and Shkarofsky I, “A Fast 2-D Fokker-Planck Solver”, Centre Canadien de Fusion Magnetique, 1993, Report CCFM RI 411e. [23] Froissard P, Brusati M, Adams J M, Ekedahl A, Gormezano C, et al, Proc. 18th EPS Conf. on Controlled Fusion and Plasma Physics, Berlin, FRG, 1991 (Geneva: European Physical Society) Vol. 15C, Part III, p. 389. [24] Litaudon X and Moreau D, Nucl. Fusion 30 (1990) 471. [25] Ikeda Y, Naito O, Kondoh T, Seki M, Ide S, et al, Proc. 15th IAEA Conf. on Plasma Physics and Controlled Nuclear Fusion Research, Seville, Spain, 1994 (Vienna: International Atomic Energy Agency), IAEA-CN-60/A3-1. [26] Ikeda Y, Naito O, Ushigusa K, Sato M, Kondoh T, et al, Nucl. Fusion 34 (1994) 871. [27] Rimini F G, Alper B, Baranov Y, DeHaas J, Dobbing J A, Ekedahl A, et al, Proc. 11th Topical Conf. on Radio Frequency Power in Plasmas, Palm Springs, CA, USA, 1993 (New York: American Institute of Physics), p. 110. [28] Cenacchi G and Taroni A, “JETTO: A Free-Boundary Plasma Transport Code (Basic Version)”, 1988, Report ENEA RT/TIB 88(5). nd [29] Fischer B, Baranov Y, Ekedahl A, Gormezano C, Lennholm M, Parail V, et al, Proc. 22 EPS Conf. on Controlled Fusion and Plasma Physics, Bournemouth, UK, 1995 (Geneva: European Physical Society), Vol. 19C, Part III, p. 361. [30] Jacquinot and the JET Team, Proc. 16th IAEA Fusion Energy Conference, Montreal, Canada, 1996 (Vienna: International Atomic Energy Agency), IAEA-CN-64/O1-4. [31] Levinton F M, Zarnstorff M C, Bahta S H, Bell M, Bell R E, et al, Phys. Rev. Lett. 75 (1995) 4417. [32] Strait E J, Lao L L, Mauel M E, Rice B W, Taylor T S, et al, Phys. Rev. Lett. 75 (1995) 4421. [33] Koide Y and the JT-60 Team, Phys. Plasmas 4 (1997) 1623. [34] Hoang G T, Gil C, Joffrin E, Moreau D, et al, Nucl. Fusion 34 (1994) 75. 15