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Transcript
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extracts may not be published prior to publication of the original, without the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.
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15