Kein Folientitel

1
Standard and Advanced Tokamak Operation Scenarios
for ITER
Hartmut Zohm Max-Planck-Institut für
Plasmaphysik, Garching, Germany EURATOM
Association

• What is a tokamak (operational) scenario?
• Short recap of fusion and tokamak physics
• Conventional scenarios
• Advanced scenarios
• Summary and conclusions

Lecture given at PhD Network Advanced Course,
Garching, 07.10.2008
2
What is a tokamak scenario?

• A tokamak (operational) scenario is a recipe to
  run a tokamak discharge
• Plasma discharge characterised by
• external control parameters Bt, R0, a, k, d,
  Pheat, FD
• integral plasma parameters b 2m0ltpgt/B2, Ip
  2p ? j(r) r dr
• plasma profiles pressure p(r) n(r)T(r),
  current density j(r)

current density (a.u.)
current density (a.u.)
total j(r) noninductive j(r)
total j(r) noninductive j(r)
3
Control of the profiles j(r)and p(r) is limited
safety factor
ITER simulation (CEA Cadarache)

• ohmic current coupled to temperature profile via
  s T3/2
• ? inductive current profiles always peaked,
  q-profiles monotonic
• external heating systems drive current, but with
  limited efficiency
• (typically less than 0.1 A per 1 W under
  relevant conditions)
• pressure gradient drives toroidal bootstrap
  current jbs (r/R)1/2 ?p/Bpol

4
Control of the profiles j(r)and p(r) is limited

• Pressure profile determined by combination of
  heating / fuelling
• profile and radial transport coefficients
• ohmic heating coupled to temperature profile via
  s T3/2
• external heating methods allow for some
  variation ICRH/ECRH
• deposition determined by B-field, NBI has
  usually broad profile
• gas puff is peripheral source of particles,
  pellets further inside
• but under reactor-like conditions, dominant
  a-heating (nT)2

5
Control of the profiles j(r)and p(r) is limited

• Pressure profile determined by combination of
  heating / fuelling
• profile and radial transport coefficients
• anomalous (turbulent) heat transport leads to
  stiff temperature
• profiles (critical gradient length ?T/T)
  except at the edge
• density profiles not stiff due to existence of
  pinch

6
Control of the profiles j(r)and p(r) is limited

• Stiffness can be overcome locally by sheared
  rotation
• edge transport barrier (H-mode)
• internal transport barrier (ITB)

7

• What is a tokamak (operational) scenario?
• Short recap of fusion and tokamak physics
• Conventional scenarios
• Advanced scenarios
• Summary and conclusions

8
Figure of merit for fusion performance nTtE

• Aim is to generate power, so
• Pfusion/Pheat (power needed to sustain
• plasma) should be high
• Pheat determined by thermal insulation
• tE Wplasma/Pheat (energy confinement time)
• In present day experiments, Pheat comes
• from external heating systems
• Q Pfus/Pext ? Pfus/Pheat nTtE
• In a reactor, Pheat mainly by a-(self)heating
• Q Pfus/Pext ? ? (ignited plasma)
• The aim is to generate and sustain a plasma of T
  several 10 keV,

9
Optimisation of nTtE ideal pressure limit

• Optimising nT means high pressure and, for given
  magnetic field,
• high b 2m0 ltpgt / B2
• This quantity is limited by magneto-hydrodynamic
  (MHD) instabilities
• Ideal MHD limit (ultimate limit, plasma
• unstable on Alfvén time scale 10 ms,
• only limited by inertia)
• Troyon limit bmax Ip/(aB), leads to
• definition of bN b/(Ip/(aB))
• at fixed aB, shaping of plasma cross-
• section allows higher Ip (low q limit,
• see later)? higher b

10
Optimisation of nTtE resistive pressure limit

• Optimising nT means high pressure and, for given
  magnetic field,
• high b 2m0 ltpgt / B2
• This quantity is limited by magneto-hydrodynamic
  (MHD) instabilities
• Resistive MHD limit (on local current
• redistribution time scale 100 ms)

ASDEX Upgrade
11
Optimisation of nTtE resistive pressure limit

• Optimising nT means high pressure and, for given
  magnetic field,
• high b 2m0 ltpgt / B2
• This quantity is limited by magneto-hydrodynamic
  (MHD) instabilities
• Resistive MHD limit (on local current
• redistribution time scale 100 ms)
• Neoclassical Tearing Mode (NTM)
• driven by loss of bootstrap current
• within magnetic island

12
Optimisation of nTtE density limit
qedge 2
1/qedge (normalised current)

• Since T has an optimum value at 20 keV, n
  should be as high as possible
• density is limited by disruptions due to
  excessive edge cooling
• empirical Greenwald limit, nGr Ip/(pa2) ?
  high Ip helps to obtain high n

13
Optimisation of nTtE current limit

• BUT for given Bt, Ip is limited by current
  gradient driven MHD instabilities
• Limit to safety factor q (r/R) (Btor/Bpol)
• for q lt 1, tokamak unconditionally unstable ?
  central sawtooth instability
• for qedge ? 2, plasma tends to disrupt (external
  kink) limits value of Ip

qedge 2
1/qedge (normalised current)
Hugill diagram for TEXTOR
JET
normalised density
14
Optimisation of nTtE confinement scaling

• Empirical confinement scalings show linear
  increase of tE with Ip
• note the power degradation (tE decreases with
  Pheat!)
• H-factor H measures the quality of confinement
  relative to the scaling

Empirical ITER 98(p,y) scaling tE H Ip0.93
Pheat-0.63 Bt0.15
15
Tokamak optimisation steady state operation

• N.B for steady state tokamak operation, high Ip
  is not desirable
• Tokamak operation without transformer current
  100 noninductive
• external CD has low efficiency (remember less
  than 0.1 A per W)
• internal bootstrap current high for high jbs
  (r/R)1/2 ?p/Bpol
• ? fNI Ibs/Ip p/Bpol2 bpol
• Advanced scenarios, which aim at steady state,
  need high b, low Ip,
• have to make up for loss in tE by increasing H
• Without the steady state boundary condition, a
  tokamak scenario
• is called conventional
• N.B.2 Long Pulse stationary on all intrinsic
  time scales

16

• What is a tokamak (operational) scenario?
• Short recap of fusion and tokamak physics
• Conventional scenarios
• Advanced scenarios
• Summary and conclusions

17
The (low confinement) L-mode scenario

• Standard scenario without special tailoring of
  geometry or profiles
• central current density usually limited by
  sawteeth
• temperature gradient sits at critical value over
  most of profile
• extrapolates to very large (R gt 10 m, Ip gt 30
  MA) pulsed reactor

18
The (high confinement) H-mode scenario

• With hot (low collisionality) conditions, edge
  transport barrier develops
• gives higher boundary condition for stiff
  temperature profiles
• global confinement tE roughly factor 2 better
  than L-mode
• extrapolates to more attractive (R 8 m, Ip
  20 MA) pulsed reactor

19
Hot (low collisionality) edge by divertor
operation

• plasma wall interaction in well defined zone
  further away from core plasma
• possibility to decrease T, increase n along
  field lines (pconst.)
• high edge temperature gives access to edge
  transport barrier, if
• enough heating power is supplied (power
  threshold for H-mode)

20
Mechanism for edge transport barrier formation

• in a very narrow (1 cm) layer at the edge very
  high plasma
• rotation develops (E v x B several 10s of
  kV/m)
• sheared edge rotation tears turbulent eddies
  apart
• smaller eddy size leads to lower radial
  transport (D dr2/tdecor)

21
Stationary H-modes usually accompanied by ELMs

• Edge Localised Modes (ELMs) regulate edge plasma
  pressure
• without ELMs, particle confinement too good
  impurity accumulation

22
Stationary H-modes usually accompanied by ELMs
acceptable lifetime for 1st ITER divertor

• But ELMs may pose a serious threat to the ITER
  divertor
• large type I ELMs may lead to too high
  divertor erosion

23
b-limit in H-modes usually set by NTMs

• Present day tokamaks limited by (3,2) or (2,1)
  NTMs (latter disruptive)
• while onset bN is acceptable in present day
  devices, it may be quite
• low in ITER due to unfavourable rp scaling
  (rather than machine size)

24
H-mode is ITER standard scenario for Q10
Pressure driven MHD instabilities
Density limit (Greenwald)
Access to edge transport barrier
H tE,ITER/tE,predicted

• The design point allows for
• achieving Q10 with conservative assumptions
• incorporation of moderate surprises
• achieving ignition (Q ? ?) if surprises are
  positive

25
but some open issues remain

• Need to minimise ELM impact on divertor
• reduce power flow to divertor by radiative edge
  cooling
• special variants of the scenario (Quiescent
  H-mode, type II ELMs)
• ELM mitigation pellet pacing or Resonant
  Magnetic Perturbations
• Need to tackle NTM problem
• NTM suppression by Electron Cyclotron Current
  Drive demonstrated,
• but have to demonstrate that this can be used
  as reliable tool

26
ELM control by pellet pacing
ASDEX Upgrade

• Injection of pellets triggers ELMs allows to
  increase ELM frequency
• at the same time, ELM size decreases and peak
  power loads are mitigated

27
ELM control by Resonant Magnetic Perturbations
DIII-D

• Static error field resonant in plasma edge can
  suppress ELMs
• removes ELM peaks but keeps discharge stationary
  with good confinement
• very promising, but physics needs to be
  understood to extrapolate to ITER

28
NTM control by Electron Cyclotron Current Drive
Active control is possible by generating a
localised helical current in the island to
replace the missing bootstrap current
29
Suppression of (2,1) NTM by ECCD
ASDEX Upgrade
30
Suppression of (2,1) NTM by ECCD
ASDEX Upgrade
Current drive (PECRH / Ptotal 10-20 ) results
in removal method has the potential for
reactor applications
31

• What is a tokamak (operational) scenario?
• Short recap of fusion and tokamak physics
• Conventional scenarios
• Advanced scenarios
• Summary and conclusions

32
Advanced tokamak the problem of steady state

• Advanced scenarios aim at stationary
  (transformerless) operation
• external CD has low efficiency (remember less
  than 0.1 A per W)
• internal bootstrap current high for high jbs
  (r/R)1/2 ?p/Bpol
• ? fNI Ibs/Ip p/Bpol2 bpol
• Recipe to obtain high bootstrap fraction
• low Bpol, i.e. high q elevate or reverse
  q-profile (q(r/R)(Btor/Bpol))
• eliminates NTMs (reversed shear, no low resonant
  q-surfaces)
• high pressure where Bpol is low, i.e. peaked
  p(r)
• Both recipes tend to make discharge ideal MHD
  (kink) unstable!

33
Advanced tokamak the problem of steady state
ASDEX Upgrade
conventional
j(r)
q(r)
p(r)
jbs(r)

• A self-consistent solution is theoretically
  possible
• reversing q-profile suppresses turbulence
  internal transport barrier (ITB)
• large bootstrap current at mid-radius supports
  reversed q-profile

34
Problems of the Advanced tokamak scenario

• Broad current profile leads to low kink stability
  (low b-limit)
• can partly be cured by close conducting shell,
  but kink instability
• then grows on resistive time scale of wall
  (Resistive Wall Mode RWM)
• can be counteracted by helical coils, but this
  needs sophisticated feedback
• Position of ITB and minimum of q-profile must be
  well aligned
• needs active control of both p(r) and j(r)
  profiles difficult with limited
• actuator set (and cross-coupling between the
  profiles)

35
RWM control by Resonant Magnetic Perturbations

• Feedback control using RMPs shows possibility to
  exceed no-wall b-limit
• rotation plays a strong role in this process and
  has to be understood
• better (ITER is predicted to have very low
  rotation)

36
Advanced Tokamak Stability is a tough Problem
Hybrid scenarios
Reversed shear scenarios
QDB scenarios
0
.
8
0
.
8
0
.
6
0
.
6
2
95
/q
N
ITER reference (Q10)
b
0
.
4
0
.
4
89
H
ITER advanced (Q5)
0
.
2
0
.
2
0
.
0
0
.
0
0
20
40
60
80
t
t
d
ur
ation
/
d
ur
ation
/
E
E
Good performance can only be kept for several
confinement times, not stationary on the current
diffusion time (10 50 tE in these devices)
37
Control of ITB dynamics is nontrivial
CRONOS Code (CEA)
b
Example modelling of steady-state scenario in
ITER delicate internal dynamics that may be
difficult to control with present actuator set
38
A compromise the hybrid scenario

• Reversed shear, ITB discharges
• very large bootstrap fraction
• steady state should be possible
• low b-limit (kink, infernal, RWM)
• delicate to operate
• Zero shear, hybrid discharges
• higher b-limit (NTMs)
• easy to operate
• smaller bootstrap fraction
• have to elevate q(0)

Hybrid operation aims at flat, elevated q-profile
discharges with high q(0) Not clear if this
projects to steady state, but it will be very
long pulse
39
A hybrid scenario the improved H-mode
0.7
q-range
H89bN/q952
3-4
4-5
5-6
6-7
0.5
ITER (shape corrected)
0.3
0.1
?(a/R)bp
0.1
0.3
0.5
0.7
0.9
1.1
Bootstrap fraction

• Improved H-mode (discovered on AUG) is best
  candidate for hybrid operation
• projects to either longer pulses and/or higher
  fusion power in ITER
• flat central q-profile, avoid sawteeth need
  some current profile control

40
A hybrid scenario the improved H-mode
Presently, hybrid scenarios perform better in
fusion power and pulse length
41

• What is a tokamak (operational) scenario?
• Short recap of fusion and tokamak physics
• Conventional scenarios
• Advanced scenarios
• Summary and conclusions

42
Summary and Conclusions
Advanced operation
Safety factor q
Hybrid
L-mode H-mode

• A variety of tokamak operational scenarios exists
• L-mode low performance, pulsed operation, no
  need for profile control
• H-mode higher performance, pulsed operation,
  MHD control needed
• Advanced modes higher performance, steady
  state, needs profile control

43
Summary and Conclusions

• ITER aims at operation in conventional and
  advanced scenarios
• demonstrating Q10 in conventional
  (conservative) operation scenarios
• demonstrating long pulse (steady state)
  operation in advanced scenarios
• One mission of ITER and the accompanying
  programme is to develop and
• verify an operational scenario for DEMO
• DEMO scenario must be a point design (no longer
  an experiment)