fusion plasma physics and iter: an introduction · fusion plasma physics and iter: an introduction...

CERN Academic Training Programme, 12-13 April 2011

Fusion Plasma Physics and ITER: An Introduction

D J Campbell ITER Organization, Route de Vinon sur Verdon, 13115 St Paul lez Durance, France

Acknowledgements: Many colleagues in the ITER Organization and the international fusion programme

The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

2. Physics of Tokamak Plasmas


•  Controlling the plasma: −  heating −  current drive −  measurement −  control

•  Key physics challenges: −  confinement −  magnetohydrodynamic stability −  power and particle exhaust −  fusion power and the burning plasma −  Steady-state operation

Lecture 2 - Synopsis


Controlling the plasma


•  A magnetically confined plasma is a complex state −  it cannot be created and sustained without a sophisticated feedback control

system

•  Free energy which is available within the plasma tends to generate turbulence and magnetohydrodynamic instabilities (mhd) which reduce plasma confinement quality −  much effort is expended to control and sustain a high quality plasma

capable of producing significant fusion power

•  In a thermonuclear plasma with significant fusion power, additional control requirements are imposed in ensuring the fusion power can be sustained for extended periods −  failures of the control system lead to conditions which tend to reduce fusion

power or extinguish the plasma

Plasma Control


•  There are several “actuators” which can influence the plasma behaviour:

−  magnetic geometry: changing the plasma equilibrium shape

−  plasma heating

−  injecting current – changing the current profile

−  rotation – injecting torque

−  fuelling – injecting particles (even impurities)

•  Over the years we have learned how to use these tools to influence plasma behaviour −  “optimizing plasma performance” means largely finding the right

combination of control tools

How can we control the plasma?


The plasma shape can be very flexible

•  Plasma shape can affect both confinement and stability properties

R Stambaugh, APS (2000)


•  Tokamaks have a built in heating scheme: “ohmic” heating by the current −  but plasma resistivity varies as Te

-3/2, so heating power declines with increasing Te

−  so plasma temperatures of several keV are possible, but additional heating is required to achieve 10-20 keV

•  Two basic heating schemes: −  injecting neutral particle beams −  injecting radiofrequency waves – because the plasma refractive index

depends on density and magnetic fields, several RF options are possible

•  Each heating technique also provides some current drive

•  A further question: −  how much of each type do we need to optimize plasma performance?

Plasma Heating


•  Neutral beam injection (NBI): −  intense particle beams are accelerated,

neutralized and injected into plasma −  Eb ~ 100 keV, Pb up to 40MW −  very effective:

−  heating −  current drive −  fuelling −  rotation drive

•  For ITER: −  Eb ~ 1 MeV to penetrate plasma/ drive

current −  negative ion source technology −  higher energy ⇒ little fuelling, little rotation

drive

Injection of Neutral Particle Beams fuelling

heating

TFTR


•  Ion Cyclotron Radiofrequency Heating (ICRF): −  launched at frequencies ~ ωci ⇒ f ~ 50 MHz

−  technology conventional −  wave coupling to plasma problematic – penetration through edge

•  Electron cyclotron resonance heating (ECRH): −  launched at frequencies ~ ωce ⇒ f > 100 GHz

−  source technology non-conventional: “gyrotrons” −  coupling, absorption, space localization very good

•  Lower Hybrid Heating/ Current Drive (LHCD): −  “lower hybrid” a complex wave resonance in plasma: f ~ 5 GHz −  technology fairly conventional (source: klystrons) −  wave coupling to plasma problematic – penetration through edge

Radiofrequency Heating


•  Current drive provides: −  replacement of the transformer drive ⇒ towards steady-state plasma

−  manipulation of the current profile to improve confinement/ stability −  direct suppression of plasma instabilities

Current Drive

•  Current drive efficiency (driven current/ input power - ηCD): −  typically increases with Te

−  for beams, also increases with Eb

⇒ favourable for ITER

C Gormezano et al, Nucl Fusion 47 S285 (2007)


Paux for Q=10 nominal scenario: 40-50MW

NBI Layout

DNB

120GHz

Heating the ITER Plasma

126 or 170GHz


•  Technology: −  ICRF and LHCD fairly conventional −  NBI and ECRH source technology challenging

•  Coupling to plasma: −  NBI and ECRH straightforward

−  ICRF and LHCD problematic: antenna design challenging due to difficulty in coupling wave through (evanescent) plasma edge

•  Radial localization: −  Resonance condition favours ECRH and ICRF radial localization −  NBI and LHCD more global in effects

•  Current drive: −  NBI and LHCD most efficient −  ECRH and ICRF used in more specialized applications where space

localization important

Why 4 Heating Systems?


1 – 10 eV 10 – 100 eV 1 – 20 keV

Measurement: “Seeing” a Burning Plasma


•  About 40 large scale diagnostic systems are foreseen: •  Diagnostics required for protection, control and physics studies •  Measurements from DC to γ-rays, neutrons, α-particles, plasma species •  Diagnostic Neutral Beam for active spectroscopy (CXRS, MSE ….)

UPPER PORT (12 used)

EQUATORIAL PORT (6 used)

DIVERTOR PORT (6 used)

DIVERTOR CASSETTES (16 used)

VESSEL WALL (Distributed Systems)

Analyzing the Plasma - ITER Diagnostics


•  Puffing a small amount of tritium into the edge of a deuterium plasma allows hydrogenic particle transport to be investigated: •  The 14MeV neutron signal provides a distinct marker for the location of the

tritium as it diffuses into the plasma

14MeV neutron signal Derived tritium profile

P C Efthimion et al, Phys Rev Lett 75 85 (1995)

Example: Particle Transport Studies

TFTR


•  Sophisticated imaging systems in various spectral regions allow detailed internal images of plasma to be constructed

•  X-ray tomography allows internal magnetic structure of the plasma to be investigated: •  figure shows an image of a

“magnetic island” formation inside the plasma

G Huijsmans, JET

Example: Magnetic Island Structures


•  Controlling a burning plasma in ITER will require an extensive set of control functions based on measurement systems and actuators described here:

−  Plasma equilibrium control: (routine and robust) •  plasma shape, position and current

−  Basic plasma parameter control: (routine and robust) •  plasma density

−  Plasma kinetic control: (ongoing research) •  fuel mixture, fusion power, radiated power …

−  Control for advanced operation: (ongoing research) •  current profile

−  Control of magnetohydrodynamic instabilities: (ongoing research) •  several key MHD instabilities

•  All of these elements have been developed and demonstrated to varying extents in existing tokamak devices

Plasma Control in ITER


•  Plasma equilibrium control is routine: •  based on magnetic sensors which

provide signals to reconstruct plasma boundary or equilibrium

•  feedback signals control voltages to poloidal field coils to maintain required plasma equilibrium parameters

•  very long pulses require particular care to avoid drifts in magnetic diagnostic signals

Plasma Equilibrium Control

-5

-4

-3

-2

-1

0

1

2

3

4

5

3 4 5 6 7 8 9

Z, m

R, m

0.4MA

1.5MA

2.5MA

3.5MA4.5MA

5.5MA

SOH,SOB

XPF

11.5MA

6.5MA

ITER simulation


(M R Wade et al, Fusion Energy Conf 2006)

Control of MHD Instabilities

•  An MHD instability is detected magnetically: −  localized electron cyclotron

current drive is used to suppress the instability

DIII-D


Key Physics Challenges


•  A well developed “neoclassical” theory of transport processes in toroidal plasmas has been developed: −  unfortunately, it doesn’t do a good job

of describing plasma processes ⇒ turbulence normally dominates

−  turbulence produces “radial streamers” which represent regions of rapid heat transport

−  rotational shear in the plasma can break up turbulent structures, reducing transport

Plasma Confinement

(P Beyer et al, Uni Provence/ CEA, 1999)


Plasma Confinement: H-mode

JET

•  It is found that the plasma confinement state (τE) can bifurcate: −  two distinct plasma regimes, a low confinement (L-mode) and a high

confinement (H-mode), result −  this phenomenon has been shown to arise from changes in the plasma

flow in a narrow edge region, or pedestal


τth ∝ IpR2P-2/3

•  Turbulent transport is difficult to predict quantitatively: −  so, we use scaling experiments to predict the level of energy

confinement in ITER −  H-mode turns out to be robust enough to provide the basis for the ITER

design ⇒ significant reduction in size of device

Plasma Confinement: τE Scaling

Plasma Current Major Radius

Input Power


• Predictions of fusion performance in ITER rely essentially on a small number of physics rules: •  H-mode energy confinement scaling (IPB98(y,2)):

•  H-mode threshold power:

(ie a certain level of power needs to flow across the plasma boundary to trigger an H-mode)

!

"E,th98(y,2) = 0.144 I0.93B0.15P#0.69n0.41M0.19R1.97$0.58%0.78 (s)

!

"E # IR2P$2 / 3

!

PLH = 0.098M"1B0.80n 200.72S0.94 (MW)

!

H98(y,2) = "E,thexp / "E,th

98(y,2)NB:

ITER Physics Basis I


•  MHD stability:

•  Divertor physics:

!

q95 = 3

!

n/ nGW " 1

!

nGW (1020 ) =I(MA)"a2

!

"N = "(%) aBI(MA)

!

"N # 2.5

!

", # det ermined by control considerations

!

q95 = 2.5 a2BRI

f(",#,$)

!

Peak t arget power ~ 10MWm"2

!

Helium transport : "He* / "E ~ 5

!

Impurity content : nBe / ne = 0.02 (+ ~ 0.1% Ar for radiation)

ITER Physics Basis II


•  The interaction of the plasma fluid and the magnetic field is described by magnetohydrodynamic (MHD) stability theory

−  provides a good qualitative, and to a significant extent quantitative, description of stability limits and the associated instabilities

•  There are two basic types of instability: −  “ideal” instabilities produce field line bending – can grow very rapidly −  “resistive” instabilities cause tearing and reconnection of the magnetic

field lines ⇒ formation of “magnetic islands”

MHD Stability - Plasma Operational Limits

•  Plasma control techniques are being applied to suppress the most significant instabilities −  allows access to higher fusion performance


MHD Stability: Disruptions

H-mode L-

mode CQ TQ

Plasma current Plasma energy

RE current

t Typical chain of events during

plasma disruption

•  The ultimate stability limit in tokamak plasmas is set by major disruptions: large scale MHD instabilities −  loss of plasma energy in milliseconds (thermal quench – TC) −  Plasma current decays in 10s of milliseconds (current quench – QC)

•  Produces: −  very large heat loads on plasma facing surfaces −  significant electromagnetic forces in vacuum vessel −  large runaway electron beam

Mitigation techniques

essential


•  li-qa diagram describes stable plasma operating region, limited by disruptions: −  low li typically has to be negotiated during the plasma current ramp-up −  high-li limit typically occurs due to excessive radiation at plasma edge,

resulting in cold edge plasma and narrow current channel (eg at density limit)

qa=2 limit

JET Limiter plasmas

MHD Stability - Plasma Equilibrium Limits


•  Experiments have shown that tokamak plasmas can sustain a maximum density: −  limit depends on operating regime

(ohmic, L-mode, H-mode …)

−  limit may be determined by edge radiation imbalance or edge transport processes

−  limit can be disruptive or non-disruptive

•  Comprehensive theoretical understanding still limited −  “Greenwald” density:

nGW = I(MA)/ πa2

−  operational figure of merit

JET

MHD Stability - Density Limits


• Maximum value of normalized plasma pressure, β, is limited by mhd instabilities:

• Typically, “Troyon” limit describes tokamak plasmas:

βN ≤ 2.8-3.5

• More generally, “no-wall” limit: βN ≤ 4×li

!

"(%) = 100 < p >

B2 / 2µo

!

"N ="(%)

Ip(MA) / aB

Plasma MHD Stability – Pressure Limit: β


•  Essential problem is: −  handle power produced by plasma with

(steady-state) engineering limit for plasma facing surfaces of 10 MWm-2

−  extract helium from the core plasma to limit concentrated below ~6%

−  prevent impurities from walls penetrating into plasma core

−  ensure plasma facing surfaces last sufficiently long

Power and Particle Exhaust

Core plasma

Scrape-off layer (SOL) plasma: region of open field lines

Divertor targets

Private plasma

X-point


•  The divertor is a significant element of the solution −  surfaces for high heat fluxes (10 MWm-2) −  cryopumping to extract particles leaving the plasma, including helium


ITER divertor cassette – 54 cassettes make up the complete toroidal ring


•  The concept of divertor “detachment” is fundamental to power exhaust in the burning plasma environment: −  additional impurities are added to

edge plasma to increase radiation −  a large pressure gradient develops

along the field lines into the divertor −  the divertor plasma temperature

falls to a few eV −  a large fraction of the plasma

exhaust power is redistributed by radiation and ion-neutral collisions

0 1 2 3 4 5MW/m3

Deut.

Carbon

Radiationfrom

DIII-DRadiationLevelsDetachedRegion

0

1

2

3

4

1.0 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8

Radius (m)

IRTV Divertor Heat Flux

Before Gas Puffing

After Detachment

Hea

t Flu

x (M

W/m

2 )



•  CFC divertor targets (~50m2): −  high thermal conductivity and good

thermal shock resistance (doesn’t melt)

−  but combines chemically with hydrogen (ie tritium)

•  Be first wall (~700m2): −  good thermal conductivity −  low-Zi – low core radiation

•  W-clad divertor elements (~100m2): −  high melting point and sputtering

resistance −  but might still melt during thermal

transients −  will eventually replace CFC in entire

divertor

Plasma Facing Components - Challenges


•  Access to plasmas which are dominated by a-particle heating will open up new areas of fusion physics research, in particular:

−  confinement of α’s in plasma −  response of plasma to α-heating −  influence of α-particles on MHD stability

•  Experiments in existing tokamaks have already provided some positive evidence

−  ”energetic” particles (including α-particles) are well confined in the plasma

−  such particle populations interact with the background plasma and transfer their energy as predicted by theory

−  but energetic particles can induce MHD instabilities (Alfvén eignemodes) - for ITER parameters at Q=10, the impact is expected to be tolerable

Burning Plasma Physics in ITER


•  In existing experiments single particle theory of energetic ion confinement confirmed: −  simple estimate, based on banana

orbit width shows that Ip ≥ 3MA required for α-particle confinement

•  Classical slowing down of fast ions well validated: −  data range 30keV NBI (ISX-B) to

3.5MeV α-particles (TFTR)

•  Energetic ion heating processes routinely observed in additional heating experiments

.

2.0

1.5

1.0

0.5

0.010-2 10-1 1 10

SLOWING-DOWN TIME (s)

EXPE

RIM

ENT

/ THE

ORY

****

ITER reference value

W W Heidbrink, G J Sadler, Nucl Fusion 34 535 (1994)

Energetic Ion Confinement


•  Electron heating by α-particles validated in JET and TFTR DT experiments: −  in JET, maximum electron

temperature correlated with optimum fuel mixture for fusion power production

P R Thomas et al, Phys Rev Lett 80 5548 (1998)

Electron Heating by α-Particles


•  In a tokamak plasma, the Alfvén wave continuum splits into a series of bands, with the gaps associated with various features of the equilibrium: •  a series of discrete frequency Alfvén eigenmodes can exist in these gaps:

•  toroidicity-induced (TAE) gap created by toroidicity •  ellipticity-induced (EAE) gap created by elongation •  triangularity-induced (NAE) gap created by additional non-

circular effects

•  beta-induced (BAE) gap created by field compressibility •  kinetic toroidal (KTAE) gap created by non-ideal effects

such as finite Larmor radius … and others!

•  These modes can be driven unstable by the free energy arising from energetic particle populations with velocities above the Alfvén velocity, eg α-particles

Alfvén Eigenmodes


A Q=10 scenario with: Ip=15MA, Paux=40MW, H98(y,2)=1 Te

Ti

ne

10nHe

Zeff

fHe

qΨ

χe

keV

%

MA

/m2

1019m

-3 m

2s-1

Current Ramp-up Phase

-5

-4

-3

-2

-1

0

1

2

3

4

5

3 4 5 6 7 8 9

Z, m

R, m

Plasma start-up in PF Scenario 2

4.5MA (XPF)

0.5MA - XPF

0.5MA

1.5MA

2.5MA

3.5MA

-5

-4

-3

-2

-1

0

1

2

3

4

5

3 4 5 6 7 8 9

Z, m

R, m

Plasma start-up in PF Scenario 2

4.5MA (XPF)

XPF - SOF15MA (SOF)

Dashed line: reference plasma

ITER Plasma Scenario - H-mode


•  Discovery of internal transport barriers ⇒ “advanced scenarios”

•  But development of an integrated plasma scenario satisfying all reactor-relevant requirements remains challenging

plasma with reversed central shear + sufficient rotational shear

internal transport barrier ⇒ enhanced confinement

reduced current operation + large bootstrap current fraction

reduced external current drive + current well aligned for mhd stability and confinement enhancement

active mhd control

Steady-state operation + High fusion power density

Steady-State Operation


•  Calculations of plasma profiles for a possible ITER steady-state scenario with Q=5

0

5

10

15

20

25

30

0.0 0.2 0.4 0.6 0.8 1.0

T, keV

X

1

0

ne, 1020m-3

Te

Ti ne

0.0

0.2

0.4

0.6

0.8

1.0

0.0 0.2 0.4 0.6 0.8 1.0

j, MA/m2

X

10

0

q

jbs

jtot

jnb+lh

q

(A Polevoi et al, IAEA-CN-94/CT/P-08, IAEA FEC2002)

ITER Steady-State Operation


•  To fulfill its missions ITER must carry out an ambitious and exciting physics programme

•  Its essential design features give it the capability to do this: −  pulse length and duty cycle −  flexible heating and current drive systems

•  total power •  variety of systems

−  diagnostic access and facilities −  additional plasma engineering systems

•  flexible fuelling systems •  capability for control of important MHD instabilities

−  equilibrium shape flexibility −  divertor and first wall exchange capability

ITER has many assets as a burning plasma experiment and the key step towards the realization of fusion energy

Conclusions


ITER Physics Basis, ITER Physics Expert Groups et al, Nucl Fusion 39 2137-2638 (1999)

Progress in the ITER Physics Basis, ITPA Topical Physics Groups et al, Nucl Fusion 47 S1-S413 (2007)

Hawryluk, R. J. Results from deuterium-tritium tokamak confinement experiments, Reviews of Modern Physics 70, 537 (1998).

Gibson, A. et al. Deuterium-tritium plasmas in the Joint European Torus (JET): Behavior and implications, Physics of Plasmas 5, 1839 (1998).

Keilhacker, M. et al. High fusion performance from deuterium-tritium experiments in JET, Nuclear Fusion 39, 209 (1999).

Jacquinot, J. et al. Overview of ITER physics deuterium-tritium experiments in JET, Nuclear Fusion 39, 235 (1999).

http://www.iter.org - and associated links

References: Tokamak Fusion Physics

fusion plasma physics and iter: an introduction · fusion plasma physics and iter: an introduction...

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