OVERVIEW OF ITER PHYSICS

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OVERVIEW OF ITER PHYSICS

V. Mukhovatov1, M. Shimada1, A.E. Costley1, Y. Gribov1,

G. Federici2,A.S. Kukushkin2, A. Polevoi1, V.D. Pustovitov3,

Y. Shimomura1, T. Sugie1, M. Sugihara1, G. Vayakis1

1 International Team, ITER Naka Joint Work Site, Naka, Ibaraki, Japan2 International Team, ITER Garching Joint Work Site, Garching, Germany

3 Nuclear Fusion Institute, RRC Kurchatov Institute, Moscow, Russia

ITER

V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia

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Contents

V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russi

Introduction ELMy H-mode

Operational limits Confinement Instabilities

Improved H-mode Internal Transport Barriers

Formation Performance Control

Summary

3

Introduction


Predictive methodologies for tokamak Burning Plasma Experiment (BPX) have been summarized in the ITER Physics Basis (IPB) published in 1999 [Nucl. Fusion 39 (1999) 2137-2638].

In recent years, significant progress has been achieved in many areas of tokamak physics

New achievements have had significant impact on new ITER design (stronger shaping, methods to suppress NTMs and RWMs)

This talk reviews the ITER physics basis taking account of the recent progress in tokamak studies

4

Major ITER-Relevant Confinement Modes


H-mode (High Confinement Mode) associated with formation of edge transport barrier (ETB)

Reference mode for ITER inductive high-Q operation

Improved H-mode Candidate mode for

inductive and/or hybrid ITER operation

Advanced Tokamak (AT) mode associated with formation of Internal Transport Barrier (ITB)

Candidate mode for steady-state ITER operation

5

Physics Rules for Selection of ITER Design Parameters


Q ≥ 10 Q = 5P/Paux

ELMy H-mode reference operation mode

ITERH-98P(y,2) scaling for energy confinement time

Safety factor q95 ≥ 2.5 q95 (5B/I)(a2/R)

Electron density ne ≤ nG nG= I/(a2), Greenwald density

Normalized beta N ≤ 2.5 [N = (%)(aB/I)]

Strong plasma shaping sep = 1.85, sep = 0.48

Heating power P ≥ 1.3 PL-H P = P + Paux- Prad

PL-H is H-mode power thresh.

6

ELMy H-MODE

7

ELMy H-mode


ELMy H-mode: H-mode with bursts of Edge Localized Modes (ELMs)

Reference ITER mode for inductive high-Q operation Robust mode observed in all tokamaks under wide

variety of conditions at heating power above the threshold, P>PL-H

Good prospects for long-pulse operation >20 years of studies Rich experimental database High confidence that ELMy H-mode will be obtained

in ITER

8

Energy Confinement Projections for ELMy H-mode in ITER


Three approaches (discussed in details in IPB) predict compatible results for ITER reference high Q scenario

Transport models based on empirical scalings for the energy confinement time

Physics-based transport models

Dimensionless analysis

9

ITER Reference Scalings


ITERH-98P(y.2) confinement scaling ITER:E = 3.66s ±14% [2.78, 4.83]s

95% nonlinear interval estimate

O.Kardaun, Nucl. Fusion 42 (2002) 841J A Snipes et al PPCF 42 (2000) A299

H-mode power threshold scaling ITER: PL-H= 49 MW [28.4, 84.1]MW

95% interval estimate

10

Effect of Plasma Dilution with Helium ITER performance depends on

plasma dilution with He

B2/Eirene code:

Helium content in ITER plasma reduces due to Helium atom elastic collisions with D/T ions

Reduction of He content improves ITER performance


1/2D ITINT1.SAS code with Psep ≥ PL-H

O.J.W.F. Kardaun NF 42 (2002) 841

11

Theory Based Transport Models


WEILAND, MMM, GLF23 and IFS/PPPL transport models

Transport driven by drift wave turbulence

Detailed treatment is somewhat different

Boundary conditions taken from experiments or from empirical or semi-empirical scalings

Reasonable agreement with experimental data for plasma core

12

ITER Predictions by Physics Based Models


Pedestal scalings

(a) J G Cordey, et al 19th FEC Lyon (b) J G Cordey, et al19th FEC, Lyon

(c) M Sugihara, et alNF 40 (2002) 1743

(d) A H Kritz, et al29th EPS D-5.001

(e) M Sugihara, et al submitted to PPCF

(g) K S ShaingT H Osborne et al19th FEC, Lyon

13

ITER Predictions by Physics Based Models


Predictions for ITER by different models at the same input parameters (G. Pereverzev et al. 29th EPS 2002 P-1072)

14

Edge Pedestal in ELMy H-mode


J.G.Cordey et al IAEA Lyon Conf. 2002 M Sugihara et al , submitted tp PPCF 2003

Two-term confinement scalings for thermal energy

W = Wcore+ Wped

Edge temperature gradient limited by thermal conduction

ITER: Wped = 174 MW

Tped = 5.2 keV

Edge gradient limited by ELMs (MHD limit):

ITER: Wped= 98 MW

Tped ≈ 3.0 keV

15

Non-Dimensional Confinement Scalings


GyroBohm like scalings have been found in experiments with ELMy H-mode:

BE (*)-3.15 0.03 (*)-0.42 in DIII-D

BE (*)-2.7 -0.05 ( *)-0.27 in JET (*= i/a)

JET DT discharge with all dimensionless parameters, *, q, R/a, etc, except *, the same as ITER:

JET #42983: *= 4.25 10-3

JET-like ITER: *= 1.88 10-3

==> Q = 6 - 13

16

High Performance H-Modes at High Density Demonstrated

One of the major achievements in recent tokamak experiments was demonstration of good confinement in H-mode at high plasma density required for ITER, i.e.

H98(y,2)= 1 at n ≥ 0.85 nG

There are several ways to improve confinement at high density

Increase in plasma triangularity; gentle gas fuff Impurity seeding High field side pellet fueling


17V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia

Good Confinement at High DensityEnergy confinement reduces with density but improves with plasma triangularity or shaping parameter q95/qcyl

H(y,2)corr = 0.46 + 1.35 ln(q95/qcyl) - 0.17 n/nG + 0.38(n/nped -1)

ITER: H(y,2)corr=0.91 at n/nped=1; H(y,2)corr=1.05 at n/nped=1.3

HH

JET

ITER

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Power and Particle Control in ITER


B2/Eirene code: steady state divertor power loads are within the proven limits

He density at the separatrix reduces by 3-5 times due to elastic collisions of He atoms with D/T ions

A S Kukushkin, H D Pacher PPCF 44 (2002) 943

19

Major Instabilities in ELMy H-mode


Sawteeth

Edge localized modes (ELMs)

Neoclassical tearing modes (NTMs)

Alfven instabilities

Disruptions

20

EDGE LOCALIZED MODES (ELMs)

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H-mode Regimes with Smaller ELMs Expected energy fluxes on the ITER divertor associated

with ELMs are close to being marginal for an acceptable divertor target life time

There are alternative high confinement modes with small ELMs found at q95 > 3.6-4 and high triangularity

H-mode with ‘grassy’ or ‘minute’ ELMs in DIII-D and JT-60U Enhanced D (EDA) mode in Alcator C-Mod with quasi-

coherent density fluctuations Advanced H-mode with Type II ELMs in ASDEX-U Impurity seeded H-mode in JET with reduced Type I ELMs High density H-mode with rear small ELMs in JET Quiescent Double Barrier (QDB) H-mode in DIII-D

ELM mitigation with frequent pellet injection is promising


22

ELM Mitigation Using Pellet Injection

.

A. HerrmannPSI 2002 ?


0

0.1

0.2

0.3

0.01 0.1 1 10

W

ELM

/W

p

∗

ITER

/w o

pellet

∗

p

safe ELMs

Δ

p

4Hz pellet injection in ITER can reduce the energy loss per ELM to acceptable level (A Polevoi et al 19 FEC Lyon

2002)

ELM induced energy loss is reduced in ASDEX Upgrade at sufficiently high frequency of pellet injection (P Lang, 2002)

23

NEOCLASSICAL TEARING MODES (NTMs)

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Neoclassical Tearing Modes (NTMs)


Neoclassical tearing modes (NTMs) are induced by reduction of bootstrap current inside magnetic islands

Deteriorate confinement and determine the lowest beta limit

NTMs methastable: ‘seed’ islands are required

NTM can be stabilized with localized current drive within magnetic island


Neoclassical Tearing Modes (NTMs) Complete 3/2 NTM suppression demonstrated (AUG,

DIII-D, JT-60U) with localized ECCD Complete 2/1 NTM suppression demonstrated (DIII-D) Real-time ECCD position control demonstrated (DIII-D)


Suppression of NTMs in ITER

Extrapolation to ITER: PECCD = (30 ± 15) MW

(G Giruzzi and H Zohm,

ITPA MHD Meering, Naka, Feb 2002)

Early injection would

enable NTM stabilization

with PECCD< 20 MW

ITER design:

PECCD = 20 MW A Zvonkov , 2000

m/n = 2/1

27

DISRUPTION MITIGATION

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Disruption Mitigation

Mechanical loads during disruptions are within the design limits (confirmed by DINA) (M.Sugihara et al, this Conference)

Promising disruption mitigation technique DIII-D: High-pressure noble gas jet injection

(D G Whyte FEC 2002, Lyon)

V. Riccardo, this Session


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Preliminary modeling: the technique is feasible for ITER Operation space limited by melting/ablating the first

wall

Noble Gas Jet Injection in ITER


2

0 1

0

8

00 1 2 3 4 t (ms)

(1021 m-3)

ITER-98

D G Whyte 19th FEC 2002, Lyon

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IMPROVED H-MODE

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Regime with lower current (higher q95) would be beneficial to reduce disruption forces and for access to benign (Type II) ELM regime but requires improved confinement

Recently ASDEX Upgrade, DIII-D and JET demonstrated a possibility to obtain plasmas with improved confinement,

H98(y,2) = 1.2-1.4, at q95 =3.6-4.2

(correspond to I = 12.5 - 10.5 MA in ITER)

Q=10 Scenario at Reduced Current



Advanced H-mode with Type II ELMs

18 MW / m2 6 MW / m2

inner divertor

outer divertor

No sawteeth

q(0) ≥1

N = 3.5

q95 = 3.6

H98(y,2) = 1.3

n = nG

Δt = 40 E

Low divertor heat load (Type II ELMs)

ASDEX Upgrade

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INTERNAL TRANSPORT BARRIERS (ITBs)

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Steady-State Q≥5 Operation in ITER


Requirements

H98P(y.2) > 1.3-1.5

High beta N > 2.6

High bootstrap current fraction,fBS ≥50%

Advanced Tokamak Mode Regimes with Internal Transport Barriers (ITBs)

Weak or negative magnetic shear

Resistive wall mode stabilization

35

ITB Power Threshold


The rarefaction of resonance surfaces at low/zero magnetic shear helps ITB formation while the barrier width is probably controlled by the ExB shear

JET and ASDEX-U indicate importance of rational q in the vicinity of zero magnetic shear

[E Joffrin et al 19th FEC Lyon 2002]

The target plasmas with weak or negative magnetic shear require lower heating power for ITB formation [G T Hoang et al, 29th EPS Conf. 2002]

36V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia 50

Real-Time Control of ITBs in JET

37

JT-60U: ITB and Current Hole


• Current hole and ITB at strong negative shear has been sustained for ~5 s in JT60-U at I = 1.35 MA, q95=5.2, HH98y,2~1.5, N~ 1.7

• T(r) and n(r) are flat inside the current hole

Transiently:I=2.6 MA, q95=3.3, E=0.89 s, Qeq=1.2 HH98y,2~1.5, N~ 1.6

ne(0) = 1020 m-3

38

RESISTIVE WALL MODES (RWMs)

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DIII-D: Dynamic error field corrections by feedback control allows rotational stabilization of RWMs N=N(ideal wall) ~ 2N(no-wall limit) at wrot > 2% Alfven

DIII-D: Negative central shear plasma fBS = 65%, fnon-ind = 85%, T ≥ 4% (E J Strait et all 19th FEC Lyon 2002)

Suppression of Resistive Wall Modes


40

Extrapolation to ITER Model developed taking account realistic vessel and coil

geometry and plasma rotation (A Bondeson, next report) Side correction coils will be used for RWM stabilization

(similar to that in DIII-D)

Suppression of RWM in ITER


Cβ =

βN −βNnowall

βNidealwall −βN

nowall

C = 0.8 is achievable


Requirements for Power Reactor

Analysis study suggests that it is possible to achieve most normalized plasma parameters in ITER to enable projection to fusion power reactor, i.e. demonstration of Pfus~0.7GW and simulation of Pfus ~ 1 GW

(M.Shimada, this Conference, Thursday 10 July)

42

Requirements for Plasma Measurements


The requirements for plasma and first wall measurements on ITER are well developed and many diagnostic systems have been designed to an advanced level

Solutions to many of the difficult implementation issues that arise on a DT machine have been found, and design and R&D is in progress on outstanding issues

It is believed that the measurements necessary for the machine protection and basic plasma control can be made at the required level of accuracy etc, and also many of those now identified as necessary to support the advanced operation

There are several papers on ITER diagnostics presented in the diagnostic sessions on Thursday and Friday afternoons including an overview oral by A Costley on long pulse issues in ITER diagnostics

43

Summary - I


The reference plasma parameters required for inductive high-Q operation in ITER (N = 1.8, q95 = 3, H98(y,2) = 1, n/nG = 0.85) are demonstrated on present machines

The feasibility of achieving Q ≥10 in H-mode predicted by transport model based on empirical confinement scaling is confirmed by dimensionless analysis and theory-based transport modeling

Active control of NTMs and mitigation of ELMs and disruptions may be necessary. Relevant control and mitigation techniques suggested and tested. Extrapolation to ITER needs further work

44

Summary - II


Requirements for ITER steady-state Q≥5 operation (N > 2.6, H98(y,2) > 1.3, fBS > 0.5, n ~ nG) developed. Normalized parameters demonstrated in experiments

More sophisticated control schemes (i.e. current and pressure profiles) will be necessary for steady state operation. Such schemes are under development

Achievement of more demanding normalized parameters (N > 3.6) and high fusion power, 700MW, necessary to facilitate extrapolation of plasma performance to fusion power reactor is under study and looks possible

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LIST OF ITER IT REPORTS AT THIS CONFERENCE

V. Mukhovatov Overview of ITER Physics (Wednesday, July 9) I-3.3AM. Shimada High Performance Operation in ITER (Thursday, July 10) P-3.137M. Sugihara Examination on Plasma Behaviors during Disruptions on Existing

Tokamaks and Their Extrapolations to ITER (Tuesday, July 8) P-2.139A.S. Kukushkin Effect of Carbon Redeposition on the Divertor Performance in ITER

(Thursday, July 10) P-3.195A. Costley Long Pulse Operation in ITER: Issues for Diagnostics (Friday, July 11)

O-4.1D K. Itami Study of Multiplexing Thermography for ITER Divertor Targets (Friday,

July 11) P-4.62T. Kondoh Toroidal Interferometer/Polarimeter Density Measurement System for Long

Pulse Operation on ITER (Friday, July 11) P-4.64T. Kondoh Prospects for Alpha-Particle Diagnostics by CO2 Laser Collective Thomson

Scattering on ITER (Friday, July 11) P-4.65T. Sugie Spectroscopic Measurement System for ITER Divertor Plasma:

Divertor Impurity Monitor (Friday, July 11) P-4.63C. Walker Erosion and Redeposition on Diagnostic Mirrors for ITER: First Mirror Test

at JET and TEXTOR (Friday, July 11) P-4.59C.I. Walker ITER Generic Diagnostic Components and Systems for Integration (Friday,

July 11) P-4.61

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OVERVIEW OF ITER PHYSICS

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