1
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
2
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
1/2D ITINT1.SAS code with Psep ≥ PL-H
O.J.W.F. Kardaun NF 42 (2002) 841
11
Theory Based Transport Models
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
18
Power and Particle Control in ITER
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
Sawteeth
Edge localized modes (ELMs)
Neoclassical tearing modes (NTMs)
Alfven instabilities
Disruptions
20
EDGE LOCALIZED MODES (ELMs)
21
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
22
ELM Mitigation Using Pellet Injection
.
A. HerrmannPSI 2002 ?
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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)
24
Neoclassical Tearing Modes (NTMs)
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
25V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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)
26V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
28
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
29
Preliminary modeling: the technique is feasible for ITER Operation space limited by melting/ablating the first
wall
Noble Gas Jet Injection in ITER
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
2
0 1
0
8
00 1 2 3 4 t (ms)
(1021 m-3)
ITER-98
D G Whyte 19th FEC 2002, Lyon
30
IMPROVED H-MODE
31
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
32V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
33
INTERNAL TRANSPORT BARRIERS (ITBs)
34
Steady-State Q≥5 Operation in ITER
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
• 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)
39
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
Cβ =
βN −βNnowall
βNidealwall −βN
nowall
C = 0.8 is achievable
41V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
V. Mukhovatov et al., 30th EPS Conf. on Control. Fusion and Plasma Phys., July 7-11, 2003, St Petersburg, Russia
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
45
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
Top Related