k.lackner*) max-planck institut für plasmaphysik, d-85748 garching *) based largely on work of efda...
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K.Lackner*)
Max-Planck Institut für Plasmaphysik, D-85748 Garching
*) based largely on work of EFDA and the EU DEMO-Working Group
Technology and Plasma Physics Developments Needed for DEMO
DEMO: implicite
ly defined by FAST TRACK
discussion:
single interm
ediate step between ITER and
a (potentia
lly) first o
f a kind fu
sion power plant
EFDA (D. Campbell, D. Maisonnier, P. Sardain) + M.Q. Tran; G. Janeschitz, K. Lackner, G. Marbach, M. Ravnik, B. Saotic, D. Stork, D. Ward;
A.Kallenbach, A. Sips
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ROOTS:
FAST TRACK discussion
Power Plant Conceptual Studies
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a Fast Track version 2002
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DEMO Working Group
following completion of PPCS
•identical or scalable with high confidence to a first generation power plant (physics technology AB↔C)
•physics and technology demands – except availability – similar to PP
•for DEMO (vs. PP): construction costs rather than COE decisive → Pel ≤ 1.0 GW
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can a DEMO be based on a (largely) demonstrated physics scenario?
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DEMO base-line assumptions
2 basic physics operation modes considered
0
1
2
3
4
5
0 0.5 1
r/a
strong
weak
q
Standard H-mode
~ zero shear
Reversedshear ITER standard
operating scenario
„improved H-mode“ a.k.a. „hybrid mode“
„internal transport barrier“: ITB -modes
ITER- baseline
ITER-steady
1st generation reactor designs
“advanced” reactor designs
n 1.8 3.1 3.5 - 4 > 4
<> [%]
2.5 2.9 2.2 - 3 3 - 5
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why “hybride” mode considered
•much broader physics base •originally considered for pulsed scenarios
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a pulsed DEMO/PP option?
known objections
•pulsed loads
•need for continuous power output (energy storage requirements)
•power supplies for rapid restart
considered in the expectation:
• could be designed largely on demonstrated physics base
• inductive current drive energetically favourable 210O
P
UI
CD
loopp
preliminary conclusions (D.Ward et al., based on PROCESS-Code):
•same physics basis as pulsed device, allows also (more favourable) DC device
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why “hybrid mode” considered
a 1 GWel DEMO(Process-Code )
achieved parameter sets start overlapping with DEMO, PPCS assumptions
even an established physics scenario needs
extrapolations (to be verified)
development into an integrated scenario
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PROs and CONs of more “advanced” scenarios
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what are the “PROs” of ITB scenarios?
cause: suppression of turbulence in a layer in core (analogy to H-mode)
precondition: weak or reversed shear
efficient use of bootstrap current (high fraction & distribution)
good confinement (H-factor)
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intrinsic problem of ITB scenarios
pressure and current profiles (li..internal inductance) unfavourable for stability
→ only weak barriers, at large radius stable
1
2
3
4
N
2 4 6Pressure peaking: p0/<p>
unstable
ConventionalH-mode
a
pla
sma
pre
ssu
re
0
ITB H-mode
a
pla
sma
pre
ssu
re
0
ITB
H-mode
AUGDIII-DJT-60UJET
?
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extrapolations: to be verified (or based) on ITER
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confinement
confirm assumptions for H and “hybrid” H-modes
establish a scaling for ITB - modes
at constant n*, for ITER98(y,2)
AUG
JET
ITER
device operating regimes in dimensionless
“engineering” variables
222
222
/108
/10/*
)/(0032.0/
tet
emfp
tii
BTn
TqRnRq
RBTR
t
heat
tt
B
nan
aPP
aBB
4/3
4/3
4/5
*
*
*
dimensionless physics parameters only known after experiment
close to Greenwald
extrapolation to ITER/DEMO
small in β
large in ρ*, and particularly! in ν*
ρ*
ν*
β
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current drive: efficiency and controllability
“hybrid”:
efficiency very important (small fbootstrap) . γ ≈ = 0.5-0.6 needed
modest control requirements, central current drive o.k.
“ITB scenarios”:
high control requirements
off-axis c.d. probably needed
controllability : differing
cross-diffusion of fast particles
excitation of AE modes
NBI 0.35-0.4 *)
LH 0.3-0.35
ICRF 0.3-0.4 *)
ECCD 0.15 *)
ITER-estimates
*) extrapolated to ITER-temperatures – to be demonstrated!
figure of merit of efficiency
RICDP
n
1020(m 2AW 1)
discrepancy between predicted and observed distribution of NBI driven current on ASDEX Upgrade
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(largely) new territory entered with ITER
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α-particle behaviour (fusion heating)
fast particles (due to NBI or ICRH) cause range of resonant interactions, potentially leading to their loss
fusion-αs different through isotropy
figures of merit:
further increase in
reactor
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α-particle behaviour (fusion heating)
again more serious issue for ITB-scenarioes thermal ion orbits in an
extreme ITB (“current hole”) discharge on JT60U
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needs of significant quantitative progress
(new concepts)
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achievable β-values: limits depend on discharge duration
wall stabilization
NTMs
nonstationarity of current (i.e. q) - distribution
ARIES -AT
PPCD - D
PPCD - A
ITER-FEAT, reference
type of intervention:
external current drive
feedback by localized current drive (ECCD)
magnetic feedback + resistiv wall
most demanding (least demonstrated): control of
resistive wall modes
needed for ITER
needed for DEMO
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achievable β-values: resistive wall mode control important for ITB-scenarioes
for high li (hybrid H-mode) modest need and gain
for low li (“ITB-scenarios”) strong need and significant gain
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achievable β-values: resistive wall mode control
method: similar to vertical position control, but on a helical perturbation:
DIII-D
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integrated physics/engineering issues
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physics/technology interface: plasma wall interaction
tritium retention and material erosion → full high-Z (tungsten) pfc solution:
not in ITER starting configuration → to be added – at latest – in phase 2 of operation
divertor load issue more severe on DEMO/PP than ITER
•higher power & power density
•divertor cooling (He; high duty cycle) not more efficient
P f u s
[G W ]
R o
[m ]
P r a d / P h e a t1 ) Q q d iv , n o m 2 )
[M W / m 2 ]
q d e s ig n
[M W / m 2 ]
IT E R - re f. 0 . 5 6 . 2 0 . 8 1 0 5 1 0
IT E R - S S 0 . 3 6 6 . 2 0 . 8 5 5 1 0
P P C S - B 3 . 4 7 . 5 0 . 8 1 5 2 2 1 0
P P C S - D 2 . 5 6 . 1 0 . 8 3 5 2 0 5
A R IE S - A T 2 . 2 5 . 5 0 . 8 5 1 1 0 - 2 0 3 ) 5
1 ) P h e a t = ( P f u s ( 1 / 5 + 1 / Q ) )
2 ) q d iv , n o m = ( P h e a t- P r a d ) / ( 4 R o F ) w ith g e o m e tr y fa c to r F = 1 0 a n d m id p la n e h e a t fl u x
w id th s c a l in g l ik e 5.0003.0 oR
3 ) d e p e n d in g o n d o u b le n u l l c r e d it
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reduction of divertor load by radiation:
higher fraction of radiative losses than ITERlimits to edge radiation? → higher-Z radiators
•less dilution & Zeff
•more core losses effect on H-mode pedestalbenefit from profile stiffness
ITER´s power handling limit, and scaling of problem with size
→ no direct test of solution possible DEMO solution will have to be an extrapolation based on quantitative understanding of carefully chosen experiments on ITER & elsewhere
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pulsed loads and anomalous events
cyclic pulsed loads (ELMs) .. DEMO constraints even more severe than ITER (because of duty cycle and availability requirements)
anomalous events: specification 0.1 – 1*) disruption /year
•multifaceted nature of disruptions •dedicated campaign phase on ITER to demonstrate achievability (during stage 2 with tungsten)..discharge number rather than time counts*) depending on mitigation success
successive elimination of causes of disruptions: analogy to radioactive decay characteristics of realistic materials
→ when disruption control is improved, previously hidden causes (isotopes) dominate
improved control measures
disr
uptio
n r
ate
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Development of
Integrated & Controlled Scenario
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plasma control: a multifacted issue requiring a highly integrated approach
example: control of divertor load and tungsten concentration
dangers: mitigation (actuators):
high heat load to divertors
high radiation losses supress ELMs, absence of ELMs reduces W-impurity screening
central electron heating by ECRH,ICRH causes impurity pump-out
flat heating profile or peaked density causes W-accumulation at center
impurity and gas puffing increases radiation losses
artificial triggering of ELMs (pacemaking) by pellets screens impurities
show on ITER:
how does α-particle heating work?
peaked density profiles on ITER/DEMO?
scaling of needed central heating power?
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proof of the working of individual actuators
effect of a missing pellet on edge impuríty density
effect of switching on ECRH on central tungsten concentration
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example: control of divertor load and tungsten concentration
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top-level requirements on technology
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DEMO technology: credible 1st generation PP
•from day1 of DT operation: self-sufficiency of tritium
•satisfy same high levels of safety and environmental compatibility as demanded in EU PPCS (requiring, among others, use of low activiation materials)
•aim at a high availability:
•to produce the neutron fluences needed for testing
•(during later stage) to extrapolate to an attractive reactor
•technology requirements similar to 1st generation PP (also not beyond)
•exception: operational experience
•in this regard: DEMO an experiment
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technology develoment needs
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DEMO technology: progress beyond ITER
•use of low activation structural and functional materials (operating temperature window critical) – IFMIF tested including joining (to 80 dpa for first wall/blanket components
•RAFM (EUROFER, possibly modified by ODS)
•divertor materials t.b.d. (tungsten based)
•ITER-like magnet technology – or HTSC?
•tritium breeding and handling
•as base-line for first stage a blanket validated in modules on ITER phase 1 in thermo-mechanics, thermohydraulics
•helium cooled (DC, if SiC-SiC timely available)
•full fuel self-sufficiency
•tritium accountability O(100) more demanding than in ITER
*)classification as established predates Ciacynski-presentation
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DEMO technology: progress beyond ITER
•divertor and first wall
•material tested on ITER
•divertor cooling concept compatible with blanket (development of He-cooling)
•heating and current drive systems
•reduce to 2 out of the 4 systems included or options for ITER
•raise plug efficiency
•possibly push to higher performance (NBI →2MeV ?)
•demonstrate the long-pulse, long-term reliability (testing)
NBI ≤ 0.6
LH ≤ 0.6
ICRF ≤ 0.5
ECCD ≤ 0.45
plug efficiencies expected*)
*) conclusions of EFPW 2005
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Availability: where DEMO is in a different category from ITER
•remote maintenance and repair
•segmentation driver of effort
•compromise between modularity (use testing on ITER) & limited number of elements
T. Ihli et al., this conference
•design target for availability:
•testing of internal components to 50dpa before start of design of FPP -> availability ≥ 33 %
•second stage: make credible that if operated in a routine fashion an availability >75% could be achieved
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Conclusions: how do requirements map to
“broader approach”
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DEMO requirements consistent with „broader-approach“?
IFMIF
Tokamaks
ITER + TBM
temperature
density
Modelling
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ITER (scaled)
50 Years of Fusion Power Plant Studies