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
ROOTS:
FAST TRACK discussion
Power Plant Conceptual Studies
a Fast Track version 2002
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
can a DEMO be based on a (largely) demonstrated physics scenario?
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
why “hybride” mode considered
•much broader physics base •originally considered for pulsed scenarios
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
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
PROs and CONs of more “advanced” scenarios
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)
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
?
extrapolations: to be verified (or based) on ITER
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
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
(largely) new territory entered with ITER
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
α-particle behaviour (fusion heating)
again more serious issue for ITB-scenarioes thermal ion orbits in an
extreme ITB (“current hole”) discharge on JT60U
needs of significant quantitative progress
(new concepts)
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
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
achievable β-values: resistive wall mode control
method: similar to vertical position control, but on a helical perturbation:
DIII-D
integrated physics/engineering issues
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
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
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
Development of
Integrated & Controlled Scenario
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?
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
example: control of divertor load and tungsten concentration
top-level requirements on technology
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
technology develoment needs
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
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
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
Conclusions: how do requirements map to
“broader approach”
DEMO requirements consistent with „broader-approach“?
IFMIF
Tokamaks
ITER + TBM
temperature
density
Modelling
ITER (scaled)
50 Years of Fusion Power Plant Studies
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