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: implicitely defined by FAST TRACK

discussion:single intermediate step between ITER and

a (potentially) first of a kind fusion 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 1r/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

PUI

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

apl

asm

a pr

essu

re0

ITB H-mode

a

plas

ma

pres

sure

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

Bnan

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 : differingcross-diffusion of fast particles

excitation of AE modesNBI 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

n1020 (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 fu 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 f a c to r F = 1 0 a n d m id p l a 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 ra

te

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