K.Lackner*) Max-Planck Institut für Plasmaphysik, D 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 ABC) 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 r/a strong weak q Standard H-mode ~ zero shear Reversed shear ITER standard operating scenario improved H-mode a.k.a. hybrid mode internal transport barrier: ITB -modes ITER- baseline ITER- steady 1 st generation reactor designs advanced reactor designs n > 4 [%]

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 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 GW el 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 (l i..internal inductance) unfavourable for stability only weak barriers, at large radius stable

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 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*ρ* ν*ν* β

current drive: efficiency and controllability hybrid: efficiency very important (small f bootstrap ). γ = 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 *) LH ICRF *) ECCD0.15 *) ITER-estimates *) extrapolated to ITER- temperatures – to be demonstrated! figure of merit of efficiency 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 l i (hybrid H-mode) modest need and gain for low l i (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

reduction of divertor load by radiation: higher fraction of radiative losses than ITER limits to edge radiation? higher-Z radiators less dilution & Z eff more core losses effect on H-mode pedestal benefit 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 disruption rate

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