Wendelstein 7-X in the European Roadmap to Fusion Electricity Enrique Ascasíbar for the W7-X Team This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the European Union‘s Horizon 2020 research and innovation programme under grant agreement number 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. Andreas DINKLAGE | Hungarian Plasma Physics and Fusion Technology Workshop | Tengelic, Hungary | 09. Mar. 2015 | Page 2 Outline 1. Introduction and Motivation:Stellarators in the EU Roadmap 2. Operational Phases of W7-X: OP1.1, OP1.2 3. Aspects of SSO 4. Summary Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 3 Wendelstein 7-X, the Engineers’ View mass: 725 t ECRH heating plasma vessel 254 ports 117 diagnostics ports R = 5.6 m OP1.1: 5.3 MW a = 0.5 m 3 Vplasma= 30 m B ≤ 3 T = 5/6 - 5/4 outer vessel 20 planar SC coils 10 divertor units support structure 50 non-planar SC coils plasma 1st ECPD 14.-17. Apr. 2015 4 Europe and W7-X / W7-X and Europe Introduction and Motivation: Stellarators in the Roadmap http://www.efda.org/wpcms/wp‐content/uploads/2013/01/JG12.356‐web.pdf?5c1bd2 EU Roadmap: Eight Missions: Mission 8 Bring the HELIAS line to maturity Long-term alternative to tokamaks: mitigate risks and enhance synergies: Wendelstein 7-X is an integral part of the European fusion development strategy. Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 5 Why stellarators[1]? Introduction and Motivation: Stellarators in the Roadmap Stellarators: the main alternative magnetic confinement concept to the tokamak. external coils generate rotational transform: 3D-confinement without plasma current + intrinsically steady-state - 3D engineering + no current disruptions - 3D core impurity transport + no current driven instabilities - 3D plasma/fast ion confinement + no need significant current drive high neoclassical losses + no runaway electrons - 3D MCF: one generation behind + operation above Greenwald-limit feasible - divertor concept to be verified + lower -particle pressure (for given Pfus) - operation scenarios to be developed Stellarator Optimization[2]: mitigate 3D losses to pave the way to a Fusion Power Plant [1]Helander, Rep. Prog. Phys. 77, 0877001 (2014), [2]Nührenberg, Zille, Phys. Lett. A 114, 129 (1986) Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 6 Why Wendelstein 7-X? Introduction and Motivation: Stellarators in the Roadmap Stellarator Optimization[1]: mitigate 3D losses HELIcal Axis Advanced Stellarator (Neoclassical Optimization) taming locally trapped particles by proper shaping of mod B 75keV D+ W7‐AS W7‐X 75keV D+ © R.Kleiber • proof the concept of stellarator optimization to be a viable path to FPPs • fundamentally new scientific field: lmfp-3D plasma physics 3D impurity transport fast particles & Alfvénic instabilities 3D turbulence high- operation at low i~e improved confinement modes new (island) divertor & SOL physics [1]Nührenberg, Zille, Phys. Lett. A 114, 129 (1986) Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 7 And what is needed for a HELIAS-FPP? Introduction and Motivation: Stellarators in the Roadmap Engineering Study HELIAS 5-B electro-mechanical feasibility of HELIAS fusion-power plants Schauer et al., Fusion Eng. Design 88, 1619 (2013) Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 8 And what is needed for a HELIAS-FPP? Gaps in dimensionless , *, * parameters ITER‐like shaped tokamaks Wendelstein-line 4 4 Lackner, Fusion Sci Technol. 54, 989 (2008) HELIAS 5‐B 3 DEMO 3 2 * * ITER 2 2 10* 10* 2 log(P*) 1 JET log(P*) 1 /2 W7‐X 100* AUG W7‐X 0 100* 0 1st phase /2 2* * */2 W7‐AS 2* * */2 ‐1 ‐1 ‐1 ‐0.5 0 0.5 1 1.5 ‐1 ‐0.5 0 0.5 1 1.5 log(B*) log(B*) From Wendelstein 7-X there is still a large gap to the HELIAS reactor but W7-X is large enough to assess reactor physics aspects. Andreas DINKLAGE | Hungarian Plasma Physics and Fusion Technology Workshop | Tengelic, Hungary | 09. Mar. 2015 | Page 9 High-level objectives and implications Mission 8: bring HELIAS to maturity Operational Phases of W7‐X exploit W7-X to prepare a decision on the next-step HELIAS device: HELIAS-FPP? or HELIAS-BPX? steady-state, lmfp, high-, (low *, *) plasmas and demonstration of favourable 3D confinement/operation high-level priorities : address key issues in 3D confinement safe operation schemes and effects of optimization density control, impurities, fast particles, island divertor/edge implement actions along high-level priorities theory driven exploitation and predictive capabilities delivery of key components and their operation preparation of experimental schemes and FPP physics basis specific actions to comply with WP14-18: prepare SSO + develop in OP1 feasible island divertor scenarios in a robust environment to mitigate later technical risks with water-cooled in-vessel components (OP2) - increase density - explore magnetic configurations compliant with reliable divertor operation + assess aspects of stellarator optimization, develop interpretative and predictive capabilities Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 10 Timeline OP1.1, OP1.2, OP2 Device commissioning Test divertor HHF divertor (cryostat, fields, vacuum) installation installation weeks 2014 2015 2016 2017 2018 2019 divertor commissioning 1st plasma C uncooled divertor water‐cooled HHF div. C limiter; pulsed op. plasmas steady‐state plasmas OP 1.1 pulsed operation OP 2 OP 1.2 11 W7-X: first OPs in figures Operational Phases of W7‐X NC OP 1.1 uncooled carbon limiter PECRH ~ 2 MW (…5MW) Te < 3.5 keV NC 2015 He, (H) gas puff Ti < 0.9 keV surveillance diag. n < 2 x 1019 m-3 13 wks pulse limit: Emax < 2MJ magnetics ISS04 < 0.6% Pulse < 1 s ~ basic n, T, imp. diagnostics NC < 1.6% OP 1.2(a) uncooled test-divertor (C) PECRH ~ 8 MW H, (D) P H ~ 7 MW 2016 NBI NC pulse limit: E < 80 MJ +profiles MHD (n,T, E ,…) Te < 3.5 keV 29 wks max r NC < 10s … min +impurity diagnostics Ti < 3 keV Pulse ~ n < 1.6 x 1020 m-3 OP 1.2(b) test scraper elements + PICRH ~ 1.6 MW ISS04 < 1.2% 2017 P < 10 MW < 3% tot ~ NC 29 wks + blower gun pellet inj. + diagn. upgrades NC OP 2 actively cooled divertor (CFC) PECRH ~ 10 MW Te < 4.5 keV H NC >2019 steady-state capable PNBI ~ 7 MW or Ti < 4 keV D 20 -3 D, H PNBI ~ 10 MW n < 2.4 x 10 m technical limit: 30 min/10MW PICRH ~ 4 MW SS04 < 2 % P /A < 10 MW/m2 P < 20 MW < 5 % target tot ~ NC + quasi cw pellet injection + steady-state upgrades Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 12 Minimum goals of OP1.1 OP1.1 uncooled carbon limiter, He, (H), Emax < 2MJ OP1.1 priorities: first plasma operation and integral commissioning 1. Integral commissioning of all systems needed for successful plasma operation 2. Existence of closed flux surfaces all the way to the limiter (at B=2.5 T) 3. Measurement and adequate reduction of B11 field errors 4. Reliable ECRH plasma startup scenario in He 5. Basic ECRH interlocks and safe operation scenarios 6. Basic impurity content monitoring 18 -3 7. Central Te>1 keV at ne>5*10 m in at least 10 discharges in He Plus some more physics: • Electron root transport studies • Scrape-off layer studies • First experiments with ECRH and ECCD • First comparison He vs H (startup, pumping, confinement) Confirmation of optimization goals of W7-X will be done in later operation phases. T.S. Pedersen, EPS2014 & W7-X PC Meeting, Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 13 Closed flux surfaces view into the W7-X vacuum vessel OP1.1 EDICAM video camera system Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 14 Field errors -4 OP1.1 no magneticB11 = 10 fieldT errors high-iota configuration standard configuration assessment of small field errors development of divertor configurations Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 15 Limiter SOL physics OP1.1 • connection length is short Cross field diffusion rate visibly compared to divertor phase affects heat load patterns • two distinct regions on limiter D (m2/s) 0.5 1.0 2.0 Langmuir probes Langmuir probes Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 16 Scrape‐off layer physics with a limiter • Langmuir probes provide direct measurements ofpower decay length, λq • Can be done piggy-back on confinement studies (vary density and temperature, reach quasi-steady conditions). Enrique ASCASIBAR | 8th IAEA TM ‘Steady-State Operation of Magnetic Fusion Devices’ | Nara, Japan | 26.-29. May 2015 | Page 1717 Predictive simulations of 1st plasmas 100 OP1.1 Plasma 50 energy,kJ 0 4 Te, Ti, keV 2 0 0.9665 0.9662 edge iota 0.9658 0.9655 1500 © Yu.Turkin 1000 bootstrap and total current, A 500 • 500 kW ECRH; 100 ms after burn-through • bootstrap current shielded by the plasma response 0 • strong electron-root feature in the core – high Te 0 0.2 0.4 0.6 0.8 Time/s Andreas DINKLAGE | Hungarian Plasma Physics and Fusion Technology Workshop | Tengelic, Hungary | 09.