The Initial Program of Wendelstein7-X on the Way to a HELIAS Fusion Power Plant

1 FIP/3-1 The Initial Program of Wendelstein7-X on the Way to a HELIAS Fusion Power Plant A. Dinklage 1, R. König 1, H. Maaßberg 1, T. Sunn Pedersen 1, F. Warmer 1, R. Wolf 1, A. Alonso 2, E. Ascasibar 2, J. Baldzuhn 1, C.D. Beidler 1, C. Biedermann 1, H.-S. Bosch 1, T. Bräuer 1, R. Brakel 1, S. Bozhenkov 1, S. Brezinsek 3, R. Burhenn 1, I. Calvo 2, F. Castejon 2, M. Drevlak 1, V. Erckmann 1, F. Effenberg 4, T. Estrada 2, Y. Feng 1, J.-M. García-Regaña 1, D. Gates 5, J. Geiger 1, O. Grulke 1, D. Hartmann 1, P. Helander 1, C. Hidalgo 2, M. Hirsch 1, H. Hölbe 1, W. Kernbichler 6, R. Kleiber 1, T. Klinger 1, A. Könies 1, A. Krämer-Flecken 3, M. Jakubowski 1, G. Kocsis 7, M. Kubkowska 8, H. Laqua 1, M. Laux 1, Y. Liang 3, O. Marchuk 3, N. Marushchenko 1, P. McNeely 1, A. Mishchenko 1, V. Moncada 9, D. Naujoks 1, O. Neubauer 3, J. Ongena 10, M. Otte 1, N. Pablant 5, M. Preynas 1, E. Puiatti 11, M. Rack 3, J. Riemann 1, N. Rust 1, F. Schauer 1, O. Schmitz 4, H. Smith 1, T. Stange 1, T. Szepesi 7, H. Thomsen 1, J. Travere 9, Y. Turkin 1, J.-L. Velasco 2, A. Werner 1, P. Xanthopoulos 1, K. Tanaka 12, K. Ida 12, S. Kubo 12, R. Sakamoto 12, H.Yamada 12, M. Yokoyama 12, M. Yoshinuma 12, The W7-X Team 1 (cf. author list of Ref. [1]), The TJ-II Team 2, The LHD-Experiment Team 12 1 Max-Planck-Institut für Plasmaphysik, Greifswald, Germany 2 CIEMAT, Madrid, Spain 3 Forschungszentrum Jülich, Jülich, Germany 4 U Wisconsin, Madison WI, USA 5 Princeton Plasma Physics Laboratory, Princeton NJ, USA 6 Technische Universität Graz, Graz, Austria 7 Wigner Institute, Budapest, Hungary 8 IPPPLM, Warsaw, Poland 9 CEA/IRFM, Cadarache, France 10 ERM-KMS, Brussels, Belgium 11 Consorzio RFX, Padova, Italy 12 National Institute for Fusion Science, Toki, Japan E-mail contact of main author: dinklage@ipp.mpg.de Abstract. The stellarator concept offers a possible alternative to a tokamak Fusion Power Plant (FPP). Worldwide efforts on different stellarator lines cover a substantial range of candidate magnetic configurations; the EU program focuses on helical-axis advanced stellarator (HELIAS) line. Wendelstein 7-X (W7-X) is the first optimized stellarator to prove the concept of physics-based, optimized shaping of the magnetic field structure aiming an improvement of the plasma performance. To demonstrate reactor potential, W7-X needs to operate reliably at high-power, high-density quasi-continuously with a viable divertor concept. To mitigate technical risks, the first operation phase will use carbon limiters later replaced by an un-cooled test divertor unit (TDU). The robust TDU allows for target-oriented developments towards steady-state, high-performance operation but at limited pulse lengths. The primary goal of the first phase is to develop discharge scenarios at high densities with full density control by successively increasing the density. Systematic variations of the magnetic configuration will allow studying the effect of stellarator optimization. In ECRH heated plasmas, fuelling

2 FIP/3-1 schemes have to be qualified to avoid central density depletion due to neoclassical thermodiffusion. Furthermore, impurity transport needs to be addressed in view of potential accumulation on the way to long pulses. New, basic physics aspects, e.g. the effect of optimization on turbulent transport, MHD stability aspects and the search for improved confinement modes are addressed along the line of steady-state developments. Comprehensive studies, however, will require later steady-state capabilities and high plasma performance. Particularly, the full assessment of fast-ion confinement will likely not be achievable in the first phase but requires long-term preparation. The physics plan to steady-state operation in the initial phase of W7-X with regard to a HELIAS FPP is outlined. 1. Introduction: Perspectives of the HELIAS/Wendelstein Line Stellarators [2] are the main alternative magnetic confinement concept to the tokamak. Since the rotational transform is predominantly generated by external coils, neither a principal limitation of pulse lengths exists nor are strong toroidal plasma currents required. Consequently, current-driven instabilities, disruptions and a Greenwald density limit are not observed in stellarators. Current-drive is considered to be a tool for minor adjustments but does not substantially affect the projected recirculating power in a fusion power plant (FPP). For fast plasma terminating events (e.g. thermal quenches), the confining field remains intact even when the plasma decays. Since a Greenwald-density limit does not exist (but radiative limits depending on B exist at even higher densities), stellarators can work at higher densities than tokamaks. Beyond such economic considerations, it is anticipated that higher densities ease divertor operation [3] and lead to a lower alpha-particle pressure at the same fusion power reducing the drive of fast-ion driven instabilities. Preceding experiments showed that Stellarators provide intrinsic steady-state capabilities with more quiescent plasma behavior also towards operational boundaries, e.g. with regard to the plasma beta [4]. The advantages of stellarators result from the external generation of the magnetic geometry. But the resulting 3D magnetic field also leads to issues to be overcome for a FPP. Beyond consequences for 3D engineering and construction [1,5], the 3D field leads to particles locally trapped in magnetic mirrors. These particles are susceptible to large, so-called neoclassical drift losses. The losses affect both the thermal plasma and alpha particle confinement in an FPP. Moreover, non-intrinsically-ambipolar neoclassical radial electric fields and a lack of temperature screening (as in tokamaks) [6] require concepts to avoid impurity accumulation [7]. The specific 3D magnetic field structure also affects the plasma rotation [2] and is anticipated, e.g., to affect the occurrence of improved confinement modes. A concept to mitigate drawbacks resulting from the 3D magnetic geometry is stellarator optimization [8]: the magnetic field geometry is optimally shaped to reduce neoclassical losses to arrive at improved confinement. In line with the transport optimization, further optimization criteria like the existence of good flux surfaces, favorable equilibrium properties, MHD stability at high plasma beta, fast-ion confinement have been attained in the optimization of concepts named Helical-Axis Advanced Stellarators (HELIAS). W7-X-like magnetic equilibria are characterized by minimized internal plasma currents which result in stiff magnetic configurations, i.e. the influence of the plasma pressure on the magnetic field structure is relatively small (small Shafranov shift) and the plasma diamagnetism increases the magnetic well at high plasma beta. Using low-order rational resonances at the edge of a low-shear configuration, exhaust is provided by means of the island divertor. W7-X can be regarded to be the first experiment realizing the concept of an optimized stellarator at sufficient size to operate at reactor relevant plasma beta values and at collisionalities in the long-mean-free path regime.

3 FIP/3-1 This background and the resulting potentials make the stellarator concept not only a possible alternative to a tokamak FPP but contribute to magnetic confinement in general by offering physics solutions to potential issues on the route to burning plasmas. Nonetheless, since stellarator optimization became feasible only once conceptual and computational methods were available, stellarators can roughly be said to be about one generation behind tokamaks. 2. Introduction: Perspectives of the HELIAS/Wendelstein Line Wendelstein 7-X is undergoing device commissioning at present. First plasma is expected in 2015. The device has a major radius of R=5.5m and an average minor radius of a=0.55m resulting in a plasma volume of about 30 m 3. The five-fold symmetric magnetic field (B=2.5 T), generated by 50 superconducting modular coils and 20 superconducting planar coils, will allow to substantially vary the magnetic configuration along with additional coils (e.g. trim coils) to directly affect the SOL plasma. The divertor will make use of magnetic islands at the 5/4, 5/5 and 5/6 resonances of the rotational transform. Heating is planned to be supplied by ECRH, NBI and ICRH which will be upgraded along the program. ECRH is planned to be the stationary high-power heating source (~ 10MW cw at 140GHz; X2, O2 heating). 3. Operation Phases (OP) of Wendelstein 7-X The technical key for the physics program is steady-state heat- and particle exhaust by qualified divertor operation with actively water-cooled plasma facing components (PFC) [9]. The experimental exploitation of W7-X begins with OP1.1 with carbon limiters. After a brief shut-down, an uncooled test divertor unit will be installed for OP1.2. These first experimental campaigns have the goal to acquire a sound basis for the development of a high-performance (high nt E ) operation after the installation of water-cooled PFCs for OP2 making W7-X steady-state capable. Operation at high densities is required for safe (highly radiating) divertor operation and to achieve high plasma performance. Possible upgrades to the divertor, e.g. introducing metallic PFCs, will benefit from the first steady-state results in OP2 and developments from tokamaks but are beyond the scope of a present discussion. The operation phases are expected to begin in 2015 (OP1.1), 2016 (OP1.2) and 2019 (OP2) after the installation of respective PFCs. The specific sequence is a staged approach to mitigate technical risks. The shut-down after OP1.1 will be used for fixing technical problems in diagnostics, control and device technology and for upgrades. A shut-down after OP1.2 serves for the installation of the actively cooled divertor. Technically, OP1.2 is regarded to make W7-X ready for steady-state operation but without risks from water-cooled PFCs. 4. OP1.1 This very first phase OP1.1 is planned to take three months. Primary goal is the first plasma operation and the commissioning of diagnostics, heating systems (ECRH), components, control systems and data acquisition. Error field correction (in the range of B/B~10-5 ) with trim coils is possible even in this early phase and will be used to explore heat exhaust. Vacuum flux surface measurements are planned to verify the existence of good flux surfaces (one of the W7-X optimization criteria). In order to verify the predicted flexibility of the coil system and as a basic reference for divertor scenario developments, a systematic exploration of magnetic configurations (e.g. variation of the toroidal mirror cf. [10]) is planned.

4 FIP/3-1 FIG. 1. Predictive simulations for first discharges in W7-X (0.5 MW ECRH, 100ms, n=2x10 19 m -3 ). First plasmas will be created with a simplified set of in-vessel components (five inboard limiters). The deposited energy is limited to E < 2MJ. Plasma operation will start with a technical demonstration of ECH plasma break-down and device conditioning and the operation of X2 heated plasmas. Since plasma break-down and the discharge control are easier with helium, the program starts with helium and will eventually go to hydrogen once first experience has been attained. With this background, low density plasmas (n e ~ some 10 19 m -3 ) with short pulses (~s) appear to be feasible. Estimates of the achievable temperatures are shown in the example in Fig. 1. The results show time-traces of predictive simulations of the plasma temperature and stored energy for the initial discharges (flat density profiles at maximum densities of 2 x 10 19 m -3 ). The simulations indicate high electron temperatures of some kev due to stellarator-specific electron heat transport (central electron root confinement [11]). The assessment of this neoclassical effect along with the exploration of ECRH heating scenarios offers room for physics investigations. The ECRH will be ready to allow for first ECCD studies. The limiter configurations have a shorter connection length and can be therefore used for scrape-off layer investigations. Cross field diffusion will become visible in the limiter heat-load pattern. First assessments of radiative mantle discharges appear to be feasible and are being prepared. 5. OP1.2 After a shut-down to install an un-cooled but robust graphite divertor (TDU), OP1.2 experiments will be conducted. The TDU closely matches the shape of the actively cooled high-heat flux (HHF) divertor which will be installed for OP2. Due to the lack of active cooling, the pulse lengths in OP1.2 are limited; the allowable deposited heating energy to W7- X is E max ~80MJ. OP1.2 is planned to consist of two campaigns of 29 weeks each. The staged approach OP1 OP2 mitigates risks associated with the HHF targets because the TDU is more robust to overloads. Although the HHF divertor is steady-state capable, it bears the risk of failure if its load limits are locally exceeded. Consequently, any scenario qualification will have to be performed more carefully in OP2. In contrast, OP1.2 offers a unique possibility for a much more aggressive and free exploration, since one can run scenarios that are not a priori guaranteed to stay within the limits of the actively cooled divertor. Typical discharge lengths in OP1.2 will be limited to 5-10 s at maximum anticipated heating powers (8MW ECRH, 7MW NBI in hydrogen) but may be extended to a minute if the deposited energy stays below E max. The working gas is hydrogen; deuterium discharges will be limited by technical reasons. H-modes in stellarators have been observed but the basic

5 FIP/3-1 confinement scenario in stellarators is characterized by optimized neoclassical transport. Improved confinement modes (e.g. like the so-called HDH mode [18]), however, might be required for high-density operation and impurity control. The physics plan needs to effectively address questions of importance for OP2. The physics plan is not resource driven but is guided by physics priorities, logical sequences and operational constraints. Those cover not only the aforementioned shorter pulse lengths, but also lower heating power and a lack of divertor pumping, relative to OP2. Nonetheless, OP1.2 will allow addressing central elements of stellarator optimization and objectives of highest scientific interest for the first time. In line with the mission of the Wendelstein 7-X project, i.e. to bring the HELIAS line to maturity, the top-level goals of OP1.2 can be summarized to be: 1. Proof and assessment of elements of stellarator optimization (small Shafranov shift, aspects of MHD stability, good neoclassical confinement, and small bootstrap currents) 2. The preparation of high-density, high-power steady-state plasma operation (scenario development, qualification of divertor operation, heating, fuelling, exhaust and impurity control) To attain at these goals, the physics plan is based on guiding principles for the specific development of an experimental program; physics topics of relevance (see below) are aligned to these guiding principles. The first guiding principle for the program in OP1.2 is to successively increase the plasma density while remaining in the long-mean free path regime by adequately increasing the heating power. High-density operation is the prerequisite both for safe high-power operation of the steady-state divertor and to achieve high plasma performance. The increase of density gives rise for a sequence of qualification actions for the later operation phase OP2. Specifically, the qualification actions in OP1.2 are roughly in temporal order but likely evolving during OP1.2: achieve tractable (symmetric) power loads on the divertor plates achieve plasma density control and examine elements of appropriate fuelling schemes qualify techniques and tools for configuration control (e.g. X2-ECCD, field variations and corrections (trim coils) and determine whether scraper elements could and should be used in OP2) qualify heating schemes in view of high-density discharges (transition from X2- to O2 heating) develop concepts for the avoidance of impurity sources and accumulation and for controlled PWI explore operation parameters and control actuators for the safe operation of the island divertor. achieve high separatrix (upstream) density scenarios with the island divertor develop techniques for highly radiating and high-recycling divertor scenarios The second guiding principle is to exploit the flexibility in magnetic configuration space to assess the benefits of stellarator optimization along the path of increasing the density. This will allow addressing key aspects of the W7-X stellarator optimization under different conditions. One key element is to define magnetic configurations compatible with divertor operation; an option is to avoid bootstrap currents as in high-mirror configurations [10].

6 FIP/3-1 Experimentally, the flexibility of the W7-X magnet system will be systematically exploited and the effect of additional coils (trim-coils) and components (scraper element) will be investigated. A simultaneous verification of all W7-X optimization criteria will be left for OP2 and later campaigns. Nonetheless, the following important topics can be started to be addressed to a significant extent already in OP1.2: increasing beta; high beta discharges will be done in OP2 first assessment of fast particle confinement (NBI, ICRH, Diagnostic Injector), development of required diagnostics and study its configuration dependence [12] 3D impurity transport (source control, asymmetries, PWI, neoclassical effects) turbulent transport in 3D fields [13] with the interplay of neoclassical electric field shear, zonal-flows and configuration effects on micro-instabilities MHD stability in specific de-optimized configurations (interchange (magnetic hill variation), ballooning instabilities (variation of the toroidal mirror)) improved confinement modes (HDH-mode, H-mode) It is understood that the implementation of the plan for each of the aforementioned topics depends on resources. The program will be adjusted to the progress of device capabilities, such as plasma performance and the availability of various subsystems (heating, fuelling, diagnostics, control, and data acquisition). At the same time, modeling tools will be employed for discharge simulations and analyses and will thereby be validated taking benefit from a theory driven experimental exploitation. 6. OP2 and beyond For the second operation phase OP2, more heating power will be available (P ECRH = 10 MW (long pulse heating), P ICRH = 5MW, P H+ NBI = 7MW, P D+ NBI=10MW, NBI will be available already earlier in OP1.2). The HHF divertor with actively cooled carbon fibre composite target elements is specified to handle heat fluxes up to (P/A) target = 10MW/m 2. These two enhancements open the gate to high-power steady-state operation and the exploration of high- plasmas. With regard to reactor operation, true steady-state discharge scenarios need to integrate fusion relevant performance parameters. High- operation in W7-X with stiff equilibria will allow pursuing studies on crucial physics issues for reactor operation, e.g. the effectiveness of fast particle confinement [12]. The validation of understanding of 3D Alfvénic instabilities [15] is highly relevant to the operation of FPPs. While the alpha particle pressure is lower at the same fusion power in stellarators, the Alfvénic spectrum is specific to HELIAS devices and little is known about collective 3D effects. From the present perspective, later modification of the in-vessel components could be expected beyond OP2, e.g. to metallic divertor targets. These upgrades require a detailed understanding of the plasma-wall interaction. In view of the development of the HELIAS line, the time for a decision point about a burning plasma stellarator remains to be specified depending on the physics basis. It appears to be self-evident, however, that this decision point needs to be attained sufficiently early to make the HELIAS line an alternative route to fusion. 7. Projections to Stellarator FPPs and Programmatic Implications Beyond W7-X and in view of FPPs, step-ladder plots (Fig. 3) allow one to assess the progress of magnetic confinement devices (of similar geometry) on the way to a fusion

7 FIP/3-1 reactor. These plots show the operation regime of fusion devices in figures of a convenient choice of leading linear combinations of powers of engineering variables [16]. With the provision of a relation of the energy confinement time (usually a scaling law for E, here ISS04 [17]) the plots allow to indicate the leading dimensionless physics variables (plasma beta normalized gyro-radius * and collisionality of the plasma *) to elucidate both gaps in physics and engineering quantities at the same time. The scaling relation also allows one to determine figures for the fusion performance like Q and nt E. FIG 3. Step-ladder plots the HELIAS stellarator line. Since a Greenwald limit does not exist for stellarators, the density dimensionless engineering variable is not kept constant and the *, * lines vary a bit as indicated by different line styles in Fig. 2. The W7-X operation region is plotted for the first and the second operation phase indicating the central role of W7-X for the HELIAS/Wendelstein line from W7-AS [18] to power-plant concepts (HSR4 [19]) and reactor studies (e.g. HELIAS 5-B [5]). In addition, the region Q=10 indicates the operation region of an ITER-like burning plasma stellarator. It will be part of the W7-X program to prepare a decision about the necessity for such a device allowing on to study collective burning plasma effects but without a breeding-blanket and relaxed material requirements. Physics aspects from burning plasmas (e.g. Alfvénic instabilities) do not yet enter the extrapolation in Fig. 2. W7-X and ITER results will need to be assessed for a decision about a burning plasma, ITER-like HELIAS device. For the realization of a HELIAS FPP, benefits from tokamak DEMO-related technology developments are expected to be included. Having these long-term goals of the HELIAS line as a background, the physics program of W7-X will need to demonstrate reliable operation scenarios. These discharges need to show steady-state, reactor-relevant performance in figures of, * and * (high-beta, long-meanfree path plasmas). The gaps in Fig. 2 emphasize again the important role of theory-based predictive capabilities. 8. Summary Wendelstein 7-X is the crucial milestone for the development of HELIAS reactors. The route to a potential FPP is guided by clear high-level physics requirements (,, ), required

8 FIP/3-1 solutions for the power and particle exhaust and the development of steady-state scenarios. Arising from the specific physic of 3D confinement, fuelling and density control, divertor operation, impurity transport, fast-ion confinement, the interplay of turbulent and neoclassical transport and MHD need to be addressed to develop a physics basis for a HELIAS FPP. The physics plan for the first operation phase will start to address these key elements but will also allow the flexibility to cope with new insights; concepts for scientific risk mitigation take benefit from collaborations with stellarators in operation. The specific experiment plan will need to bring the physics goals in line with specific device capabilities and available resources. While modeling results have already been extensively used for the development of this plan, it is worthwhile to emphasize the role of modeling, both for a theory driven experiment planning for Wendelstein 7-X and the development of predictive capabilities for a reliable extrapolation to a HELIAS FPP. Acknowledgement: 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 9. References [1] BOSCH, H.-S. et al, NUCL. FUSION 53, 126001 (2013). [2] HELANDER, P., REP. PROG. PHYS. 77, 087001 (2014). [3] FENG, Y. et al., PLASMA PHYS. CONTR. FUSION 53, 024009 (2011). [4] WELLER, A. et al., NUCL. FUSION 49, 065016 (2009). [5] SCHAUER F. et al., FUSION Eng. Design 88 1619 (2013). [6] MAASSBERG, H. et al., PLASMA PHYS: CONTR. FUSION 41, 1135 (1999). [7] BURHENN, R., NUCL FUSION 49, 065005 (2009). [8] NÜHRENBERG, J. and ZILLE, R., PHYS. LETT. A 114, 129 (1986). [9] KLINGER T. et al, FUSION ENG. DESIGN 88, 461 (2013). [10] GEIGER, J., et al., PLASMA PHYS. CONTR. FUSION (ACCEPTED, 2014). [11] YOKOYAMA M., et al., NUCL. FUSION 47, 1213 (2007). [12] DREVLAK, M. et al., NUCL. FUSION 54, 073002 (2014). [13] XANTHOPOULOS, P., et al. PHYS. REV. LETT. 107, 245002 (2011). [14] TANAKA, K. et al., FUSION SCI. TECHNOL 58, 80 (2010). [15] KÖNIES, A., and KLEIBER R., PHYS. PLASMAS 19, 122111 (2012). [16] LACKNER, K., FUSION SCI. TECHNOL. 54, 989 (2008). [17] YAMADA, H. et al., NUCL. FUSION 45, 1684 (2005). [18] HIRSCH, M. et al., PLASMA PHYS. CONTR. FUSION 50, 053001 (2008). [19] WOBIG, H. et al., NUCL. FUSION 43, 889 (2003).