ARTICLE IN PRESS Fusion Engineering and Design xxx (2010) xxx xxx

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1 Fusion Engineering and Design xxx (2010) xxx xxx Contents lists available at ScienceDirect Fusion Engineering and Design journal homepage: Review of stellarator/heliotron design issues towards MFE DEMO Akio Sagara a,, Yuri Igitkhanov b, Farrokh Najmabadi c a National Institute for Fusion Science, Fusion Engineering Research Center, Oroshicho, Toki , Japan b Karlsruher Institut für Technologie (KIT), IHM, Karlsruhe, Germany c University of California, San Diego, La Jolla, CA, USA article Article history: Available online xxx Keywords: Stellarater Heliotron Reactor design Modular coil Blanket Maintenance info abstract The first review is presented for the recent stellarator designs of FFHR, HSR, ARIES-CS, which are based on the experimental device projects of LHD, W-7X, NCSX, respectively. The main advantageous features of those designs are common on (1) current-free plasmas, which leads to steady, high Q and disruption-free reactors and (2) high density operations, which largely reduces the diverter heat load. Design optimization and engineering feasibility are highlighted including maintainability, fabricability and showing future R&D key issues with prospect towards DEMO Elsevier B.V. All rights reserved. 1. Introduction 2. Main advantageous features Due to inherently current-less plasma, stellarator concepts [1] including heliotron [2], helias [3] and quasi-axisymmetry (QAS) [4] have attractive advantages in magnetic fusion energy (MFE) reactors, such as steady operation without current drive and dangerous current disruption. This means high potentiality on reactor economy and engineering feasibility. Therefore, from 1970s in parallel to rising of tokamak experiments, many reactor designs have been proposed based on heliotron with continuous winding or modular coils [5 13], helias [14], modular helias-like heliac (MHH) [15] or QAS [16] as shown in Fig. 1. On the other hand, non-axisymmetric magnetic field configurations originate challenging subjects on both of plasma physics and device engineering. In particular, high-energy particle confinement related to alpha-heating and highly accurate three-dimensional superconducting magnet systems with sufficient blanket space and maintenance ports are major constraints on reactor designs. On the basis of physics and engineering results achieved recently in major projects such as LHD (heliotron) [17,18], W7-AS [19], W7- X (helias) [20], and NCSX (QAS) [21], much progress has been made in power reactor designs, namely the LHD-type reactor FFHR (Force Free Helical Reactor), the W7-X type reactor HSR (Helical Stellarator Reactor), and the NCSX type reactor ARIES-CS (compact stellarator). Therefore, from the aspect towards MFE DEMO, this paper makes first comparative review of main progress and key issues for those stellarator designs. Corresponding author. Tel.: ; fax: address: sagara.akio@lhd.nifs.ac.jp (A. Sagara) FFHR The main feature of FFHR is force-free-like configuration of helical coils. This gives three merits: (i) simplification of coils supporting structures by opening areas for maintenance works; (ii) widening of the coil-to-plasma clearance needed for the blanket and nuclear shielding; and (iii) use of high magnetic fields. In the LHD-type reactor, the coil pitch parameter of continuous helical winding can be adjusted beneficially to reduce the magnetic hoop force while expanding the blanket space, where =(ma c )/(lr c ) with a coil major radius R c, a coil minor radius a c, a pole number l, and a pitch number m. The second feature is the adoption of self-cooled molten salt Flibe blanket due to attractive merits on safety aspects: low tritium solubility, low reactivity with air and water, low pressure operation, and low MHD resistance [22]. This blanket design has encouraged and initiated the Flibe blanket R&D activities as international collaborations [23]. The third important feature is the neutron wall loading lower than 2 MW/m 2. Keeping this condition, the blanket lifetime can be about 10 years, and if the STB (Spectral-shifter and Tritium breeder Blanket) proposed for the first wall is feasible, the replacement-free blanket concept is possible [12] HSR The Helias reactor based on a Helical Advanced Stellarator configuration, which is optimized so as to achieve simultaneously several important criteria: smooth magnetic surfaces without /$ see front matter 2010 Elsevier B.V. All rights reserved.

2 2 A. Sagara et al. / Fusion Engineering and Design xxx (2010) xxx xxx Fig. 1. Stellarator reactor designs and current typical designs of HSR, FFHR and ARIES-CS. significant islands in the confinement region; good finite-beta equilibrium and stability properties; small (neoclassical) losses; negligible bootstrap currents; good alpha-particle confinement at finite-beta-values and good feasibility of modular coils. The design criteria of the Helias reactor concept are based on presently known physics from stellarator experiments (conservative approach). The magnetic field is chosen as small as possible in order to facilitate the mechanical problems associated with forces and stresses. The ignition condition and self-sustained burn, based on presently known confinement scaling laws, requires a plasma radius 1.8 m. No improvement factor is assumed. To meet these requirements all dimensions of W7-X is scaled up by a factor of 4 leading to a major radius of R = 22 m (HSR5/22). The modular coil system comprises 50 coils. The SC cable for HSR is an upgrade version of a cable-in-conduit conductor, proposed for W-7X, which uses NbTi at 1.8 K with forced-flow cooling of super-fluid helium. Reduction of the mechanical stresses in modular coils requires a proper support structure [24] ARIES-CS ARIES-CS study is an integrated assessment of compact stellarator configuration [16] and was concluded in The major goals of ARIES-CS study was: (1) to investigate if stellarator power plants can be made to be comparable in size with advanced tokamak power plants in order to reduce the cost of electricity and (2) to understand and quantify the impact of complex 3D shape and geometry of fusion core components on the design, manufacturing, and operation of these components. Study focused on quasiaxisymmetric (QAS) configurations as they can to be operated at a lower plasma aspect ratio ( 4 5) compared to other stellarator configurations and can potentially lead to smaller devices. In addition, the particle drift orbits in a QAS configuration are similar to that of a tokamak and it is argued that QAS configuration would have good confinement. Starting from NCSX [21] configuration, effort to reduce -particle loss rate led to the new criteria for optimizing QAS configuration [25]. Extensive optimizing of external coils was also performed to increase the plasma coil spacing. A new approach was adopted to downsize the blanket and utilize a highly efficient WC (tungsten carbide)-based shield in the spaceconstrained regions where plasma is close to the coil [26,27]. As a result, ARESI CS major radius was reduced to 8m. On the engineering side, ARIES-CS study found that the device configuration, assembly and maintenance drive the design optimization and technology choices in many cases [26]. Examples include port-based maintenance scheme which drives the internal design of fusion core and led to the choice of a ferritic steel, dual-coolant blanket and the irregular shape of the superconducting coil that necessitates development of inorganic insulators for high-field magnets. 3. Design optimization and engineering feasibility 3.1. FFHR As a key feature of heliotron reactors, magnetic field configurations are directly coupled to the blanket space as well as to the core plasma performance under key physics and engineering constraints. Therefore, after concept definition of the initial FFHR1 (l = 3) design, optimization studies have been needed on the reactor size, based on the LHD-type (l =2, m = 10) compact design FFHR2 ( = 1.15, R c = 10 m) and modified FFHR2m1 ( = 1.15 and outward shifted plasma axis, R c = 14 m) and FFHR2m2 (inward shifted plasma axis, R c = 17 m) with the ISS95 enhancement factor less than 1.8 [28] as shown in Fig. 2, comparing with ITER, W7-X, NCSX and designs of HSR and ARIES-CS. In those studies, it is found that, with increasing the reactor size, the capital cost does not drastically increase, because the total magnetic energy, which provides the mass of coils supporting structure under the Virial theorem, increases only in proportion to R 0.4 due to the decrease of B 0 [29].

3 A. Sagara et al. / Fusion Engineering and Design xxx (2010) xxx xxx 3 increase the total TBR over 1.2 but also to reduce the radiation effects on magnets [36] HSR Fig. 2. Design history of FFHR. From the requirement of -heating efficiency over 0.9, the importance of the ergodic layers surrounding the last closed flux surface has been found by collision-less orbits simulation of 3.52 MeV alpha particles [28]. Therefore, to avoid the interference between the first walls and the ergodic layers at the inboard side in particular, the reactor size is increased with alternative options, and the design is improved as FFHR2m2, in which = 1.20 is selected with inward shifted magnetic axis. In this case, it is found that there is the optimum major radius of plasma around 16 m with B 0 of about 5 T by taking into account the neutron wall loading below 2 MW/m 2, cost analyses based on the ITER (2003) design [30] and engineering feasibility on large scaled magnets. The magnetic stored energy is reduced less than 150 GJ by selecting the location of poloidal coils, then it is about 3 times as large as ITER but the maximum magnetic field of 13 T and mechanical stress can be comparable. Minimization of the external heating power to access selfignition is advantageous to increase the reactor design margin. Because any fusion power rise-up time can be employed in a heliotron reactor, it has been recently found in a zero-dimensional simulation that a lower density limit margin reduces the external heating power, and over 300 s of the fusion power rise-up time can reduce the heating power from 100 MW to a minimized 30 MW in the FFHR2m1 [31]. Recent discovery of the super dense core (SDC) plasma up to m 3 in LHD [18] has led the new ignition scenario [32], in which the control of thermally unstable operation is crucial and a new and simple feedback method is proposed in FFHR2m1 using proportional-integration-derivative (PID) control of the fueling [33]. Such high density and low temperature operation is generally advantageous to reduce the divertor heat flux due to an enhanced radiation loss rate [34]. For continuously wound large superconducting magnet systems under the maximum nuclear heating of 200 W/m 3, cable-in conduit conductor (CICC) of 90 ka with Nb 3 Al are proposed with react and wind method and quench protection candidates and with the maximum cooling path of about 500 m [35] and a robust design of LHD-type cryogenic support posts ( 16,000 tonnes/30 posts). The blanket designs have been improved to obtain the total TBR over 1.05 for the standard design of Flibe + Be/JLF-1 and the STB blanket with the blanket cover rate over 90%, which is effectively possible by a new proposal of Discrete Pumping with Semi-closed Shield (DPSS) concept [28] and is very important not only to The concept to design the coils for magnetic configuration ensuring the best plasma performance offers the chance for the advanced stellarator optimization. In W7-AS this concept has been utilized to replace the helical windings of the previous experiment W 7-A. The concept of modular coils and the principle of optimization have been combined in the W7-X device, which will demonstrate the reactor capability of the advanced stellarator line. The HELIAS reactor (HSR) design studies are an extrapolation from the W7-X design to a reactor size stellarator [37,21]. In particular the chosen quasi-symmetry and the optimization procedure are the same, decoupling as much as possible plasma and magnetic field properties. Three different reactor configurations have been investigated: (i) the HSR5/22, a scaled up version of W7-X, with five field periods, a major radius of R = 22 m and an aspect ratio of A = 12, (ii) the HSR4/18 with four field periods, 18 m major radius and a reduced aspect ratio of A = 9, and finally the HSR3/15 with only three field periods, R = 15 m and an even further reduced aspect ration of A = 6. The top view of the modular coils for the different period HSR is shown in Fig. 3. As the minor radius increases with decreasing major radius, the plasma volume is similar for all three configurations. Also the average magnetic field (on axis about 5 T or slightly below which is twice that of W7-X), the aspired (4 5%) and hence the fusion power (3 GW) are comparable. Analysis of the power balance in HSR4 and HSR5 shows that (1) ignition can be achieved with an average beta 3.2% and (2) a fusion power of 3 GW can be reached in self-sustained operation close to the beta MHD limit, if the empirical scaling time (Lackner- Gottardi (LG) scaling) is used [38]. The LG scaling is used as a basis in designing the Helias reactor plasma, although in the high density discharges of W7-AS plasma confinement is better than predicted by LG. Extrapolation to HSR yields confinement times in a range of 1 3 s, and does not require H-mode improvement. The ISS04 scaling (which is essentially the ISS95 gyro-bohm type scaling) predicts the required confinement time for HSR with the amplification factor H Neoclassical transport in a Helias reactor is strongly suppressed. In HSR4/18 an effective helical ripple of considerably less than 1% over the entire plasma cross-section is expected [38]. It was shown [39], that stable fusion burn at high density operation in HSR4/18 can be sustained at some finite residual level of 1/ transport, ε h ε eff 2%. Analysis of the MHD stability has shown stability up to an average beta of 4.2% [38]. In HSR4/18 direct losses of fast alpha particles will be negligible, but the stochastic diffusion can lead to the energy lost fraction of 2.5%. This is tolerable with respect to the energy balance of the reactor; however the impact of the highly energetic alpha particles on the first wall could be a matter of concern. The blanket options envisaged in the Helias reactor are a helium cooled solid breeder blanket (HCPB) and a water-cooled Li Pb blanket (WCLL). In the present concept of HSR5/22 the area of first wall is 2600 m 2, which leads to an average neutron wall loading of less than 1 MW/m 2 (fusion power 3 GW). In HSR4/18 the area of the first wall is 2200 m 2, which results in a 20% higher average neutron wall load than in HSR5/22. The peaking factor is 1.7 compare with 1.5 in a tokamak DEMO reactor and the lifetime of the first wall and structural components is about 2 times larger [40] ARIES-CS The ARIES-CS reference plasma is a NCSX-like configuration with 3 field periods, has a major radius of 7.75 m and a plasma aspect ratio of 4.5. The configuration has excellent quasi-

4 4 A. Sagara et al. / Fusion Engineering and Design xxx (2010) xxx xxx Fig. 3. Top view of the modular coils for the different HELIAS reactor studies. The colors indicate repeating coils types. axisymmetry, as measured by the effective helical ripple, ε eff (ε eff < 0.6% at LCMS and 0.1 in the core region). The shaping of the plasma results in a vacuum rotational transform from 0.4 to 0.5. The plasma current (bootstrap current) is 4 MA which raises the rotational transform to 0.7 near the plasma edge. The distinct feature of this configuration is that a bias is introduced in mirror and helical terms of the magnetic spectrum in order to alter the structure of the ripple and improve particle confinement [25]. In this manner, the particle loss in ARIES-CS is reduced to 5%. While the heat flux from these lost particles may be handled on the divertor target plate, erosion and, in particular, exfoliation due the accumulation of He atoms in the armor are major concerns. Extensive optimization of the coils showed that a Ribbon-like coils (i.e., thin in radial direction and wide in toroidal/poloidal direction) allow a larger coil-plasma spacing. The ARIES-CS coils set consists of 18 coils (three different coil shapes) with a 0.19 m 0.74 m cross-section. The maximum field on the coil is 15 T (for B 0 = 5.7 T) necessitating the use of Nb 3 Sn superconductors. The required energy confinement time for ARIES-CS is 1 s and is a factor of H = 2 larger than the value predicted by the ISS95 scaling. ARIES-CS focused on operation with a high density because of the ñ e 0.51 dependence in ISSS95 scaling, increased a slowing-down time which reduced a losses, and also the belief that high density operation would help reducing the heat flux on the divertor plates. Because of the compact size, a large portion of plasma power should be radiated in order to obtain a reasonable heat flux on the divertors. A dual-coolant configuration with a self-cooled Pb 17Li zone and He-cooled reduced-activation ferritic steel (RAFS) structure was selected as reference concept [26]. The blanket is coupled to a Brayton cycle through a heat exchanger where both coolants (He and Pb 17Li) transfer their energy to the cycle working fluid (He). A modular concept was adapted for the ARIES-CS compact stellarator geometry compatible with the port-based maintenance scheme [26]. A key engineering parameter affecting the size of a compact stellarator is the minimum coil plasma distance. A novel approach has been developed where a highly efficient WC shield is used in the critical area to minimize this distance. In this way, the normal blanket/shield/manifold thickness of 1.76 m can be reduced to 1.3 m by optimizing the shield while the loss of breeding in these regions can be compensated by optimizing the breeding in other regions (to maintain the 3D TBR at 1.1) [27]. The coil system consisting of the inter-coil structure, coil cases, and winding packs, is enclosed in a common cryostat. The coils are wound into grooves at the inside of a strong supporting toroidal tube for each field period. These tubes are then connected to each other to provide a continuous ring structure to react the large centering forces pulling each coil towards the center of the torus. The out-of-plane forces acting between neighboring coils inside a field period are reacted by the inter-coil structure. Because of high-field on the coil, the superconductor is Nb 3 Sn, operating at 4K.Itis installed with a wind (including an inorganic insulator) and react method and heat treated [41]. 4. Prospect and main issues 4.1. FFHR The road map towards the helical-demo at 2030s consists of three pathways [30]; LHD experiments to accumulate enough amount of database to build a helical-demo, contribution of LHD research to physics of burning plasma in ITER, development of the LHD numerical test reactor with integration of physics and technology based on simulation science. One of important targets is the SDC (super-dense core) ignition plasma, in which the external heating methods such as the electron Bernstein wave (EBW) [42], fueling methods [33] and He ash removal [43] are important issues. Optimization of poloidal coiles arrangement and operation is important as coupling issues on maximizing plasma volume by minimizing the ergodic layer which works efficiently as impurity shielding, reducing the total stored magnetic energy by reducing leakage field and maximizing maintenance port areas with simplified cryogenic support structures. Divertor physics, high heat target and pumping engineering are key issues and will be soon examined in the closed divertor program in LHD [18]. Mockup studies of large size superconducting magnet system with quench protection, long-life blanket system under high temperature operations with low activation materials and high efficiency tritium and heat recovery system are crucial and common issues as in tokamaks HSR The Helias reactor, which demonstrates ignition (ignition experiment, HSR4/18i) has been considered [37]. The necessity to accommodate only a shield between coils and plasma determines the dimensions of the HSR4/18i reactor. The size of reactor is large enough to allow for a self-sustained burn within presently known laws of confinement. In HSR4/18i, a magnetic field can be reduced to 8.5 T on the coils. The modular coils for HSR4/18i are shown in Fig. 4. The divertor concept in the Helias reactor is the island divertor concept, which in HSR4/18 uses the existence in a separatrix region of the 4/4 islands. This concept was successfully verified in W7- AS experiment. However, a heat load on the target plates in HSR, like in any reactor, could be a critical issue. To keep thermal load below the technical limits of 5 10 MW/m 2, up to 90% of the alphaparticle power must be radiated. This can be achieved in partly

5 A. Sagara et al. / Fusion Engineering and Design xxx (2010) xxx xxx 5 a high degree of precision in manufacturing, assembly and alignment, as well as lead to a reduction in the performance due to the large peaking factors and non-uniformities. In principle, a substantially larger (at least 50%) plasma coil spacing is probably necessary compared to similar components in a tokamak. ARIES-CS study also showed that device configuration, assembly, and maintenance drive the design optimization and technology choices in many cases. As such, an integrated optimization of physics performance as well as technology feasibility is an essential element of stellarator development. Development of high-field superconducting magnets with irregular shape (e.g., inorganic insulators and/or high temperature superconductors) as well as development and demonstration of methods to fabricate, assemble, and maintain large superconducting stellarators free of resonance-inducing field errors are critical issues. Finally, it should be stressed that fusion power technology and materials are in their early stages of development and an extensive R&D program in these areas is urgently needed. Fig. 4. Modular coils of HSR4/18: 10 coils per period (5 coils with different shape); plasma also has shown in red. (For interpretation of the references to color in this figure caption, the reader is referred to the web version of the article.) detached operation with the high value of separatrix density. The divertor plates are aligned along the iota equal one islands as it is shown in Fig. 5. Ten longitudinal divertor modules cover 50% of the torus circumference and provide about 400 m 2 area of target and baffle plates (40 m 2 wetted area). Every divertor plate is 15 m long and 1 m wide. For the reason of maintenance, segmentation of the target plate is foreseen ARIES-CS ARIES-CS study showed the configuration space for QAS stellarators is quite rich and many desirable configurations are possible. An important constraint is optimizing a configuration is the plasma limit which is not understood in stellarators and is an important physics topic which should be addressed in the next generation of stellarators. Not only would relaxing linear MHD stability constraint allow optimization of other parameters (e.g., -particle losses), a larger ˇ value can also be used to reduce the technological requirement of the superconducting magnets. While -particle loss does not impact the plasma power balance, it leads to serious issues for in-vessel components (localized heating, exfoliation of blistering). Of course, QAS configuration with reduced a losses should be developed and demonstrated experimentally together demonstration of profile control (including bootstrap current effects) as well as development and demonstration of pumped divertor configurations with a highly radiative plasma without any impurity accumulation. ARIES-CS study has also highlighted the difficulties associated with the 3D and complex shapes of the fusion core. These lead to Fig. 5. Divertor plates in HSR4/18 configuration. 5. Summary and conclusions The first review is presented for the recent stellarator designs of FFHR, HSR, ARIES-CS, which are based on the experimental device projects of LHD, W-7X, NCSX, respectively. In the series of FFHR-1, 2, 2m1 and 2m2 designs, the coil pitch parameter of continuous helical winding and the major radius R with poloidal coil positions have been optimized to reduce the magnetic stored energy below 150 GJ while expanding the blanket space, and a self-cooled Flibe blanket has been proposed as a long-life blanket under neutron wall loading less than 2 MW/m 2. Recently the high density ignition operation has been proposed with drastically reducing divertor heat load by newly proposing a thermal instability control method. Large size superconducting helical coil is shown conceptually feasible. Divertor design, external heating design and unscheduled blanket replacement are engineering key issues. The Helias reactor can reach ignition and run in stable operation within the physics limits presently known from stellarator experiments. The HSR4/18 is the most appropriate configuration, which demonstrates the best reactor features. Extrapolation predicts stable operation for ˇ 3.5 4%, B =5T, E 1 2 s and fusion power 1 3 GW. The sizes of superconductive NbTi modular coils are large enough to accommodate blanket and to have sufficient ports for maintenance. The appropriate support structure is designed with mechanical stress within technical limits. An average neutron power density in the blanket is estimated roughly 1 MW/m 2. The lifetime of blanket system can exceed that for tokamak reactor ( 2 times). Natural islands at the edge can be utilized for the divertor function. The future perspectives for Helas reactor mainly depend on the achievements on W7-X experiment. ARIES-CS study has explored compact stellarator configurations. By focusing on quasi-axisymmetric (QAS) configurations, it was managed to simultaneously reduce the machine size and -particle losses while maintaining desirable stellarator properties. 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