“Development of 3-D divertor solutions for stellarators through coordinated domestic and international research”
Oliver Schmitz1 on behalf of U.S. stellarator collaborators*
U.S. stellarator collaborators: * J.P. Allain3, D.T. Anderson1, A.H. Boozer2, D. Curreli3, J.P. Freidberg4, D.A. Gates5, J.H. Harris6, S.R. Hudson5, M. Landreman7, J.D. Lore6, G.H. Neilson5, D.A. Maurer8, A. Reiman5, D. Ruzic3, D.A. Spong6,J.N. Talmadge1, F.A. Volpe2, H. Weitzner9, G.A. Wurden10, M.C. Zarnstorff5
1 – University of Wisconsin 6 – Oak Ridge National Laboratory 2 – Columbia University 7 – University of Maryland 3 – University of Illinois 8 – Auburn University 4 – Massachusetts Institute of Technology 9 – New York University 5 – Princeton Plasma Physics Laboratory 10 – Los Alamos National Laboratory
I.) Introduction: This paper, a supplement to the stellarator community white paper and presentations, describes a set of three initiatives aiming on development of a 3-D divertor solution for stellarators within the next decade. Development of a 3-D divertor solution for Long Pulse Burning Plasma science is a key challenge on the way to qualify the stellarator as a potential candidate for a demonstration fusion reactor.. Optimization of 3-D confinement and 3- D equilibrium properties in stellarators go hand-in-hand with the development of a viable 3-D divertor solution in an integrated approach for confinement of stationary, disruption free, high performance plasmas. A coherent initiative is proposed to enhance international stellarator collaboration towards this mission goal and link it to a thriving, well-targeted domestic research program. Our aim is to expand on our strong standing in the field and take up leadership on an international scale.
Key issues which have to be addressed for the development of a credible 3-D divertor solution are (a) a divertor structure that is robust to finite beta and bootstrap current effects on the equilibrium (b) the ability to handle the heat exhaust through a combination of radiation with detachment (partial or full) and expanded target surface area (c) capability to reduce neutral recycling and impurity influx from the wall to the plasma edge and (d) compatibility to core transport performance which is intrinsically bound in stellarators to the 3-D equilibrium choice (optimization route). The issue of a robust divertor with respect to stability of the magnetic configuration concerns the residual plasma currents existing even in optimized stellarators. These currents change the rotational transform and hence the edge magnetic field structure employed as divertor. This has consequences for the heat and particle loads of divertor and first wall elements as well as on the particle exhaust properties. Today two concepts are used - the helical divertor at the Large Helical Device in Japan and the island divertor at Wendelstein 7-X in Germany. Both are quite different concepts, but both require a stable edge magnetic topology to work as feasible 3D divertor concepts. There are significant concerns as far as the internal plasma response (in particular, the residual bootstrap and Pfirsch-Schlüter currents) interacts with the divertor structures. The edge magnetic field defines the neutral and impurity transport as well as the particle and heat exhaust characteristics. Hence, there is a direct link established between the internal plasma currents and the plasma edge transport and plasma wall interaction. Reducing and controlling the level of internal currents is a means to control the plasma edge features. At the same time, the transport across good flux surfaces in the plasma defines the particle and heat inflow into the plasma boundary and hence the resulting exhaust requirement. This points out that development of a 3-D divertor solution in stellarators requires taking plasma transport inside of the last closed flux surface into account. The edge structure will determine the inward transport of neutrals and impurities and hence – in an ion root transport regime (i.e. a negative radial electric field) – also the level of impurity accumulation and helium retention in the core. Consequently, core transport phenomena are directly linked to the optimization of the edge magnetic field structure. This sequence of key issues of our development strategy for a 3-D divertor solution shows that we seek to setup an integrated approach, which concentrates on the 1 3-D divertor structure and related edge transport and plasma wall interaction. However, with a suite of diagnostics and modeling tools we will assess at the same time how feasible a given 3-D divertor concept is as an integrated component of a long pulse plasma including plasma core transport and performance aspects. The goals and means of these initiatives are shown in figure 1. The goal of these initiatives is to develop a credible 3-D divertor solution for stellarators through a coordinated research effort internationally and on domestic experiments. The initiatives are described in the following section II with the three budget scenarios stated in the charge to FESAC discussed at the end of each initiative. Concluding remarks addressing prioritization are described in section III.
Figure 1: Goals and means of initiative on 3D divertor physics in stellarators
II.) Initiatives
Initiative I: Enhance international stellarator collaboration enabling access and leadership participation on forefront, large-scale stellarator facilities
Two large-scale superconducting devices constitute the experimental basis of the international stellarator collaboration program - the LHD stellarator in Japan and the Wendelstein 7-X (W7- X) stellarator in Germany. Both devices feature quasi-stationary plasmas with specific divertor concepts which have to be qualified as viable divertor solutions.
At LHD, recently a closed helical divertor has been installed. The divertor structure consists of a stochastic magnetic field layer in the very edge and long divertor legs in which the magnetic field lines approach the divertor surfaces under very shallow angles. This first of its kind setup allows the testing of the quasi-stationary particle and heat exhaust properties. This setup is in strong contrast to the W7-X island divertor as shown in figure 2.
Figure 2: Plasma edge structure of the helical divertor at LHD, NIFS, Japan on the left and the modular island divertor structure at Wendelstein 7-X, MPI-IPP, Germany on the right.
2 Here, magnetic islands in the plasma edge are used to form the divertor structure. Comparing both divertor geometries and qualifying them towards heat and particle exhaust properties as well as impurity retention and density control is a unique opportunity with W7-X coming online at the same time LHD is operating as a mature large scale stellarator device. The LHD directorate has endorsed participation of U.S. scientists and recently intensified the encouragement to participate in LHD research. We have already provided a crystal soft x-ray imaging spectrometer and a strong collaboration path exists. The U.S. XICS system provides an available opportunity for the U.S. to participate in their exciting plans for the next several years, including their new divertor configuration and long-awaited deuterium plasma operation. On W7-X, we are particularly interested in 3-D edge plasma/divertor physics as a key aspect of an integrated physics program including 3-D core transport and plasma control in long pulse operation. We have provided trim coils, which are an actuator for fine-tuning the edge magnetic field structure and hence are a key tool to systematically change divertor topology. The ongoing design of an active protection element, or so-called scraper element, for critical high heat flux divertor regions has been driven by U.S. involvement and in concert with the trim coils and current US diagnostic activities on W7-X provides for an excellent basis for a U.S. leadership role in establishing the island divertor physics basis for successful W7-X operations. Moreover, this particular activity aids generic diagnostic and hardware developments for other long pulse devices like surface heat load control by IR spectroscopy and pellet fueling at ITER. A recent workshop on possible university contributions to W7-X has highlighted, that a very strong alignment exists between W7-X programmatic needs and U.S. University and national laboratory activities and expertise. A strong national lab programmatic setup is already in place and the Wendelstein 7-X directorate has strongly endorsed the U.S. collaboration in particular in the field of 3-D divertor physics as key to the success of the device. The U.S. research agenda for W7-X will make use of its contributions to date and will focus strongly on edge and divertor physics in its first years. The TDU scraper element, with its embedded diagnostics combined with U.S. infrared imaging instruments and U.S. modeling efforts, will enable measurements of edge plasma transport coefficients. Critical to these studies is the ability to control the edge magnetic field structure, including the use of U.S. trim coils to correct for error fields. The U.S. has the edge physics and magnetics modeling tools needed for these studies, which will provide in addition opportunity for validation. Investigation of the edge cannot be decoupled from understanding of the core and the U.S. will use its XICS to contribute with temperature and velocity profile measurements of the core. Over time, U.S. core physics research on W7-X will grow; plans include density profile control with pellet fueling, energetic particle physics studies, and transport physics studies with fluctuation diagnostics. U.S. participation in both the LHD and W7X world-class stellarator programs will enable advances in understanding through comparison based on their similarities and differences, as well as model validation against data from multiple sources. The planned U.S. activities on both devices have been aligned with the programmatic interests of the local scientific team and other international collaborators. Exploitation of the strong positioning of the U.S. stellarator program and of the significant investments made requires enhanced funding for international stellarator research. The program proposed shall focus initially on 3-D divertor physics, including measurements of transport and equilibrium effects in the plasma.
Initiative 1: Enhance international stellarator collaboration enabling access and leadership participation on forefront, large-scale stellarator facilities
Present funding status: $2.5 M/yr (National Laboratories) and $0.5 M/yr (Non-lab). Scenario 1 (modest growth): Expand to $5 M/yr immediately and grow to $10-12 M/yr over the next five years Scenario 2 (cost of living): Expand to $5 M/yr over next five years in $1.0 M/yr increments Scenario 3 (flat funding): Expand to $5 M/yr over next ten years with $0.5 M/yr increase 3 Initiative II: Form a coherent domestic program by well targeted upgrades to U.S. devices focusing on fundamental physics understanding and code validation for a 3-D divertor solution.
A well-aligned domestic initiative is proposed that addresses the gaps in critical issues identified within the international research proposal component. Existing domestic facilities represent a well-suited platform to address development of the tools required to address 3-D divertor physics behavior in a coherent fashion. Enhancement of the existing domestic facilities is proposed targeting development of new concepts to overcome the issue of topological stability - like that addressed with the trim coils at W7-X. A 3-D divertor solution should be qualified alongside with good neoclassical confinement at reduced plasma current levels as pioneered by the U.S. stellarator program in particular towards impurity confinement and transport. Optimization of the divertor, while minimizing anomalous transport, is a key aspect in an integral core-edge-divertor solution for stellarators. Exploring the effect of 3-D shaping on plasma stability in conjunction with the impact on optimized 3-D topologies is another key goal. The U.S. fields a unique set of small scale stellarator devices of appropriate parameter range to address these edge physics issues. In proposing this initiative, we have identified the strength of each device to address a specific set of key physics questions. Each device has proposed an upgrade strategy to make use of existing hardware and focus on specific divertor physics issues. The mission of each device is stated in the corresponding headline in figure 3 and a list of the relevant upgrades is presented. In the following, the proposed strategy for each device is presented including a budget scenario to conduct this upgrade. An overall budget summary of this initiative is presented at the end of the document.
Figure 3: Coherent setup of U.S. stellarator facilities to address key aspects for development of a 3D divertor solution for stellarators.
For each device, a brief statement of the role within the coherent U.S. domestic program is included. Then a mission goal, which is enabled by the upgrade, is stated including upgrade goals and means. For each machine a layout of the upgrade budget is discussed.
• Helical Symmetric Experiment (HSX), University of Wisconsin - Madison HSX demonstrated for the first time that a quasisymmetric magnetic field structure can be designed and built to reduce neoclassical losses and optimize plasma confinement. With this background the device and the HSX team are ideally suited to act as a test bed to develop an integrated 3D divertor solution together with confinement of a plasma with significant ion temperature and hence “ion-root” confinement and transport. A key aspect is to advance the optimization of neoclassical confinement achieved to tackle the challenge of optimization towards turbulent transport. An initiative is underway to assess 4 if the field structure of HSX can be modified to maintain good neoclassical confinement and at the same time optimize for reduced turbulent losses. Addressing this experimentally requires a plasma which provides drives for electron and ion turbulent modes. Hence providing substantial ion heating is a key aspect for this endeavor.
Mission enabled by upgrade: Test of new 3D divertor concepts and impurity transport in integrated fashion with optimized neoclassical and turbulent confinement in ion root confinement regime
Upgrade goals: - Increase ion temperature (through NBI) to: • Study role of high-effective transform (ieff ~3) on NBI deposition and ion confinement in moderate ion and electron temperature plasmas (500 eV at
- Achieve higher density operation to: • Increase divertor plasma parameters and compare to model calculations • Study impurity transport and accumulation • Reduce charge exchange losses and increase operational flexibility • Realize a relevant core plasma, i.e. largely opaque to neutral refueling together with ion root transport and ion driven turbulence and test in integrated manner with innovative divertor
Upgrade means: - Upgrade Electron Cyclotron Heating (ECH) system: 500 kW, 0.3s, 56 GHz (present system is at 28 GHz) to provide feasible target for neutral beam and study energetic electron transport. This upgrade will also largely increase reliability of HSX operation. The ECH systems envisioned are off shelf components and do not need development work. - Implement two Neutral Beam Injectors (NBI): 2 x 300 kW, 40 keV, 0.3s to increase ion temperature and go to ion root transport. Beams envisioned are off shelf beam components (Budker Inst. Compass-D units) and do not need development. - Design, build and implement new type 1 coil. This coil type is used at the four corners of HSX and will be redesigned to make room for a test divertor and tangential beam injection. An activity to minimize the impact of the coil design on the QH optimization of HSX has been started in collaboration with PPPL. - New vacuum vessel to give room for test divertor, increase distance between plasma and wall for a more defined plasma edge region, and provide additional diagnostic access and vessel baking capability.
Cost breakdown of major elements: • Major element cost breakdown (over 4 year period) ($k) – 2 X 300kW 0.3s Beams (w/PS-torus interface) 2400 – 500 kW 0.3s ECH (w/PS, transmission line, launcher) 1300 – Vacuum vessel/ divertor chamber/structure 2800 – New coil-type #1’s 1100 – Infrastructure and contract engineering 2400 – Contingency (@20%) 2000 Total cost 12000
• Site credits reduce cost of upgrade
5 – Only 16% of coils remanufactured/ majority support and infrastructure (DAQ/Control, etc) used – Existing 54 MJ motor/generator supply for magnets – Utilize existing diagnostics (Thom. Scat., ECE, CXRL, reflectometery, spectroscopy, etc.)
Time line of funding
A staged approach could be employed with postponing one of the NBI units and the new ECRH system (to future upgrades) with a capital cost reduction from $12.0M to $8.4M. This would preserve the major shift to warm ions, ion-root discharges, and divertor studies. Lost program elements and capabilities could then be restored in future upgrades, although at perhaps greater total cost.
• Compact Toroidal Hybrid (CTH), Auburn University The CTH device is unique in the world in its ability to span the operational space from a stellarator to a tokamak/stellarator hybrid modified by significant amounts of internal ohmic plasma current. This magnetic configuration flexibility allows it to make connection to disruptive phenomena observed on tokamak experiments and explore the effects of strong non-axisymmetric shaping on plasma stability boundaries and avoidance of the disruptions they induce in tokamaks. CTH has demonstrated stabilization of vertical motion, operation above the tokamak density limit, and ultra-low edge safety factor discharges beyond traditional tokamak operational boundaries using strong 3-D shaping provided by stellarator fields. The CTH program has plans to broaden its activities with a new 3-D island divertor research thrust to address the critical need for basic plasma physics understanding of the 3-D edge plasma. These physics issues can be addressed on CTH in a regime complementary to that of the upgraded HSX device with divertor. The proposed CTH island divertor is generated by symmetry breaking coils that are similar to ones used for ELM mitigation on present day tokamaks and are seen as a critical component for successful ITER operation. This proposal allows the combined CTH/HSX effort to explore 3-D divertor physics relevant to 3-D perturbed tokamaks, current carrying stellarators like QUASAR, and optimized stellarators like W7-X, while furthering their core missions in 3-D transport and MHD stability and disruption physics.
Mission enabled by upgrade: Test stability limits when going from tokamak to stellarator plasma equilibria in integrated fashion with understanding the 3-D plasma edge.
Upgrade goals: - Study transition from 2-D plasma edge to 3-D edge topology including divertor structure at relevant SOL temperatures and densities - Tokamak operation allows new disruption physics studies: - Well-known disruption phenomenology, then add rot. transform - Allows comparison of disruption loss of confinement dynamics with and w/o external transform (cage effect) - Explores B3D/B0 down to similar levels used to optimize current tokamaks - Auxiliary heating using NBI to: - Explore density limit physics using auxiliary heating with rot. transform - De-couple tokamak (Greenwald) and stellarator (Sudo) density limits 6 - Understand transition from tokamak to stellarator-like limit
Upgrade means: - ECH: 200 kW, 28 GHz - Vertical/ohmic PS upgrades to allow tokamak operation/control - NBI: 300 kW 40 keV 0.3s systems (Budkar Inst./ Compass D units) - Infrastructure upgrades to accommodate new capabilities
Cost breakdown of major elements: • Major element cost breakdown (over 3 year period) ($k) – 200 kW, 28GHz ECRH (w/PS, transmission line, launcher) 1000 – 300kW Beam (w/PS-torus interface) 1200 – Vertical/Ohmic PS upgrade for tokamak operation 750 – Vacuum vessel/ divertor chamber/structure 100 – Diagnostic suite upgrades (island/disruption studies) 500 – Infrastructure and contract engineering 250 – Contingency (@20%) 760 Total cost 4560
• Site credits reduce cost of upgrade – Infrastructure (DAQ/Control, etc) used – Existing motor/generator supply for magnets and capacitor banks used
Time line of funding
• Compact Non-neutral Torus (CNT), New York CNT is an amazingly simple realization of a stellarator plasma equilibrium. One of its unique features is the very long magnetic field connection length. An anticipated advantage of the stellarator over the tokamak is that a much longer field-line connection length radially spreads the heat flux in the divertor region. CNT provides an ideal, well- targeted test bed for this concept. The moderate upgrade costs needed for this will allow the operation of the device with significant heating capabilities to assess heat and particle flux spreading as a function of the magnetic field line connection lengths. The mission is favored by excellent diagnostic access, a result of the simple magnetic coil setup and large vessel.
Mission enabled by upgrade: Explore long connection length regimes for heat flux spreading
Upgrade goals: - Use access to test/optimize different island divertors, move them relative to plasma and diagnose by Langmuir probes and IR imaging. - Confirm and characterize more benign wetted area scaling in stellarators (long connection length regime)
Upgrade means: - Implement transmission line and antenna to make use of existing 4 MW, 1-3 MHz, 10 ms source for Ion Cyclotron, Lower Hybrid or sub-harmonic sub- thermal Alfven Wave heating and current drive.
7 - Implement transmission line, antenna, PS and infrastructure for klystrons to be borrowed in the 2.45 - 8 GHz range, for Electron Cyclotron heating and current drive. Discussions are under way with various European and US institutions. - Implement diagnostics for dedicated measurements tasks.
Cost breakdown of major elements: • Major upfront costs ($k) – Transmission line and antenna for existing 4 MW source 1-3 MHz, 10 ms source 150 – Transmission line, antenna, PS and infrastructure for klystrons 250 Total cost 400
• Site credits reduce cost of upgrade Existing MW-level flywheel generator, capacitor-banks, RF PSs and triodes will reduce cost of upgrade.
Time line of funding
• Hybrid Illinois plasma Device for Research and Applications (HIDRA)
HIDRA is a new contributor to the U.S. domestic stellarator program and it addresses an open gap, i.e. a focused program on plasma materials issues in 3D geometries. The device is based on the WEGA stellarator which was provided from MPI-IPP Greifswald, another testimony of their strong interest and support of a close collaboration between the U.S. and Wendelstein 7-X. The actual program is forming and the definition of funding needs is ongoing. HIDRA has five core thrusts, which align in an efficient way with the activities on the other stellarator devices, described beforehand.
Upgrade goals:
- Develop a toroidal PFC test facility aimed at liquid metals and enabling technology development. The ability to run as a tokamak as well as a stellarator will provide a test bed for fully axisymmetric PFC experiments. This will cover approaches to create workable divertors for the stellarator community as well as tests for high-powered liquid metal solutions.
- Develop a materials synthesis facility. HIDRA will be providing a test bed for advanced material testing, processing, and development of in-situ diagnostics that can measure in the time- scale of plasma-material interactions how material surfaces respond to the fusion plasma. An advanced multi-user facility for innovative materials testing and processing will also be designed and integrated in HIDRA, which will serve the general Illinois and U.S. scientific community
- Become a test bed for model validation, with specific emphasis on near-wall transport models and divertor material response.
- Create a platform for plasma diagnostic development in the 3-D divertor.
Upgrade Means • Move the WEGA stellarator/tokamak from MPI/IPP Greifswald to UIUC • Establish the new HIDRA facility at Illinois with internal labor at UIUC • Utilize ECRH power >200 kW for adequate divertor parameters • Develop in-situ diagnostics for plasma-material interaction 8 • Establish data acquisition & data management infrastructure • Develop innovative diagnostics for relevant plasma parameters
Mission enabled: Develop innovative divertor material concepts for 3-D systems
The total cost estimated is 300 k$ per year.
Initiative 2: Form a coherent domestic program by well-targeted upgrades to the U.S. devices focusing on dedicated challenges to develop a 3D divertor solution
Scenario 1 (modest growth): Fund updates as suggested including contingencies Scenario 2 (cost of living): Fund upgrades to devices on longer time scale Scenario 3 (flat funding): Fund devices with budgets as is and let devices decide if they use operation budget for upgrades
Initiative III: Develop and validate predictive modeling capabilities for 3D divertor solutions.
In addition to the coordinated experimental program, a vibrant theory program is required to link the results from various devices together and generate a strong predictive capability towards next step experiments and towards a stellarator based reactor scale experiment. Within this white paper we consider theory efforts relevant to 3-D divertor physics. M. Landreman discusses the overall theory initiative for stellarators in the white paper and talk. The EMC3-Eirene fluid plasma edge transport and kinetic neutral transport code is the only available 3D model which can address the interaction with material surfaces and the plasma edge in a sophisticated way. This model was developed for the Wendelstein 7-AS stellarator and is presently being used on a large variety of devices, including stellarators, spherical tokamaks and tokamaks with external perturbation fields. The EMC3-Eirene code is the principal workhorse for 3-D edge and PMI modeling studies. However, there are clear shortcomings in the model, which distinguish its present level of maturity from that of 2-D edge codes that are being used to design the ITER divertor, for instance. In particular, particle drifts and kinetic corrections to parallel heat conduction are standard implementations in state of the art 2-D divertor modeling tools. EMC3-Eirene has to be enhanced and – most importantly – validated against suitable experimental data. The experimental efforts on both international and domestic devices are ideally suited to provide data across a broad spectrum of magnetic configurations (isodynamic, quasi-symmetric, and 3D symmetry breaking islands) to test model physics enhancements and validate them. The combined effort of the proposed experimental work and numerical code activity will enable the model to advance to the level of 2-D edge codes and beyond. This goal is critically needed to extrapolate findings on present devices to ensure code integrity for future devices and reactor 3D divertor designs.This development also has a direct impact on 3-D boundary plasmas in tokamaks, and in particular ITER. It has been shown that a 3-D plasma boundary is induced when resonant magnetic perturbation fields are applied for control of edge localized modes on these nominally 2-D devices. EMC3-Eirene is used intensively to assess the impact of this transition of the tokamak boundary from a 2-D to a 3-D edge on the actual divertor operation and long-term integrity. The effort proposed as third initiative therefore makes a direct contribution to this important field of inquiry on tokamaks.
Mission statement: Develop and validate predictive modeling capabilities for 3-D divertor solutions
Means: - Model enhancements and combinations: Enhance the EMC3-Eirene model and consider application of kinetic models (for instance XCG0). Include finite pressure effects through state of the art MHD modeling tools (VMEC). Determine need for development of a new 3-D modeling tool (e.g. combining kinetic and fluid modeling and including MHD physics.)
9 The near term goal is to enhance EMC3-Eirene towards inclusion of particle drifts, kinetic effects and possible multi-fluid impurity tracing. The Eirene part contains the neutral physics and dedicated enhancements (implementations of renewed collisional radiative models) is also a major part of the proposed code advancement.
- Validation: Validation of the existing model and during the development steps is a key part of the strategy. While the model is capable of reproducing general trends on various stellarator and 3D tokamak devices, it lacks a thorough validation of quantitative agreement. The experimental efforts proposed together with the active research programs and innovative diagnostics suggested for W7-X and LHD will provide an excellent test bed for the model and drive model validation and model enhancement based upon the experimental input.
Initiative 3: Develop and validate predictive modeling capabilities for 3-D divertor solutions
Present funding status: no dedicated funding for edge & divertor modeling Scenario 1 (modest growth): Increase to $0.5M/yr for 3D divertor / edge tool development (2-3 persons dedicated to this topic) Scenario 2 (cost of living): Fund at least two persons to conduct validation and some development Scenario 3 (flat funding): Fund at least one person to conduct validation and explore extension capabilities This funding request is part of the overall theory initiative presented by M. Landreman
III.) Concluding remarks.
The set of initiatives suggested in this white paper represents a coordinated effort to leverage the investments and efforts taken in international stellarator research and align those with a well- targeted domestic program. The proposed focus is on the development of 3-D divertor solution for stellarators. The enhancements to domestic facilities will enable this proposal to address key aspects within this line of research in a multi-machine domestic and international effort across a broad range of 3-D magnetic configurations. Aligning these domestic efforts with the large-scale international devices, W7-X and LHD, assures that the domestic university program is focused on supporting the success of the stellarator line as a viable alternative to the tokamak. Steady state and disruption free operation are features, which make the stellarator specifically attractive. However, performance and availability of a feasible exhaust system are the present advantages of tokamaks. To advance the stellarator concept to this level, good confinement properties have to be demonstrated in an integral way with a feasible exhaust device.
Prioritization: Setting priorities among the initiatives proposed is very difficult because they represent a coherent program, which guarantees the best outcome if conducted as proposed. The cost of each activity is comparatively moderate. The same is true for the suggested funding at the leading international facilities LHD and W7-X. A yearly amount of $10-12 M/yr to do cutting edge research at the forefront of the international community is a small price given that the host countries Japan and Germany provide the major capital investment. In this regard, a prioritization actually would degrade the key philosophy of the set of initiatives. However, if only funding can be provided for a fraction of the proposal, we give priority to enhancement of the collaboration at the overseas devices on a level which continues the operation of the domestic devices discussed beforehand and which enables moderate enhancements towards the goals of the initiatives discussed. Again, considering a yearly expense of $10-12 M/yr for international and up to $8 M/yr for the domestic program is a comparatively small amount and leverages substantial previous investments by the U.S. and other countries. The return on investment in terms of research capability gained from a total funding level of $18-20 M/yr is large given that others have made the major investments in capital cost (international) and the devices are established and ready to go with comprehensive and knowledgeable science teams which have world-wide credibility and standing.
10