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Liquid metal (LM) plasma-facing components (PFCs) may provide a resolution to the challenging fusion environment, particularly the first wall and divertor surfaces. Transforming these conce...

The Fusion Nuclear Science Facility (FNSF) is examined here as part of a two step program from ITER to commercial power plants. This first step is considered mandatory to establish the materials and component database in the real fusion in-service environment before proceeding to larger electricity producing facilities. The FNSF can be shown to make tremendous advances beyond ITER, toward a power plant, particularly in plasma duration and fusion nuclear environment. A moderate FNSF is studied in detail, which does not generate net electricity, but does reach the power plant blanket operating temperatures. The full poloidal Dual Coolant Lead Lithium (DCLL) blanket is chosen, with alternates being the Helium Cooled Lead Lithium (HCLL) and Helium Cooled Ceramic Breeder/Pebble Bed (HCCB/PB). Several power plant relevant choices are made in order to follow the philosophy of targeted technologies. Any fusion core component must be qualified by fusion relevant neutron testing and highly integrated non-nuclear testing before it can be installed on the FNSF in order to avoid the high probability of constant failures in a plasma-vacuum system. A range of missions for the FNSF, or any fusion nuclear facility on the path toward fusion power plants, are established and characterized by several metrics. A conservative physics strategy is pursued to accommodate the transition to ultra-long plasma pulses, and parameters are chosen to represent the power plant regime to the extent possible. An operating space is identified, and from this, one point is chosen for further detailed analysis, with R = 4.8 m, a = 1.2 m, I P = 7.9 MA, B T = 7.5 T, β N Gr = 0.9, f BS = 0.52, q 95 = 6.0, H 98 ∼1.0, and Q = 4.0. The operating space is shown to be robust to parameter variations. A program is established for the FNSF to show how the missions for the facility are met, with a He/H, a DD and 5 DT phases. The facility requires ∼25 years to complete its DT operation, including 7.8 years of neutron production, and the remaining spent on inspections and maintenance. The DD phase is critical to establish the ultra-long plasma pulse lengths. The blanket testing strategy is examined, and shows that many sectors have penetrations for heating and current drive (H/CD), diagnostics, or Test Blanket Modules (TBMs). The hot cell is a critical facility element in order for the FNSF to perform its function of developing the in-service material and component database. The pre-FNSF R&D is laid out in terms of priority topics, with the FNSF phases driving the time-lines for R&D completion. A series of detailed technical assessments of the FNSF operating point are reported in this issue, showing the credibility of such a step, and more detailed emphasis on R&D items to pursue. These include nuclear analysis, thermo-mechanics and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations.

The design of the water-cooled ceramic breeder (WCCB) blanket includes beryllium multiplier layers. At elevated temperature, beryllium reacts with steam in an exothermic reaction producing beryllium oxide and hydrogen. Such situation may occur in WCCB in the case of the rupture of one of the cooling pipes in the blanket module. This process occurs locally in a complex 3-D geometry of the blanket containing several different granular levels and a network of cooling pipes and structural supports. The process is also inherently unsteady since reaction rate depends on concentration of steam and pure beryllium which changes in time. In order to perform the detailed analysis of the process, the model of the reacting flowthrough porous media was developed and introduced into 3-D computational fluid dynamics code. In this model, granular beds are introduced as porous solids simplifying the model geometry and reducing typical mesh size to manageable amount of tens of millions of elements. A reaction rate between solid beryllium and steam is obtained from experimental results, and depends on temperature and concentration of the reactants. The differential equation for beryllium oxide fraction is introduced, allowing obtaining distributions of beryllium oxide in space and time. Multicomponent flow consisting of a homogenous mixture of steam and hydrogen is considered flowing through the porous solid with variable properties. Sink and source terms for steam and hydrogen fractions are determined by local beryllium oxide mass fraction source according to the molar ratios of beryllium steam reaction. The conjugated heat transfer approach is applied to calculate heat transfer in support structures as well as coolant flow, simultaneously with the porous medium steam flow in a blanket’s granular beds. The model is validated using experimental data on beryllium steam reaction for granular bed samples.

In order to get first insight into net tungsten erosion in W7-X, tungsten (W) marker layers were exposed during the operational phase OP 1.2b at one position of the Test Divertor Unit (TDU), at 21 different positions of the inner heat shield, and at two scraper elements. The maximum tungsten erosion rate at the TDU strike line was 0.13 nm s−1 averaged over the whole campaign. The erosion was inhomogeneous on a microscopic scale, with higher erosion on ridges of the rough surface inclined towards the plasma and deposition of hydrocarbon layers in the recessed areas of the rough surface. The W erosion at the inner heat shield was below the detection limit of 3–6 × 1012 W-atoms/cm2s, and all inner heat shield tiles were covered with a thin B/C/O layer with thickness in the range 2 × 1017–1018 B + C atoms/cm2 (about 20–100 nm B + C). W-erosion of the marker layers on the scraper elements was also below the detection limit.

Final design studies in preparation for manufacturing have been performed for functional components of the vacuum portion of the ITER Low-Field Side Reflectometer (LFSR). These components consist of an antenna array, electron cyclotron heating (ECH) protection mirrors, phase calibration mirrors, and vacuum windows. Evaluation of these components was conducted at the LFSR test facility and DIII-D. The antenna array consists of six corrugated-waveguide antennas for simultaneous profile, fluctuation, and Doppler measurements. A diffraction grating, incorporated into the plasma-facing miter bend, provides protection of sensitive components from stray ECH at 170 GHz. For in situ phase calibration of the LFSR profile reflectometer, an embossed mirror is incorporated into the adjacent miter bend. Measurements of the radiated beam profile indicate that these components have a small, acceptable effect on mode conversion and beam quality. Baseline transmission characteristics of the dual-disk vacuum window are obtained and are used to guide ongoing developments. Preliminary simulations indicate that a surface-relief structure on the window surfaces can greatly improve transmission. The workability of real-time phase measurements was demonstrated on the DIII-D profile reflectometer. The new automated real-time analysis agrees well with the standard post-processing routine.

With this Special Issue, we are publishing 81 peer-reviewed manuscripts, a selection from the best work presented in June 2019 at the 28th IEEE Symposium on Fusion Engineering (SOFE), Jacksonville, FL, USA. This is the sixth SOFE Special Issue that IEEE TRANSACTIONS ON PLASMA SCIENCE (TPS) has published, a tradition that started with the 2009 SOFE in order to afford attendees the opportunity to publish an expanded version of their oral or poster presentations in a refereed journal. It has become a popular option. With each successive SOFE, more authors take advantage of the possibility offered by TPS to have their conference papers gain wider exposure and be cited by others working in fusion engineering and its related disciplines. Readers benefit as well. For example, this issue features a total of 22 papers on ITER and Wendelstein 7-X, a unique collection of papers on two projects that sit at the forefront of fusion engineering.

A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field (BT0 = 7.4 T), R = 6.8 m, a = 2.1 m, and a steady-state operation. The design concepts of the K-DEMO blanket system considering the cooling in-vessel components with pressurized water and a solid pebble breeder are described herein. The structure of the K-DEMO blanket is toroidally subdivided into 16 inboard and 32 outboard sectors, in order to allow the vertical maintenance. Each blanket module is composed of plasma-facing first wall, layers of breeding parts, shielding and manifolds. A ceramic breeder using Li4SiO4 pebbles with Be12Ti as neuron multiplier is employed for study. MCNP neutronic simulations and thermo-hydraulic analyses are interactively performed in order to satisfy two key aspects: achieving a global Tritium Breeding Ratio (TBR) >1.05 and operating within the maximum allowable temperature ranges of materials.

As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200 GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb3Sn superconducting conductor. The peak magnetic field reaches to 16 T with the magnetic field at the plasma center above 7 T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material.

The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF ...

Modular coil characteristics of a 2-field periods quasi-axisymmetric stellarator QAS-LA configuration with an aspect ratio Ap = 3, magnetic pressure ∼4% and rotational transform ι ∼ 0.15 per field period supplied by its own shaping have been detailed studied. In addition, the characteristics of modular coils for QAS-LA were compared with those of an intermediate QA configuration QAS-LAx and a tokamak based on the same center magnet field B0, aspect ratio and number of coils. As expected, the Bmax/B0, force F and overturning moment M, increase with the increased complexity of the coil shape. The relationships between the modular coils’ parameters (such as radius curvature ρ, distance from coil to coil Δc–c and the cross-section of coils) and the electromagnetic characteristics have been systematically summarized. The approximate formula for the maximum magnetic field in the coil body as functions of modular coil parameters (Δc–c, ρ) was derived for a simple two wire system which will be useful when optimizations of coil properties are called for.

The quasi-axisymmetric (QA) stellarator, a 3-D magnetic configuration with close connections to tokamaks, offers solutions for a steady state, disruption-free fusion system. A new experimental facility, QUASAR, provides a rapid approach to the next step in QA development, an integrated experimental test of its physics properties, taking advantage of the designs, fabricated components, and detailed assembly plans developed for the NCSX project. A scenario is presented for constructing the QUASAR facility for physics research operations starting in 2019. Operating in deuterium, such a facility would investigate the scale-up in size and pulse length from QUASAR, while a suitably equipped version operating in deuterium-tritium (DT) could address fusion nuclear missions. New QA optimization strategies, aimed at improved engineering attractiveness, would also be tested.

As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb3Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

The stellarator fusion experiment Wendelstein 7-X (W7-X) is currently under construction at the Max-Planck-Institut fur Plasmaphysik in Greifswald, Germany. The main magnetic field will be provided by a superconducting magnet system which generates a fivefold toroidal periodic magnetic field. However, unavoidable tolerances can result in small deviations of the magnetic field which disturb the toroidal periodicity. In order to have a tool to influence these field errors five additional normal conducting trim coils were designed to allow fine tuning of the main magnetic field during plasma operation. In the frame of an international cooperation the trim coils will be contributed by the US partners. Princeton Plasma Physics Laboratory has accomplished several tasks to develop the final design ready for manufacturing e.g. detailed manufacturing design for the winding and for the coil connection area. The design work was accompanied by a detailed analysis of resulting forces and moments to prove the design. The manufacturing of the coils is running at Everson Tesla Inc; the first two coils were received at IPP.

A study has confirmed the feasibility of designing, fabricating and installing resonant magnetic field perturbation (RMP) coils in JET 1 with the objective of controlling edge localized modes (ELM). A system of two rows of in-vessel coils, above the machine midplane, has been chosen as it not only can investigate the physics of and achieve the empirical criteria for ELM suppression, but also permits variation of the spectra allowing for comparison with other experiments. These coils present several engineering challenges. Conditions in JET necessitate the installation of these coils via remote handling, which will impose weight, dimensional and logistical limitations. And while the encased coils are designed to be conventionally wound and bonded, they will not have the usual benefit of active cooling. Accordingly, coil temperatures are expected to reach 350 °C during bakeout as well as during plasma operations. These elevated temperatures are beyond the safe operating limits of conventional OFHC copper and the epoxies that bond and insulate the turns of typical coils. This has necessitated the use of an alternative copper alloy conductor C18150 (CuCrZr). More importantly, an alternative to epoxy had to be found. An R&D program was initiated to find the best available insulating and bonding material. The search included polyimides and ceramic polymers. The scope and status of this R&D program, as well as the critical engineering issues encountered to date are reviewed and discussed.

The results from the investigation of neoclassical core transport and the role of the radial electric field profile (Er) in the first operational phase of the Wendelstein 7-X (W7-X) stellarator are presented. In stellarator plasmas, the details of the Er profile are expected to have a strong effect on both the particle and heat fluxes. Investigation of the radial electric field is important in understanding neoclassical transport and in validation of neoclassical calculations. The radial electric field is closely related to the perpendicular plasma flow (u⊥) through the force balance equation. This allows the radial electric field to be inferred from measurements of the perpendicular flow velocity, which can be measured using the x-ray imaging crystal spectrometer and correlation reflectometry diagnostics. Large changes in the perpendicular rotation, on the order of Δu⊥∼ 5 km/s (ΔEr ∼ 12 kV/m), have been observed within a set of experiments where the heating power was stepped down from 2 MW to 0.6 MW. The...

The National Compact Stellarator Experiment (NCSX) was designed to test a compact, quasiaxisymmetric stellarator configuration. Flexibility and accurate realization of its complex 3D geometry were key requirements affecting the design and construction. While the project was terminated before completing construction, there were significant engineering accomplishments in design, fabrication, and assembly. The design of the stellarator core device was completed. All of the modular coils, toroidal field coils, and vacuum vessel sectors were fabricated. Critical assembly steps were demonstrated. Engineering advances were made in the application of CAD modeling, structural analysis, and accurate fabrication of complex-shaped components and subassemblies. The engineering accomplishments of the project are summarized

The National Compact Stellarator Experiment (NCSX) has been under construction at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The stellarator core is designed to produce a compact 3D plasma that combines stellarator and tokamak physics advantages. The complex geometry and tight fabrication tolerances of NCSX create some unique engineering and assembly challenges. The NCSX project was cancelled in May 2008; construction activities are presently being phased out in an orderly fashion. This paper will describe the progress of the fabrication and assembly activities of NCSX. Completion of the coil fabrication is on track for the summer of 2008. All three of the vacuum vessel 120° sections have been delivered. Assembly of vacuum vessel services began in May 2006 and is now complete. Assembly of the modular coils into 3-packs for safe storage is presently underway.

Engineering innovation is required to reduce cost and risk in fabrication for the Quasi-Poloidal Stellarator being developed to test key physics issues at very low plasma aspect ratio. Complex, highly accurate, stainless steel modular coil winding forms are cast and machined; conductor is wound directly onto the winding forms; a vacuum-tight cover is welded over each coil pack; the coils are vacuum pressure impregnated; the completed coils are installed in an external vacuum vessel. An internally cooled, compacted cable conductor that can be wound into complex 3-D shapes was developed. The largest and most complex of the winding forms has been cast using a patternless process (machined sand molds) and a high-temperature pour. The resulting casting required

The super-conducting stellarator Wendelstein 7-X is presently under construction at the Max-Planck-Institute for Plasma Physics in Greifswald, Germany. Assembly of the device is almost completed and the periphery systems and the diagnostic and heating systems are well advanced. Commissioning of the device has been prepared over the last 2 years and has started in April 2014. This is the first time since decades that a superconducting fusion device is commissioned in Europe.

For the (non-axisymmetric) stellarator class of plasma confinement devices to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux surfaces; however, the inherent lack of a continuous symmetry implies that magnetic islands responsible for breaking the smooth topology of the flux surfaces are guaranteed to exist. Thus, the suppression of magnetic islands is a critical issue for stellarator design, particularly for small aspect ratio devices. Pfirsch–Schluter currents, diamagnetic currents and resonant coil fields contribute to the formation of magnetic islands, and the challenge is to design the plasma and coils such that these effects cancel.Magnetic islands in free-boundary high-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the Princeton Iterative Equilibrium Solver (Reiman and Greenside 1986 Comput. Phys. Commun. 43 157) which iterates the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to preserve certain measures of engineering acceptability and to preserve the stability of ideal kink modes. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible, the plasma is stable to ideal kink modes, and the coils satisfy engineering constraints. The method is applied to a candidate plasma and coil design for the National Compact Stellarator eXperiment (Reiman et al 2001 Phys. Plasma 8 2083).

The National Compact Stellarator Experiment (NCSX) [ http://www.pppl.gov/ncsx/Meetings/CDR/CDRFinal/EngineeringOverview_R2.pdf ] is being designed as a proof of principal test of a quasi-axisymmetric compact stellarator. This concept combines the high beta and good confinement features of an advanced tokamak with the low current, disruption-free characteristics of a stellarator. NCSX has a three-field-period plasma configuration with an average major radius of 1.4 m, an average minor radius of 0.33 m and a toroidal magnetic field on axis of up to 2 T. The stellarator core is a complex assembly of four coil systems that surround the highly shaped plasma and vacuum vessel. Heating is provided by up to four, 1.5 MW neutral beam injectors and provision is made to add 6 MW of ICRH. The experiment will be built at the Princeton Plasma Physics Laboratory, with first plasma expected in 2007.

Compact quasi-axisymmetric stellarators offer the possibility of combining the steady-state low-recirculating power, external control, and disruption resilience of previous stellarators with the low-aspect ratio, high beta-limit, and good confinement of advanced tokamaks. Quasi-axisymmetric equilibria have been developed for the proposed National Compact Stellarator Experiment (NCSX) with average aspect ratio ˜4.4 and average elongation ˜1.8. Even with bootstrap-current consistent profiles, they are passively stable to the ballooning, kink, vertical, Mercier, and neoclassical-tearing modes for β = 4%, without the need for external feedback or conducting walls. The bootstrap current generates only 1/4 of the magnetic rotational transform at β = 4% (the rest is from the coils). Transport simulations show adequate fast-ion confinement and thermal neoclassical transport similar to equivalent tokamaks. Modular coils have been designed which reproduce the physics properties, provide good flux surfaces, and allow flexible variation of the plasma shape to control the predicted MHD stability and transport properties.

Key physics issues in the design of a high- quasi-axisymmetric stellarator configuration are discussed. The goal of the design study is a compact stellarator configuration with aspect ratio comparable to that of tokamaks and good transport and stability properties. Quasi- axisymmetry has been used to provide good drift trajectories. Ballooning stabilization has been accomplished by strong axisymmetric shaping, yielding a stellarator configuration whose core is in the second stability regime for ballooning modes. A combination of externally generated shear and non-axisymmetric corrugation of the plasma boundary provides stability to external kink modes even in the absence of a conducting wall. The resulting configuration is also found to be robustly stable to vertical modes, increasing the freedom to perform axisymmetric shaping. Stability to neoclassical tearing modes is conferred by a monotonically increasing profile. A gyrokinetic f code has been used to confirm the adequacy of the neoclassical confinement. Neutral beam losses have been evaluated with Monte Carlo codes.

The Korea Superconducting Tokamak Advanced Research (KSTAR) Project is the major effort of the Korean National Fusion Program (KNFP) to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. Major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 Tesla, and plasma current 2 mA with a strongly shaped plasma cross-section and double-null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beam, ion cyclotron waves, lower hybrid waves, and electron–cyclotron waves for flexible profile control. A comprehensive set of diagnostics is planned for plasma control and performance evaluation and physics understanding. The project has completed its conceptual design phase and moved to the engineering design phase. The target date of the first plasma is set for year 2002.

Recent progress in the theoretical understanding and design of compact stellarators is described. Hybrid devices, which depart from canonical stellarators by deriving benefits from the bootstrap current which flows at finite beta, comprise a class of low aspect ratio A

Sixty representatives of the fusion community and the Department of Energy met October 22-24, 1996 in a workshop to chart the short and medium term future of the nation's fusion energy science program. The fusion scientists represented nearly all the institutions and scientific areas covered by the fusion program. Plans were crafted to implement the goals of the restructuring recommended last winter by the Fusion Energy Advisory Committee. The workshop refined the vision of the fusion program, articulated the accomplishments expected over the next 5 years, suggested changes in organizational methodology, reaffirmed the U.S. community's commitment to ITER and agreed to participate in the international collaborative process beyond the Engineering Design Activities toward construction, described future roles for the Princeton Plasma Physics Laboratory as a national laboratory, and produced the beginning of plans for education and outreach to the larger scientific community and public. Each of these items is discussed below: Taken as a whole, they represent a sea change for the U.S. fusion program. This change was embraced by the workshop attendees. It forms the basis for a fusion and plasma science program which will continue to reap far-reaching benefits to the nation in the near term, and progress toward a renewable and attractive energy source for mankind in the long term. It is expected that the new program will be enduring. It is based upon principles which will evolve, but not be overturned, as the nation's energy and science needs evolve.

K-DEMO is being studied by South Korean researchers as a follow-on to ITER and the next step toward the construction of a commercial fusion power plant. The K-DEMO mission defines a staged approach targeting operation with an initial testing phase for plasma facing components and critical operating systems to be followed by a second phase which centers on upgrading the in-vessel components for operation at 200 to 600 MWe with a planned 70% availability. This paper reviews the general arrangement of the K-DEMO device core, the novel configuration concept for the vertical maintenance of large in-vessel segments and describes the arrangement and maintenance of planned interfacing auxiliary systems and services - design features which impact the ability to operate with a staged mission strategy that ends with high availability operations.

The quasi-axisymmetric (QA) stellarator, a three-dimensional magnetic configuration with close connections to tokamaks, offers solutions for a steady-state, disruption-free fusion system. A new experimental facility, QUASAR, provides a rapid approach to the next step in QA development, an integrated experimental test of its physics properties, taking advantage of the designs, fabricated components, and detailed assembly plans developed for the NCSX project. A scenario is presented for constructing the QUASAR facility for physics research operations starting in 2019. A facility for the step beyond QUASAR, performance extension to high temperature, high pressure sustained plasmas is described. Operating in DD, such a facility would investigate the scale-up in size and pulse length from QUASAR, while a suitably equipped version operating in DT could address fusion nuclear missions, with operation starting in 2027.

The Korean DEMO program is pursuing a steady state tokamak configuration to develop a fusion energy producing facility. Systems analysis is performed to determine its geometry and operating space available. After the plasma major radius and elongation is chosen, and the maximum toroidal magnetic field at the coil is established, the operating space can be explored with a range of assumptions. A database approach for the systems analysis is used that generates a large number of solutions, that can be used to examine sensitivities and parameter uncertainties.

Control of the poloidal field (PF) in the Tokamak Physics Experiment (TPX) is critical to achieving its mission of advanced tokamak research. Extensive examination of the plasma equilibrium; plasma start-up; plasma position, shape, and current control; and plasma shape reconstruction have been performed as part of the design process. This paper reports the progress in this area. The PF coils have been designed to produce a wide range of plasmas. Plasma start-up can be achieved for multiple conditions. Fast plasma position control coils inside the vacuum vessel are used for short timescale control of the plasma vertical and radial position. Shape and total plasma-current control are provided by the PF coils over a slower timescale. A new algorithm for shape control of a few critical plasma boundary points is described and used in simulations using the Tokamak Simulation Code. Fast magnetostatic reconstruction of the plasma shape is examined to determine the impact of measurement locations and their quality.

The Tokamak Physics Experiment (TPX) is planned to develop the scientific basis for an economically competitive and continuously operating tokamak fusion power source. It has been designed to have steady-state operating capability, sufficient performance to produce reactor-like plasma configurations, and a flexible set of steady-state plasma controls. Active plasma control (e.g., current profile control, shape and position control, passive and active MHD mode stabilization, and toroidal rotation control) is a key to achieving steady state tokamak operating conditions with enhanced beta and confinement, efficient current drive, high purity, and high reliability. Inductive scenarios and steady-state operating modes with current-drive have been studied to determine the system requirements for access and maintenance of advanced steady-state modes. Industry contractors have begun detailed engineering design of the superconducting magnets, vacuum vessel, and plasma-facing components. 8 refs., 3 figs., 1 tab.

Experimental and theoretical results from around the world point to the possibility of high confinement, high- beta , and high-bootstrap-fraction steady-state tokamak operating modes. These modes of operation, if fully developed and extended to steady-state, could lead to much less expensive tokamak demonstration power reactors and to a significantly reduced cost-of-electricity from fusion, as compared to projections based on low- beta N, pulsed operating modes. Present results have clear implications in the areas of particle control, plasma shaping, and current-profile control. Thus they have strongly influenced the design of the steady-state advanced tokamak TPX, which has the mission to combine the best results from present experiments and extend them to steady-state. These results also have important implications for follow-up tests in ITER, which have the goal of studying advanced-tokamak operation in an ignited plasma, as well as for the eventual configuration of an advanced-tokamak fusion reactor.

A fusion pilot plant study was initiated to evaluate the potential benefits of following the fission development path as an approach for the commercialization of fusion. In such an approach, a fusion pilot plant would bridge the development needs in moving from ITER to a first of a kind fusion power plant. The pilot plant mission would encompass the component test and fusion nuclear science missions yet produce net electricity. In the first phase of the study scoping designs were developed for three different magnetic configuration options: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS). Critical component features have been added to the designs that impact the general arrangement and maintenance characteristics of each device. The requirements specified in defining the pilot plant challenge the machine configurations developed for each option. Developing multiple options with a consistent set of requirements enables a uniform comparison of configuration and component issues that drive each design. This paper will provide an engineering design overview of each option, address open issues and assess where further work is needed to meet the pilot plant objectives.

With the ITER era now well underway, the fusion community is considering the next major steps in magnetic fusion energy (MFE) development. It follows that there is heightened interest worldwide in understanding the roadmap to commercial MFE. In reality, there is no unique roadmap. An important differentiator among possible pathways is risk, i.e. the risks accepted in going from step to step and how risks are mitigated through R&D programs that accompany and support the progression of major nuclear devices. We consider a rollback approach, starting from a definition of what Demo (a power plant that is the last step before commercialization) must accomplish. We assess, in fusion science and technology terms, the mission and requirements for Demo, its prerequisites, and the requirements for a major nuclear devices and the accompanying programs that could precede Demo in order to satisfy its prerequisites. One option for a pre-Demo MFE device is a pilot plant, a facility that would develop and test nuclear components surrounding the plasma, prototype maintenance schemes applicable to a power plant, and demonstrate both tritium self-sufficiency and net electricity generation. An initial assessment of the pilot plant, in terms of its potential to satisfy Demo prerequisites and the associated risks, is presented.

A study has confirmed the feasibility of designing, fabricating and installing resonant magnetic field perturbation (RMP) coils in JET with the objective of controlling edge localized modes (ELM). These coils present several engineering challenges. Conditions in JET necessitate the installation of these coils via remote handling, which will impose weight, dimensional and logistical limitations. And while the encased coils are designed to be conventionally wound and bonded, they will not have the usual benefit of active cooling. Accordingly, coil temperatures are expected to reach 350C during bakeout as well as during plasma operations from resistive heating. These elevated temperatures are beyond the safe operating limits of conventional OFHC copper and the epoxies that bond and insulate the turns of typical coils. This has necessitated the use of an alternative copper alloy conductor C18150 (CuCrZr). More importantly, an alternative to epoxy had to be found. An R&D program was initiated to find the best available insulating and bonding material. The search included polyimides and ceramic polymers. Ultimately, these ELM coils must be able to withstand the elevated thermal conditions as well as the structural stresses resulting from electromagnetic loads, which include eddy current and halo current effects. Not only do these loads affect the performance of the coils and cases, but also impact the design of joints, leads, jumpers, and the mounting of the coils to the interior of the vacuum vessel wall. In order to qualify the proposed insulating and bonding materials, prototypical coil samples were built to the design specifications of the proposed JET ELM coils. These samples were impregnated with polyimide then cured. This paper will detail the R&D program, including the results of testing to determine mechanical properties of the polyimide bonded coil samples.

The National Compact Stellarator Experiment (NCSX) was designed to test physics principles of an innovative stellarator design developed by Princeton Plasma Physics Laboratory and Oak Ridge National Laboratory. Construction of some of the major components and subassemblies was completed, but the estimated cost and schedule for completing the project grew as the technical requirements and risks became better understood, leading to its cancellation in 2008. The project's risks stemmed from its technical challenges, primarily the complex component geometries and tight tolerances that were required. The initial baseline, which was established in 2004, was supported by a risk management plan and risk-based contingencies, both of which proved to be inadequate. Technical successes were achieved in the construction of challenging components and subassemblies, but cost and schedule growth was experienced. As part of an effort to improve project performance, a new risk management program was devised and implemented in 2007-2008. It led to a better understanding of project risks, a sounder basis for contingency estimates, and improved management tools. Although the risks were ultimately unacceptable to the sponsor, valuable lessons in risk management were learned through the experiences with the NCSX project.

The National Compact Stellarator Experiment (NCSX) is designed to test physics principles of an innovative stellarator design developed by the Princeton Plasma Physics Laboratory (PPPL) and Oak Ridge National Laboratory (ORNL). The project was technically very challenging, primarily due to the complex component geometries and tight tolerances that were required. As the project matured these challenges manifested themselves through all phases of the project (i.e. design, R&D, fabrication and assembly). Although the project was not completed, several major work packages, comprising about 65% of the total estimated cost (excluding management and contingency), were completed, providing a data base of actual costs that can be analyzed to understand cost drivers. Technical factors that drove costs included the complex geometry, tight tolerances, material requirements, and performance requirements. Management factors included imposed annual funding constraints that throttled project cash flow, staff availability, and inadequate R&D. Understanding how requirements and design decisions drove cost through this top-down forensic cost analysis could provide valuable insight into the configuration and design of future Stellarators and other devices.

The Compact Ignition Tokamak is a high-performance device designed to study the physics of burning plasmas. Key physics aspects of the design are described, including plasma performance, disruption...

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). Its mission is to develop the physics understanding of the compact stellarator and evaluate its potential for future fusion energy systems. Compact stellarators use 3D plasma shaping to produce a magnetic configuration that can be steady state without current drive or feedback control of instabilities. The NCSX has major radius 1.4 m, aspect ratio 4.4, 3 field periods, and a quasi-axisymmetric magnetic field. It is predicted to be stable and have good magnetic surfaces at beta > 4% and to have tokamak-like confinement properties. The device will provide the plasma configuration flexibility and the heating and diagnostic access needed to test physics predictions. Component production has advanced substantially since the first contracts were placed in 2004. Manufacture of the vacuum vessel was completed in 2006. All eighteen modular coil winding forms have been delivered, and twelve modular coils have been wound and epoxy impregnated. A contract for the (planar) toroidal field coils was placed in 2006 and manufacture is in progress. Assembly activities have begun and will be the project's main focus in the next few years. The engineering challenge of NCSX is to meet the requirements for complex geometries and tight tolerances within the cost and schedule constraints of a construction project. This paper will focus on how the engineering challenges of component production have been resolved, and how the assembly challenges are being met.

A Korean fusion energy development promotion law (FEDPL) was enacted in 2007. As a following step, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. After the thorough 0D system analysis, the parameters of the main machine characterized by the major and minor radii of 6.8 and 2.1 m, respectively, were chosen for further study. The analyses of heating and current drives were performed for the development of the plasma operation scenarios. Preliminary results on lower hybrid and neutral beam current drive are included herein. A high performance Nb3Sn-based superconducting conductor is adopted, providing a peak magnetic field approaching 16 T with the magnetic field at the plasma centre above 7 T. Pressurized water is the prominent choice for the main coolant of K-DEMO when the balance of plant development details is considered. The blanket system adopts a ceramic pebble type breeder. Considering plasma performance, a double-null divertor is the reference configuration choice of K-DEMO. For a high availability operation, K-DEMO incorporates a design with vertical maintenance. A design concept for K-DEMO is presented together with the preliminary design parameters.

The National Compact Stellarator Experiment (NCSX) is the first of a new class of stellarators known as "compact stellarators". Stellarators are characterized by three dimensional magnetic fields and plasma shapes and are the best-developed class of magnetic fusion devices after the tokamak. Stellarators are attractive because they solve critical problems of magnetic fusion energy: steady state operation without current drive and stable operation without feedback control or rotation drive. The differentiating feature of the compact stellarator is the use of a small plasma current in conjunction with external magnetic fields to provide the required plasma shaping and confinement. This permits a more compact stellarator design. Many of the components of NCSX are conventional in design and manufacture; however, two of the most critical components, the vacuum vessel and the modular coil winding forms, are also the most challenging components to manufacture. This paper describes design of these two critical elements of NCSX and the three-phase program that is being used in the manufacturing development of these components.

The engineering design status of the quasi-poloidal stellarator experiment (QPS) is presented. The overall configuration and the design, manufacturing and assembly techniques of the various components of the core are described.

Stellarators use 3D plasma and magnetic field shaping to produce a steady-state disruption-free magnetic confinement configuration. Compact stellarators have additional attractive properties-quasi-symmetric magnetic fields and low aspect ratio. The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL) to test the physics of a high-beta compact stellarator with a low-ripple, tokamak-like magnetic configuration. The engineering challenges of NCSX stem from its complex geometry requirements. These issues are addressed in the construction project through manufacturing R&D and system engineering. As a result, the fabrication of the coil winding forms and vacuum vessel are proceeding in industry without significant technical issues, and preparations for winding the coils at PPPL are in place. Design integration, analysis, and dimensional control are functions provided by system engineering to ensure that the finished product will satisfy the physics requirements, especially accurate realization of the specified coil geometries. After completion of construction in 2009, a research program to test the expected physics benefits will start

The production phase for the NCSX modular coil winding forms has been underway for approximately one year as of this date. This is the culmination of R&D efforts which were performed in 2001-4. The R&D efforts included limited manufacturing studies while NCSX was in its conceptual design phase followed by more detailed manufacturing studies by two teams which included the fabrication of full scale prototypes. This provided the foundation necessary for the production parts to be produced under a firm price and schedule contract which was issued in September, 2004. This paper will describe the winding forms, the production team and team management, details of the production process, and the achievements for the first year

The engineering design status of the quasi-poloidal stellarator experiment (QPS) is presented. The overall configuration and the design, manufacturing R&D and assembly techniques of the core components are described

NCSX [National Compact Stellarator Experiment] is the first of a new class of stellarators called compact stellarators which hold the promise of retaining the steady state feature of the stellarator but at a much lower aspect ratio and using a quasi-axisymmetric magnetic field to obtain tokamak-like performance. Although much of NCSX is conventional in design and construction, the vacuum vessel and modular coils provide significant engineering challenges due to their complex shapes, need for high dimensional accuracy, and the high current density required in the modular coils due space constraints. Consequently, a three-phase development program has been undertaken. In the first phase, laboratory/industrial studies were performed during the development of the conceptual design to permit advances in manufacturing technology to be incorporated into NCSX's plans. In the second phase, full-scale prototype modular coil winding forms, compacted cable conductors, and 20 degree sectors of the vacuum vessel were fabricated in industry. In parallel, the NCSX project team undertook R&D studies that focused on the windings. The third (production) phase began in September 2004. First plasma is scheduled for January 2008.

The n=0 axisymmetric vertical stability and vertical position control have been examined for the Tokamak Physics Experiment. The passive stabilization is accomplished by using stabilizer plates close to the plasma. The present configuration is found to provide robust stability over a wide range of plasma parameters. The active feedback control of the plasma vertical position is done using coils located inside the vacuum vessel. These are required to control random disturbances leading to /spl les/1.0 cm RMS displacements from the midplane, and acceptable coil currents and voltages are found.

Operation of the Burning Plasma Experiment (BPX) with a deuterium-tritium plasma and graphite divertor plates will result in significant retention of tritium in layers of eroded and subsequently redeposited carbon. This trapped tritium is at risk of release during a vacuum accident. An administrative limit of 2 g of releasable tritium has been established for assessment of the environmental impact of such an event. Analysis of the erosion and redeposition process indicates that the 2-g limit will be reached in about 200 full-parameter discharges. Helium-oxygen glow discharge cleaning will be used to remove the redeposited carbon and the trapped tritium. It is calculated that about 90 h of cleaning are required to remove the trapped tritium from 200 plasmas. The cleaning will be done during scheduled maintenance periods every third week. Studies indicate that the tritium inventory can be effectively managed with helium-oxygen glow discharge cleaning. >