[Show abstract][Hide abstract] ABSTRACT: 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.
[Show abstract][Hide abstract] ABSTRACT: The next step in the Wendelstein stellarator line is the large superconducting device Wendelstein 7-X, currently under construction in Greifswald, Germany. Steady-state operation is an intrinsic feature of stellarators, and one key element of the Wendelstein 7-X mission is to demonstrate steady-state operation under plasma conditions relevant for a fusion power plant. Steady-state operation of a fusion device, on the one hand, requires the implementation of special technologies, giving rise to technical challenges during the design, fabrication and assembly of such a device. On the other hand, also the physics development of steady-state operation at high plasma performance poses a challenge and careful preparation. The electron cyclotron resonance heating system, diagnostics, experiment control and data acquisition are prepared for plasma operation lasting 30 min. This requires many new technological approaches for plasma heating and diagnostics as well as new concepts for experiment control and data acquisition.
[Show abstract][Hide abstract] ABSTRACT: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The pilot plant mission encompassed component test and fusion nuclear science missions plus the requirement to produce net electricity with high availability in a device designed to be prototypical of the commercial device. Three magnetic configuration options were developed around this mission: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS). With the completion of the study and separate documentation of each design option a question can now be posed; how do the different designs compare with each other as candidates for meeting the pilot plant mission? In a pro/con format this paper will examine the key arguments for and against the AT, ST and CS magnetic configurations. Key topics addressed include: plasma parameters, device configurations, size and weight comparisons, diagnostic issues, maintenance schemes, availability influences and possible test cell arrangement schemes.
Fusion Science and Technology 09/2013; 64(3):662-669. · 0.59 Impact Factor
[Show abstract][Hide abstract] ABSTRACT: The design features developed for the Spherical Tokamak (ST) in the PPPL pilot plant study was used as the starting point in developing designs to meet the mission of a Fusion Nuclear Science Facility (FNSF) considering a range of machine sizes based on the influence of tritium consumption and maintenance strategies. The compact nature of a steady state operated ST device for this mission pushes operating conditions and places challenges in the design of components, device maintenance and the integration of supports and services. This paper reviews the general arrangement, design details and maintenance strategy of the ST-FNSF device core for a 1.6-m and 1.0-m device; operating points which bracket the region between purchasing and breeding tritium.
[Show abstract][Hide abstract] ABSTRACT: As a continuation of initial conceptual design work for a steady-state Korean fusion DEMO Reactor (K-DEMO), a bit more detailed K-DEMO magnet conceptual design is being carried out. The size of the K-DEMO is only slightly bigger than the ITER and the major radius is around 6.8 m. But the peak field of toroidal field (TF) magnets is as high as ~16 T. Due to a stability issue, the TF magnets will be made of two different cable-in-conduit conductors (CICC's) for the high and relatively low field regions. Some engineering issues, including possible inter coil joint schemes, are discussed. Both CICC's for the TF magnets are designed by assuming the use of a currently available high performance Nb3Sn wire. Preliminary CICC design parameters are presented together with simulation results using the code GANDALF. A vertical maintenance scheme is being discussed for the K-DEMO and the location of poloidal field (PF) coils are recently set. However, a preliminary work on central solenoid (CS) coil has been carried out. The CS coils are designed to generate ~83 Wb of flux swing. Preliminary design parameters for the CS CICC are also presented.
[Show abstract][Hide abstract] ABSTRACT: The maximum achievable tritium breeding ratio (TBR), the dose to the insulator of the Cu coils, and the radial build definition are among numerous design issues investigated in detail for a Fusion Nuclear Science Facility (FNSF) based on the spherical tokamak (ST) concept. The ongoing PPPL study is considering a range of machine sizes with 1-2.2 m major radius. Preliminary shielding analysis for the PF coils of the intermediate size machine (R~1.7 m) indicated excessive dose to the cyanate ester/epoxy organic insulator, suggesting a more radiation resistant ceramic insulator such as MgO. The 3-D analysis predicts a TBR of ~1 when the details of the dual-cooled LiPb blanket and outboard penetrations are included in the model. Potential means to increase the TBR were investigated. Understanding the impact of various device sizes on the TBR is another important ongoing research activity to determine the threshold in device size for achieving T self-sufficiency.
[Show abstract][Hide abstract] ABSTRACT: 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.
[Show abstract][Hide abstract] ABSTRACT: 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.
[Show abstract][Hide abstract] ABSTRACT: Stellarators have a significant advantage as a pilot plant since they do
not need current drive, reducing recirculating power, reducing required
technology development, and easing tritium breeding. In addition,
stellarators have soft performance limits without disruptions, and thus
do not require nearby conducting walls, thick plasma-facing armor,
active plasma stability control, or current profile control. A
stellarator pilot plant design based on a quasi-axisymmetric (QA)
configuration with aspect ratio 4.5, major radius 4.75 m, and magnetic
field on axis of 5.6 T is projected to have a Qeng greater than 2.7 and
a peak neutron wall load higher than 2 MW / m^2. The pilot plant
projects to net electricity production with 100-200 MW of fusion power
produced. The QA design can build on the tokamak understanding and data
base, since it is predicted to share many confinement and stability
characteristics with tokamaks. Strategies for simplified coils and
sector-based maintenance using magnetic materials for field shaping will
[Show abstract][Hide abstract] ABSTRACT: A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.
[Show abstract][Hide abstract] ABSTRACT: 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.
[Show abstract][Hide abstract] ABSTRACT: Stellarators offer robust physics solutions for MFE challenges-- steady-state operation, disruption elimination, and high-density operation-- but require design improvements to overcome technical risks in the construction and maintenance of future large-scale stellarators. Using the ARIES-CS design (aspect ratio 4.56) as a starting point, compact stellarator designs with improved maintenance characteristics have been developed. By making the outboard legs of the main magnetic field coils nearly straight and parallel, a sector maintenance scheme compatible with high availability becomes possible. Approaches that can allow the main coil requirements to be relaxed in this way are: 1) increase aspect ratio at the expense of compactness, 2) add local removable coils in the maintenance ports for plasma shaping, and 3) use passive conducting tiles made of bulk high-temperature superconducting material to help shape the magnetic field. Such tiles would be arranged on a shaped, segmented internal support structure behind the shield.
[Show abstract][Hide abstract] ABSTRACT: A potentially attractive next major DT step in fusion development is a device that produces net electricity as quickly as possible in a configuration directly scalable to a power plant. Such a device would accelerate the commercialization of magnetic fusion by both demonstrating net electricity production and also carrying forward a high neutron fluence component testing mission needed to ultimately achieve high availability in fusion systems. This paper will explore three configurations for a pilot plant: the advanced tokamak (AT), spherical tokamak (ST), and compact stellarator (CS). Overall, initial analysis indicates that the CS and AT are the most energy efficient electrically, while the ST is the most compact radially and provides the highest neutron wall loading. This work is supported in part by U.S. DOE Contract #DE-AC02-09CH11466.
[Show abstract][Hide abstract] ABSTRACT: 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.
[Show abstract][Hide abstract] ABSTRACT: A number of technical requirements and performance criteria can drive stellarator costs, e.g., tight tolerances, accurate coil positioning, low aspect ratio (compactness), choice of assembly strategy, metrology, and complexity of the stellarator coil geometry. With the completion of a seven-year design and construction effort of the National Compact Stellarator Experiment (NCSX) it is useful to interject the NCSX experience along with the collective experiences of the NCSX stellarator community to improving the stellarator configuration. Can improvements in maintenance be achieved by altering the stellarator magnet configuration with changes in the coil shape or with the combination of trim coils? Can a mechanical configuration be identified that incorporates a partial set of shaped fixed stellarator coils along with some removable coil set to enhance the overall machine maintenance? Are there other approaches that will simplify the concepts, improve access for maintenance, reduce overall cost and improve the reliability of a stellarator based power plant? Using ARIES-CS and NCSX as reference cases, alternative approaches have been studied and developed to show how these modifications would favorably impact the stellarator power plant and experimental projects. The current status of the alternate stellarator configurations being developed will be described and a comparison made to the recently designed and partially built NCSX device and the ARIES-CS reactor design study.
[Show abstract][Hide abstract] ABSTRACT: The National Compact Stellarator Experiment (NCSX) was a collaborative effort between ORNL and PPPL. PPPL provided the assembly techniques with guidance from ORNL to meet design criteria. The individual vacuum vessel segments, modular coils, trim coils, and toroidal field coils components were delivered to the Field Period Assembly (FPA) crew who then would complete the component assemblies and then assemble the final three field period assemblies, each consisting of two sets of three modular coils assembled over a 120? vacuum vessel segment with the trim coils and toroidal field coils providing the outer layer. The requirements for positioning the modular coils were found to be most demanding. The assembly tolerances required for accurate positioning of the field coil windings in order to generate sufficiently accurate magnetic fields strained state of the art techniques in metrology and alignment and required constant monitoring of assembly steps with laser trackers, measurement arms, and photogrammetry. The FPA activities were being performed concurrently while engineering challenges were being resolved. For example, it was determined that high friction electrically isolated shims were needed between the modular coil interface joints and low distortion welding was required in the nose region of those joints. This took months of analysis and development yet the assembly was not significantly impacted because other assembly tasks could be performed in parallel with ongoing assembly tasks as well as tasks such as advance tooling setup preparation for the eventual welding tasks. The crew technicians developed unique, accurate time saving techniques and tooling which provided significant cost and schedule savings. Project management displayed extraordinary foresight and every opportunity to gain advanced knowledge and develop techniques was taken advantage of. Despite many risk concerns, the cost and schedule performance index was maintained nearly 1.0 during the asse-
mbly phase until project cancellation. In this paper, the assembly logic, the engineering challenges, solutions to those challenges and some of the unique and clever assembly techniques, will be presented.
[Show abstract][Hide abstract] ABSTRACT: The National Compact Stellerator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The complex geometry and tight fabrication tolerances of the NCSX's non-planar coils and vacuum vessel necessitate the use of computerized, CAD-based metrology systems capable of very accurate and reasonably quick measurements. To date, multi-link, portable coordinate measuring machines (pCMM) are used in the fabrication of the non-planar coils. Characterization of the CNC machined coil winding form and subsequent positioning of the conductor centroid (to within +/-0.5 mm) are accomplished via multiple sets of detailed measurements. A laser tracker is used for all phases of work on the vacuum vessel including positioning magnetic diagnostics and vessel ports prior to welding. Future tasks requiring metrology include positioning of the magnet systems and assembly of the three vacuum vessel sub-assemblies onto the final machine configuration. This paper describes the hardware and software used for metrology, as well as the methodology for achieving the required dimensional control and will present an overview of the measurement results to date.
[Show abstract][Hide abstract] ABSTRACT: 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.