T. Brown

Princeton University, Princeton, New Jersey, United States

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Publications (58)20.7 Total impact

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    ABSTRACT: Early implementation of divertor components for the Wendelstein 7-X stellarator will include an inertially cooled system of divertor elements called the Test Divertor Unit (TDU). One part of this system is a scraper element that is intended to explore methods of mitigating heat flux on the ends of the TDU elements. This system will be in place in 2017, after a run period that will involve no divertor, and will precede steady state operation with actively cooled divertors scheduled for 2019. The TDU scraper element is an experimental device with uncertain requirements and with loading conditions which will developed as a part of the experiment. The pattern of heat flux may vary from currently predicted distributions and intensities. The design of the scraper element must accommodate this uncertainty. Originally the mechanical design was to be based on extensive studies for the monoblock- based design of an actively cooled system. An obvious simplification is the elimination of the manifolding needed for the water cooling. The wall panels on which the panels are mounted are to be maintained at 200C or less. Thermal ratcheting of the tiles, supporting structures, and backing structures is managed with adequate cooldown times, thermal anchors, where allowed, and radiative shields. Water cooling of the shields was proposed and rejected. Better radiation modeling is showing less need for multiple shields, but during initial run periods, the scraper element will have to be restricted to an acceptable operating envelope. Thermal instrumentation is recommended.
    No preview · Article · Sep 2015 · Fusion Science and Technology
  • T. Brown · J. Menard · L. El-Gueblay · A. Davis
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    ABSTRACT: One of the goals of the PPPL Spherical Tokamak (ST) Fusion Nuclear Science Facility (FNSF) study was to generate a self-consistent conceptual design of an ST-FNSF device with sufficient physics and engineering details to evaluate the advantages and disadvantages of different designs and to assess various ST-FNSF missions. This included striving to achieve tritium self-sufficiency; the ability to provide shielding protection of vital components and to develop maintenance strategies that could be used to maintain the in-vessel components (divertors, breeding blankets, shield modules and services) and characterize design upgrade potentials to expanded mission evolutions. With the conceptual design of a 2.2 m ST pilot plant design already completed emphasis was placed on evaluating a range of ST machine sizes looking at a major radius of 1m and a mid-range device size between 1 m and 2.2 m. This paper will present an engineering summary of the design details developed from this study, expanding on earlier progress reports presented at earlier conferences that focused on a mid-size 1.7 m device. Further development has been made by physics in defining a Super-X divertor arrangement that provides an expanded divertor surface area and places all PF coils outside the TF coil inner bore, in regions that improve the device maintenance characteristics. Physics, engineering design and neutronics analysis for both the 1.7 m and 1 m device have been enhanced. The engineering results of the PPPL ST-FNSF study will be presented along with comments on possible future directions.
    No preview · Article · Sep 2015 · Fusion Science and Technology
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    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.
    Full-text · Article · May 2015 · Nuclear Fusion
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    ABSTRACT: Substantial advances have been made in the design of stellarator configurations to satisfy physics properties and fabrication feasibility requirements for experimental devices. However, reactors will require further advances in configuration design, in particular with regard to maintenance and operational characteristics, in order to have high availability. The diamagnetic properties of bulk high temperature superconductor (HTS) material can be used to provide simple mechanisms for magnetic fieldshaping by arranging them appropriately in an ambient field produced by relatively simple coils. A stellarator configuration has been developed based on this concept. A small number of toroidal field coils carrying appropriate current would be sufficient to create a background toroidal field. Discrete HTS monoliths ("pucks " or "tiles ") are placed on a shaped structure that can be split in the poloidal direction at arbitrary locations. This allows modular stellarators to be designed with large openings that provide access to remove interior plasma facing components, no longer restricted by highly shaped back legs of the modular coil winding. Unlike a coil, the structure can be assembled and disassembled in pieces of convenient size, facilitating maintenance. Calculations of the effect of the use of monoliths for field modification in stellarators and tokamaks will be described.
    Full-text · Article · Feb 2015 · Fusion Science and Technology

  • No preview · Article · Feb 2015
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    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.
    Full-text · Article · Dec 2013 · Nuclear Fusion
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    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.
    No preview · Article · Sep 2013 · Fusion Science and Technology
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    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.
    No preview · Conference Paper · Jun 2013
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    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.
    No preview · Conference Paper · Jun 2013
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    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.
    Full-text · Conference Paper · Jun 2013
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    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.
    No preview · Conference Paper · Jun 2013
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    K. Kim · S. Oh · S.H. Baek · P. Titus · T. Brown · C. Kessel · Y. Zhai
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    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.
    Full-text · Conference Paper · Jan 2013
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    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 be discussed.
    No preview · Article · Oct 2012
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    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.
    Full-text · Article · Aug 2011 · Nuclear Fusion
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    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.
    Full-text · Conference Paper · Jul 2011
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    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.
    No preview · Article · Nov 2010
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    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.
    No preview · Article · Nov 2010
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    Full-text · Conference Paper · Oct 2010
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    T. Brown · Leslie Bromberg · M. Cole
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    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.
    Preview · Conference Paper · Jul 2009
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    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.
    Full-text · Conference Paper · Jul 2009