T. Brown

Princeton University, Princeton, New Jersey, United States

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Publications (53)14.21 Total impact

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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.
    Nuclear Fusion 12/2013; 53(12):126001. DOI:10.1088/0029-5515/53/12/126001 · 3.24 Impact Factor
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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.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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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.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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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.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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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.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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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.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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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.
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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.
    Nuclear Fusion 08/2011; 51(10):103014. DOI:10.1088/0029-5515/51/10/103014 · 3.24 Impact Factor
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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.
    Fusion Engineering (SOFE), 2011 IEEE/NPSS 24th Symposium on; 07/2011
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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.
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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.
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    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.
    Fusion Engineering, 2009. SOFE 2009. 23rd IEEE/NPSS Symposium on; 07/2009
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    T. Brown, L. 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.
    Fusion Engineering, 2009. SOFE 2009. 23rd IEEE/NPSS Symposium on; 07/2009
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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.
    Fusion Engineering, 2009. SOFE 2009. 23rd IEEE/NPSS Symposium on; 07/2009
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    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.
    Fusion Engineering, 2007. SOFE 2007. 2007 IEEE 22nd Symposium on; 07/2007
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    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.
    Fusion Engineering, 2007. SOFE 2007. 2007 IEEE 22nd Symposium on; 07/2007
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    ABSTRACT: The national compact stellarator experiment (NCSX) will have an extensive set of external magnetic diagnostics. These include flux loops on the exterior surface of the vacuum vessel. Data from these sensors will be integrated with other magnetic sensors and used for plasma control and to constrain magnetic equilibrium reconstructions. NCSX is currently under construction at the Princeton Plasma Physics Laboratory (PPPL). The ex-vessel flux loops must be installed during machine construction since they will ultimately be trapped in the space between the vacuum vessel and the modular coil support shell. Detailed designs have been completed, locator templates have been fabricated and approximately one third of the 225 total loops have been installed as of mid February 2007. Modeling was performed by PPPL to determine the optimum size, placement and number of turns. Engineering of the flux loops was challenging as they must be accurately positioned, optimized geometry maintained and they must be robust and reliable in a bake and cryogenic environment for the lifetime of NCSX. Designs for the ex-vessel flux loops that meet these requirements are presented.
    Fusion Engineering, 2007. SOFE 2007. 2007 IEEE 22nd Symposium on; 07/2007
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    ABSTRACT: The National Compact Stellarator Experiment will have a complete set of magnetic diagnostics to constrain equilibrium reconstructions and for plasma control. The flux loops lying on the exterior surface of the vacuum vessel and the flux loops cowound with the field coils must be installed during machine construction because they will later be inaccessible. Designs and installation techniques for these diagnostics are described.
    Review of Scientific Instruments 10/2006; 77(10). DOI:10.1063/1.2229227 · 1.58 Impact Factor
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    ABSTRACT: The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (βN = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κx = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher βN. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 °C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb–17Li to about 1000 °C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to an attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (4.7 ¢/kWh), which is competitive with those projected for other sources of energy.
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    ABSTRACT: The National Compact Stellarator Experiment (NCSX) is a quasi-axisymmetric facility that combines the high beta and good confinement features of an advanced tokamak with the low current, disruption-free characteristics of a stellarator. The experiment is based on a three field-period plasma configuration with an average major radius of 1.4 m, a minor radius of 0.3 m, and a toroidal magnetic field on axis of up to 2 T. The modular coils are one set in a complex assembly of four coil systems that surround the highly shaped plasma. There are six, each of three coil types in the assembly for a total of 18 modular coils. The coils are constructed by winding copper cable onto a cast stainless steel winding form that has been machined to high accuracy, so that the current center of the winding pack is within ±1.5 mm of its theoretical position. The modular coils operate at a temperature of 80 K and are subjected to rapid heating and stress during a pulse. At this time, the project has completed construction of several prototype components which validate the fabrication and inspection processes that are planned for the production coils. In addition, some advanced techniques for error-field compensation and assembly simulation using computer-aided design (CAD) have been developed.
    Fusion Engineering and Design 11/2005; DOI:10.1016/j.fusengdes.2005.06.254 · 1.15 Impact Factor