Search results for: P. Heitzenroeder
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Fusion Engineering and Design > 2013 > 88 > 9-10 > 1955-1959
The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement...
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ITER is incorporating two types of In Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required so that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are in close proximity to the plasma, mounted just behind the Blanket Shield Modules. This location...
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The NSTX TF coil assembly experienced a failure in the TF center bundle during pulse number 143912 on July 20, 2011. A short circuit between turns 14 and 30 spread to turn 13 and resulted in melting of approximately 2 kg of copper and ejection of copper and solder from the cooling channels at the lower end of the TF bundle. An autopsy of the coil was immediately begun by PPPL personnel that included...
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Fusion Engineering and Design > 2012 > 87 > 9 > 1515-1518
The national spherical torus experiment (NSTX) will be upgraded to provide increased toroidal field, plasma current and pulse length. This involves the replacement of the so-called center stack, including the inner legs of the toroidal field (TF) coil, the Ohmic heating (OH) coil, and the inner poloidal field (PF) coils. In addition the increased performance of the upgrade requires qualification of...
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ITER will incorporate In Vessel Coils (IVCs) as a method of stabilizing “Edge Localized Modes” (ELM) and providing “Vertical Stabilization” (VS). To meet the ELM and VS Coil requirements strong coupling with the plasma is required so that it is necessary for the coils to be installed in the vessel just behind the blanket shield modules. Due to this close proximity to the plasma the radiation and temperature...
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Recently, an Edge Localized Mode (ELM) control coil was developed for use on the DIII-D tokamak. The coil design represented a significant challenge due primarily to the requirement for the coil insulation to withstand bakeout temperatures of 400°C for extended periods. This requirement ruled out most common organic insulating systems and necessitated a significant prototyping and development effort,...
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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...
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The fastening of the National Compact Stellarator Experiment's (NCSX) modular coils presented a number of engineering and manufacturing challenges due to the high magnetic forces, need to control induced currents, tight tolerances and restrictive space envelope. A fastening method using high strength studs, jack nuts, insulating spacers, bushings and alumina coated shims was developed which met the...
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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...
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The purpose of the NSTX Center Stack Upgrade project is to expand the NSTX operational space and thereby the physics basis for next-step ST facilities. The plasma aspect ratio (ratio of plasma major to minor radius) of the upgrade is increased to 1.5 from the original value of 1.26, which increases the cross sectional area of the center stack by a factor of ~ 3 and makes possible higher levels of...
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Fusion Engineering and Design > 2007 > 82 > 5-14 > 575-582
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...
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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...
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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). The stellarator core is designed to produce a compact 3-D plasma that combines stellarator and tokamak physics advantages. The complex geometry and tight fabrication tolerances of NCSX create some unique engineering...
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The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its...
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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...
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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
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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...
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The National Compact Stellarator Experiment (NCSX) is a quasi-axisymmetric plasma experiment that combines the high beta and good confinement 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...
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Experiments in Fluids > 2006 > 40 > 6 > 962-966
We demonstrate experimentally that independently rotating intermediary end-rings between the cylinders of a Taylor–Couette apparatus can be utilized to reduce friction-driven secondary flow, i.e. Ekman circulation. This allows for velocity profiles in a device of small aspect ratio to be less constrained by ‘end effects’, so that the resulting wide-gap flows can be made to have a radial distribution...
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Fusion Engineering and Design > 2006 > 80 > 1-4 > 3-23
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 5.2m, a minor radius of 1.3m, a toroidal...
