C. Neumeyer
A5032639071- Magnetic confinement fusion research
- Superconducting Materials and Applications
- Particle accelerators and beam dynamics
- Fusion materials and technologies
- Power Systems and Technologies
- HVDC Systems and Fault Protection
- Plasma Diagnostics and Applications
- Ionosphere and magnetosphere dynamics
- Nuclear reactor physics and engineering
- Silicon Carbide Semiconductor Technologies
- Advanced DC-DC Converters
- Spacecraft and Cryogenic Technologies
- Multilevel Inverters and Converters
- High-Voltage Power Transmission Systems
- Mobile Agent-Based Network Management
- Robotic Mechanisms and Dynamics
- Power Systems Fault Detection
- Electromagnetic Launch and Propulsion Technology
- Advanced Electrical Measurement Techniques
- Electrical Fault Detection and Protection
- Combustion and Detonation Processes
- Electric Power Systems and Control
- Planetary Science and Exploration
- Electrical Contact Performance and Analysis
- Advanced Data Storage Technologies
Princeton Plasma Physics Laboratory
2010-2021
Princeton University
2003-2021
Higashihiroshima Medical Center
2006
Hiroshima University
2006
Lawrence Livermore National Laboratory
2006
University of California, Irvine
2006
University of Washington
2001-2006
University of Rochester
2006
University of California, Los Angeles
2006
University of California, Davis
2006
The National Spherical Torus Experiment (NSTX) is being built at Princeton Plasma Physics Laboratory to test the fusion physics principles for spherical torus concept MA level. NSTX nominal plasma parameters are R0 = 85 cm, a 67 R/a ⩾ 1.26, Bt 3 kG, Ip 1 MA, q95 14, elongation κ ⩽ 2.2, triangularity δ 0.5 and pulse length of up 5 s. heating/current drive tools high harmonic fast wave (6 MW, s), neutral beam injection (5 80 keV, s) coaxial helicity injection. Theoretical calculations predict...
As part of the ITER Design Review and in response to issues identified by Science Technology Advisory Committee, physics requirements were reviewed as appropriate updated. The focus this paper will be on recent work affecting design with special emphasis topics near-term procurement arrangements. This describe results on: sensitivity studies, poloidal field coil requirements, vertical stability, effect toroidal ripple thermal confinement, material choice heat load for plasma-facing...
Abstract The spherical tokamak (ST) is a leading candidate for Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. National Spherical Torus eXperiment (NSTX) MA-class ST facility in the US actively developing physics basis an ST-based FNSF. In plasma transport research, experiments exhibit strong (nearly inverse) scaling of normalized confinement with collisionality, if this trend holds at low high fusion neutron fluences could be achievable very...
Abstract A fusion nuclear science facility (FNSF) could play an important role in the development of energy by providing environment needed to develop materials and components. The spherical torus/tokamak (ST) is a leading candidate for FNSF due its potentially high neutron wall loading modular configuration. key consideration choice configuration range achievable missions as function device size. Possible include: fluence, demonstrating tritium self-sufficiency, electrical self-sufficiency....
A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. device ultimately capable of small net electricity production in as compact facility possible and configuration scalable to full-size power plant. key capability for pilot-plant programme the high neutron fluence enabling nuclear science technology (FNST) research. It found that physics assumptions between those assumed ITER nth-of-a-kind it provide FNST-relevant wall loading devices. Thus, may be...
In this paper, firstly, ITER power supply and its AC/DC converter in 2001 design have been introduced. Some import criteria such as FSC internal bypass are reviewed researched. The system overvoltage low frequency oscillation also studied. Then some proposals to improve the baseline presented, these validated supported by scientists experts from two Expert Groups organized IO.
The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order advance spherical torus (ST) concept. In do this, NSTX utilizes up 7.5 MW neutral beam injection, 6 harmonic fast waves (HHFWs), it operates with plasma currents 1.5 MA elongations 2.6 a field 0.45 T. New facility, diagnostic modelling capabilities developed over past two years have enabled research team make significant progress...
AbstractThe use of a fusion component testing facility to study and establish, during the ITER era, remaining scientific technical knowledge needed by Demo is considered described in this paper. This aims test components an integrated nuclear environment, for first time, discover understand underpinning physical properties, develop improved further testing, time-efficient manner. It requires design with extensive modularization remote handling activated components, flexible hot-cell...
The ITER Power Supply it will be the largest ever built in terms of power, pulse length, and energy capacity. It also responsible for fast discharge superconducting magnets whose storage at an unprecedented scale. Nearly all components that comprise system unique, custom designed items exceed prior state art. This paper describes power supply its with emphasis on extrapolation scale technology compared to TFTR/JET/JT-60/T-15 era large tokamaks normal (copper) as well present fleet (EAST,...
AbstractThe ITER project baseline now includes two sets of in-vessel coils, one to mitigate the effects Edge Localized Modes (ELMs) and another provide vertical stabilization (VS). The location presents special challenges in terms nuclear radiation temperature, requires use mineral-insulated conductors. An update preliminary design based on this conductor technology is presented for both coil designs. Results from an on-going R&D program consisting development, welding brazing process...
A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, confinement operation and its physics basis. This has been enabled by facility capabilities developed during 2001 2002, including neutral beam (up to 7 MW) harmonic fast wave (HHFW) heating 6 MW), toroidal fields up kG, plasma currents 1.5 MA, flexible shape control, wall preparation techniques. These have generation plasmas with 35%. Normalized beta values often exceed no-wall limit,...
The main aim of the National Spherical Torus Experiment (NSTX) is to establish fusion physics principles spherical torus (ST) concept. NSTX device began plasma operations in February 1999 and current Ip was successfully brought up design value 1 MA on 14 December 1999. planned shaping parameters, elongation κ = 1.6-2.2 triangularity δ 0.2-0.4, were achieved inner wall limited, single null double diverted configurations. coaxial helicity injection (CHI) high harmonic fast wave (HHFW)...
The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER.Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents 0.7 MA with non-inductive fractions above 65% while simultaneously achieving β T N values 17% 5.7 (%m -1 ), respectively.A newly available motional Stark effect diagnostic...
The ITER project is considering the inclusion of two sets in-vessel coils, one to mitigate effect Edge Localized Modes (ELMs) and another provide vertical stabilization (VS). location (behind blanket shield modules, mounted vacuum vessel inner wall) presents special challenges in terms nuclear radiation (˜3000 MGy) temperature (100 °C during operations, 200 bakeout). Mineral insulated conductors are well suited this environment but not commercially available large cross section required. An...
The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including stability global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized (ELMs), bootstrap current drive, divertor flux expansion, heat transport. Improved capability been crucial to achieving βt∼40%. Precise shape control achieved NSTX using real-time equilibrium reconstruction. simultaneously...
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 VS Coil requirements strong coupling with plasma is required so that it necessary for coils to be installed in vessel just behind blanket shield modules. Due this close proximity radiation temperature environment severe conventional electrical insulation materials processes cannot used. The development mineral insulated conductor...
The Energy Return on Investment (EROI) is an important measure of the energy gain electrical power generating facility that typically evaluated based life cycle balance a single facility. EROI concept can be extended to cover collection facilities comprise complete system and used assess expansion evolution as it transitions from one portfolio mix technologies another over time. In this study we develop dynamic model simulates perform simulation electricity production scenarios developed...
ITER is a representative prototype of fusion power plant. It contains most the features and systems that are anticipated in reactor will produce substantial (500 MW) for long periods time (100's seconds). Therefore its electrical loads generally those which exist future facilities such as pilot plant, DEMO, beyond. In this paper energy demand by PPEN SSEN be described, along with discussion their effects on grid. The consumed analyzed an accounting load attributable to each subsystem during...
The National Spherical Torus Experiment-Upgrade (NSTX-U) is the most powerful spherical torus facility at PPPL, Princeton USA. major mission of NSTX-U to develop physics basis for an ST-based Fusion Nuclear Science Facility (FNSF). FNSF has promise achieving high neutron fluence needed reactor component testing with relatively modest tritium consumption. At same time, unique operating regimes can contribute several important issues in burning plasmas optimize performance ITER. further aims...
A spherical torus (ST) fusion energy development path which is complementary to the proposed tokamak burning plasma experiments such as ITER described. The ST strategy focuses on a compact component test facility (CTF) and high performance advanced regimes leading more attractive Demo power plant scale reactors. To provide physical basis for CTF an intermediate step needs be taken, we refer 'next-step torus' (NSST) device examine in some detail herein. NSST 'performance extension' stage with...
ELM mitigation is of particular importance in ITER order to prevent rapid erosion or melting the divertor surface, with consequent risk water leaks, increased plasma impurity content and disruptivity. Exploitable "natural" small no regimes might yet be found which extrapolate but this cannot depended upon. Resonant Magnetic Perturbation has been added pellet pacing as a tool for mitigate ELMs. Both are required, since neither method fully developed much work remains done. In addition,...
The National Spherical Tokamak Experiment (NSTX) is an ultra low aspect ratio device with a plasma current of 1 MA. tokamak features auxiliary heating and drive close-fitting conducting shell to maximize the pressure. NSTX designed for experimental pulse length that will demonstrate quasi-steady state non-inductively driven advanced operation. design also takes maximum advantage existing facilities components from previous Princeton devices reduce overall program costs.
Research on the spherical torus (or tokamak) (ST) is being pursued to explore scientific benefits of modifying field line structure from that in more moderate aspect ratio devices, such as conventional tokamak. The ST experiments are conducted various US research facilities including MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized facilities: PEGASUS University Wisconsin, HIT-II Washington, CDX-U Princeton. In context fusion energy development path...
In order to reduce recirculating power fraction acceptable levels, the spherical torus concept relies on simultaneous achievement of high toroidal β and bootstrap in steady state. last year, as a result plasma control system improvements, achievable elongation NSTX has been raised from κ ∼ 2.1 2.6—approximately 25% increase. This increase led substantial for long pulse discharges. The is associated with an current at nearly fixed poloidal β, which enables higher βt constant fraction. As...