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Plasma control on the National Spherical Torus Experiment (NSTX) was previously accomplished using eight 333 MHz G4 processors built by Sky computers. Several planned improvements and additional control algorithms required significant upgrades to our real-time control computers and real-time data acquisition infrastructure. Several in-house modules have been designed and implemented including: the digital time stamp module (DITS) and for digital/analog front panel data port (FPDP) output, the FPDP output module digital/analog (FOMD/A). Standard Linux based Intel computers perform the real-time control tasks and InfiniBand as been employed for communication between a user-accessible “host” server and the real-time computer. In addition to several independent real-time processes the General Atomics developed PCS (Bell (2006) [1] ) system infrastructure continues to be used on NSTX. While maintaining previous functionality, improvements in the control system software include: an RWM feedback algorithm, beta feedback NBI control, more comprehensive error logging and trapping, more user-friendly interface, more complete archiving and restoring functionality, and better status reporting and diagnostic tools. Once completed, we succeeded in increasing overall plasma stability and decreasing control system latency by several times.

Small electrodes were installed in the outer-midplane edge of NSTX to attempt to control the local width of the scrape-off layer (SOL) by creating an outward E(pol)xB flow. When the applied voltage between electrodes was 90 V, the density between these electrodes increased by a factor of 3-10 over a radial width of similar to 4 cm. Thus a local control of the SOL plasma density was obtained. (C) 2009 Elsevier B.V. All rights reserved.

In 2003, the National Spherical Torus Experiment (NSTX) plasma control system was used for plasma shape control using real-time equilibrium reconstruction (using the rtEFIT code). rtEFIT is now in routine use for plasma boundary control. More recently, the system has been upgraded to support feedback control of the resistive wall mode (RWM). This paper describes the hardware and software improvements that were made in support of these physics requirements. The control computer is an eight processor embedded system. The real-time data acquisition system now acquires 352 channels of data at 5 kHz for each NSTX plasma discharge. The latency for the data acquisition, which uses the Front Panel Data Port (FPDP) protocol, is measured to be ∼8 μs. A Stand-Alone Digitizer (SAD), designed at PPPL, along with an FPDP Input Multiplexing Module (FIMM) allows for simple modular upgrades. An interface module was built to interface between the FPDP output of the NSTX control system and the legacy Power Conversion Link (PCLink) used for communicating with the PPPL power supplies (first used for TFTR). Additionally a module has been built for communicating with the switching power amplifiers (SPA) recently installed on NSTX. In addition to the hardware developments, the control software on the NSTX control system has been upgraded. The device driver software for the hardware described above will be discussed, as well as the new control algorithms that have been developed to control the switching power supplies for RWM control. A reliable mode detection algorithm for RWM feedback is currently under development.

In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ∼ 2.1 to κ ∼ 2.6 - approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βtwith nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented. © 2006 IAEA, Vienna.

Plasma shape control using real-time equilibrium reconstruction has been implemented on the National Spherical Torus Experiment (NSTX). The rtEFIT code originally developed for use on DIII-D was adapted for use on NSTX. The real-time equilibria provide calculations of the flux at points on the plasma boundary, which are used as input to a shape control algorithm known as isoflux control. The flux at the desired boundary location is compared with a reference flux value, and this flux error is used as the basic feedback quantity for the poloidal field coils on NSTX. The hardware that comprises the control system is described, as well as the software infrastructure. Examples of precise boundary control are also presented.

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 to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

A simple analog circuit that measures the National Spherical Torus Experiment (NSTX) axisymmetric eddy current distribution has been designed and constructed. It is based on simple circuit model of the NSTX vacuum vessel that was calibrated using a special axisymmetric eddy current code which was written so that accuracy was maintained in the vicinity of the current filaments [J. Menard, J. Fusion Tech. (to be published)]. The measurement and the model have been benchmarked against data from numerous vacuum shots and they are in excellent agreement. This is an important measurement that helps give more accurate equilibrium reconstructions.

National Spherical Torus Experiment (NSTX), like other research programs, is facing an inevitable crisis due to end-of-life issues for its 20-year-old CAMAC instrumentation. In many cases, replacement components are not available, effectively rendering a CAMAC module unusable after a failure. The proliferation of high performance, reliable, low-cost commodity computing hardware and software based on industry standard technology can provide the basis for a new generation of instrumentation. At NSTX, there have been several advances towards developing a PCI-based model for data acquisition and control systems. New hardware developments include a High Performance Signal Conditioning board and an Field Programmable Gate Array (FPGA)-based Multifunction Timing System (MTS). Extensible software interfaces have been developed to integrate these boards into the NSTX computing environment [P. Sichta, J. Dong, G. Oliaro, P. Roney ‘Overview of the NSTX Control System,’ 8th International Conference on Accelerator and Large Experimental Physics Control Systems, San Jose, CA (2001)]. This paper will illustrate these developments and how they could be used to benefit collaborative fusion research.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (?), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500?kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma.The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral‐beam heating, in a supershot discharge and 6.7 MW in a high‐βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter‐H‐mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high‐βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first‐orbit loss mechanism. No loss due to alpha‐particle‐driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fullymore » explored.« less

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium–deuterium (D–D) run to simulate expected deuterium–tritium (D–T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D–T experiments using D–D have been performed. The physics objectives of D–T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium–tritium plasmas, evaluation of α‐particle heating of electrons, and collective effects driven by alpha particles and testing of diag...

The first real-time detection of surface dust inside a tokamak was made using an electrostatic dust detector. A fine grid of interlocking circuit traces was installed in the NSTX vessel and biased to 50 V. Impinging dust particles created a temporary short circuit and the resulting current pulse was recorded by counting electronics. The techniques used to increase the detector sensitivity by a factor of ×10 000 to match NSTX dust levels while suppressing electrical pickup are presented. The results were validated by comparison to laboratory measurements, by the null signal from a covered detector that was only sensitive to pickup, and by the dramatic increase in signal when Li particles were introduced for wall conditioning purposes.

The first real-time detection of surface dust inside a tokamak was made using an electrostatic dust detector. A fine grid of interlocking circuit traces was installed in the NSTX vessel and biased to 50 v. Impinging dust particles created a temporary short circuit and the resulting current pulse was recorded by counting electronics. The techniques used to increase the detector sensitivity by a factor of x10,000 to match NSTX dust levels while suppressing electrical pickup are presented. The results were validated by comparison to lab measurements, by the null signal from a covered detector that was only sensitive to pickup, and by the dramatic increase in signal when Li particles were introduced for wall conditioning purposes.

In order to provide improved and expanded experimental capabilities, the existing Transrex power supplies at PPPL are to be upgraded and modernized. Each of the 39 power supplies consists of two six pulse silicon controlled rectifier sections forming a twelve pulse power supply. The first modification is to split each supply into two independent six pulse supplies by replacing the existing obsolete twelve pulse firing generator with two commercially available six pulse firing generators. The second change replaces the existing control link with a faster system, with greater capacity, which will allow for independent control of all 78 power supply sections. The third change replaces the existing Computer Automated Measurement and Control (CAMAC) based fault detector with an Experimental Physics and Industrial Control System (EPICS) compatible unit, eliminating the obsolete CAMAC modules. Finally the remaining relay logic and interfaces to the “Hardwired Control System” will be replaces with a Programmable Logic Controller (PLC).

NSTX at Princeton Plasma Physics Laboratory (PPPL) requires sophisticated plasma positioning control system for stable plasma operation. TF magnetic coils and PF magnetic coils provide electromagnetic fields to position and shape the plasma vertically and horizontally respectively. NSTX utilizes twenty six coil power supplies to establish and initiate electromagnetic fields through the coil system for plasma control. A power protection and interlock system is utilized to detect power system faults and protect the TF coils and PF coils against excessive electromechanical forces, overheating, and over current. Upon detecting any fault condition the power system is restricted, and it is either prevented from initializing or suppressed to de-energize coil power during pulsing. Power fault status is immediately reported to the computer system. This paper describes the design and operation of NSTX's protection and interlocking system and possible future expansion plans.

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physic Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. The NSTX power system has eleven (11) circuits. Each circuit is a floating DC system. Ground detection and protection device is provided for each circuit. The TFTR ground detection scheme had to be modified for the new application. The coaxial helicity injection system (CHI) requires special scheme to detect grounds in the main power loop since the circuit can be back fed from during neutral beam injection. For CHI, the vacuum vessel is divided into a) inner vessel and b) outer vessel, insulated from one another. Special ground protection devices are being designed for the CHI. An additional system to detect grounds and generate alarms while performing maintenance on the vacuum vessel has also been installed. This was necessary to maintain the insulation integrity of the vessel. Also techniques used to trouble shoot vessel grounds are outlined. This paper describes the changes and addition to the electrical power circuit ground detection and monitoring for the NSTX magnet systems and CHI electrodes.

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. Until 2004, the NSTX power system was feeding twelve (12) circuits in the machine. In 2004, resistive wall mode (RWM) coils were installed in the machine to suppress resistive wall modes and to correct error fields. There are six of these coils installed around the machine on the mid-plane. Since these coils need fast and accurate controls, a switching power amplifier (SPA) has been procured, installed and commissioned along with other circuit components. One of the existing thyristor rectifiers is used as the DC source to the SPA. The controls for the RWM have been integrated into the overall computer control of the DC power systems for NSTX. This paper describes the RWM power supply for NSTX

A simple analog circuit that measures the NSTX axisymmetric eddy current distribution has been designed and constructed. It is based on simple circuit model of the NSTX vacuum vessel that was calibrated using a special axisymmetric eddy current code which was written so that accuracy was maintained in the vicinity of the current filaments. The measurement and the model have been benchmarked against data from numerous vacuum shots and they are in excellent agreement. This is an important measurement that helps give more accurate equilibrium reconstructions.

Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} 500 kA. In parallel, start-up using radio-frequency current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

The National Compact Stellarator Experiment (NCSX) is being designed to be built in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). The design of the NCSX electrical power system is tailored to suit the available infrastructure and electrical equipment on site. The equipment at PPPL is located primarily in two sites called C-Site and D-Site. D-Site houses the majority of electrical equipment. Most of the equipment being used for the currently operating National Spherical Torus Experiment (NSTX) located in D-Site and the earlier Fusion Experimental devices in C-Site, are proposed to be used with suitable modifications to reflect NCSX needs. The NCSX is proposed to be located in C- Site, based on the available Neutral Beam equipment and a Test Cell to house the device. Thus NSTX and NCSX will share some of the available power supplies. The number of circuits and the location of the NCSX device in the C-Site, drove the major changes in the power system hardware. Only the D-Site power supplies can meet the time response and energy requirements of NCSX. DC power will be transmitted from D-Site to C-Site. The NCSX has a total of nine circuits to be fed from D-Site. Three of the nine circuits are bipolar and the rest are unipolar. Two circuits known as Trim Coils will be fed from the available thyristor power supplies in C-Site. Additional power supplies are available for two more circuits. The hardware needs to be modified to address the requirements of NCSX power loops. The aim is to make the changes in the most cost-effective manner without compromising technical requirements. This paper describes the design changes and addition to the electrical power system components for the NCSX magnet system.

The NSTX is a new national facility for the study of plasma confinement, heating, and current drive in a low aspect ratio, spherical torus (ST) configuration. The ST configuration is an alternate magnetic confinement concept which is characterized by high /spl beta/ (ratio plasma pressure to magnetic held pressure) and low toroidal field compared to conventional tokamaks, and could provide a pathway to the realization of a practical fusion power source. The NSTX depends on a real time, high speed, synchronous, and deterministic control system acting on a system of thyristor rectifier power supplies to 1) establish the initial magnetic field configuration; 2) initiate plasma within the vacuum vessel; 3) inductively drive plasma current; and 4) control plasma position and shape.

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. The design of the NSTX electrical power system was tailored to suit the available infrastructure and electrical equipment on site. Components were analyzed to verify their suitability for use in NSTX.

The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program. The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, the first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations.

The NSTX is a new national facility for the study of plasma confinement, heating, and current drive in a low aspect ratio, spherical torus (ST) configuration. The ST configuration is an alternate magnetic confinement concept which is characterized by high /spl beta/ (ratio plasma pressure to magnetic field pressure) and low toroidal field compared to conventional tokamaks, and could provide a pathway to the realization of a practical fusion power source. The NSTX defends on a real time, high speed, synchronous, and deterministic control system acting on a system of thyristor rectifier power supplies to 1) establish the initial magnetic field configuration; 2) initiate plasma within the vacuum vessel; 3) inductively drive plasma current; and 4) control plasma position and shape.

In April of 1993 PPPL embarked upon the tritium operations phase of the TFTR project. Four years later, in the early morning of April 4, 1997, TFTR pulsed with tritium for the last time. In those four years, approximately 1 mega Curie of tritium was processed, and a Tritium Purification System (TPS) was commissioned. During this period >1,000 invasive (line break) operations were performed with

In TFTR plasmas heated by neutral beam injection, the fusion power yield increases rapidly with the plasma pressure. However, the pressure is limited by the onset of instabilities which may result in plasma disruptions that increase the risk of damage to internal components. The likelihood of disruption has been found to correlate with the normalized beta

The NSTX is a new national facility for the study of plasma confinement, heating, and current drive in a low aspect ratio, spherical torus (ST) configuration. The ST configuration is an alternate magnetic confinement concept which is characterized by high beta (ratio plasma pressure to magnetic field pressure) and low toroidal field compared to conventional tokamaks, and could provide a pathway to the realization of a practical fusion power source. The NSTX depends on a real time, high speed, synchronous, and deterministic control system acting on a system of thyristor rectifier power supplies to (1) establish the initial magnetic field configuration; (2) initiate plasma within the vacuum vessel; (3) inductively drive plasma current; and (4) control plasma position and shape. For the initial ``day 0'' 1st plasma operations (Feb. 1999), the system was limited to closed loop proportional-integral current control of the power supplies based on preprogrammed reference waveforms. For the ``day 1'' phase of operations beginning Sept. 1999 the loop has been closed on plasma current and position. This paper focuses on the Power Supply Real Time Controller (PSRTC).

A peak fusion power production of 9.3±0.7 MW has been achieved on the Tokamak Fusion Test Reactor (TFTR) in deuterium plasmas heated by co and counter injected deuterium and tritium neutral beams with a total power of 33.7 MW. The ratio of fusion power output to heating power input is 0.27. At the time of the highest neutron flux the plasma conditions are: Te(0)=11.5 keV, Ti(0)=44 keV, ne(0)=8.5×1019 m−3, and 〈Zeff〉=2.2 giving τE=0.24 s. These conditions are similar to those found in the highest confinement deuterium plasmas. The measured D‐T neutron yield is within 7% of computer code estimates based on profile measurements and within experimental uncertainties. These plasmas have an inferred central fusion alpha fraction of 0.2% and central fusion power density of 2 MW/m3 similar to that expected in a fusion reactor. Even though the alpha velocity exceeds the Alfven velocity throughout the time of high neutron output in most high power plasmas, MHD activity is not substantially different from that in co...

In TFTR plasmas heated by neutral beam injection, the fusion power yield increases rapidly with the plasma pressure. However, the pressure is limited by the onset of instabilities which may result in plasma disruptions that would have had an adverse effect on the performance of subsequent discharges and increase the risk of damage to internal components. The likelihood of disruption has been found to correlate with the normalized beta, defined as is the volume-average plasma perpendicular pressure, a the mid-plane minor radius of the plasma, BT the toroidal magnetic field and I P the plasma current. Other variables, such as the peaking of the plasma pressure and current profiles, have been found to influence the threshold of β N at which the probability of disruption begins to increase significantly. For TFTR plasmas with high fusion performance (TFTR “supershots”) the probability of disruption has been found to increase rapidly for β N > 1.8. Since confinement in this regime is affected by plasma-wall interaction, which can vary from shot to shot, operation at high β N with preprogrammed heating power pulses can produce an unaccept-ably high risk of disruption. To reduce the risk of producing beta-limit disruptions during neutral beam heating experiments, a control system, the Neutral Beam Power Feedback System (NBPFS), has been developed to modulate the total heating power by switching individual neutral beam sources on and off in response to the evolution of the normalized beta so that the limit will not be exceeded.

A dense array of 99 Langmuir probes has been installed in the lower divertor region of the National Spherical Torus Experiment (NSTX). This array is instrumented with a system of electronics that allows flexibility in the choice of probes to bias as well as the type of measurement (including standard swept, single probe, triple probe, and operation as passive floating potential and scrape-off-layer SOL current monitors). The use of flush-mounted probes requires careful interpretation. The time dependent nature of the SOL makes swept-probe traces difficult to interpret. To overcome these challenges, the single- and triple-Langmuir probe signals are used in complementary fashion to determine the temperature and density at the probe location. A comparison to midplane measurements is made.

An experiment on the National Spherical Torus Experiment (NSTX) was designed to test the theory that biased electrodes can affect the local scrape-off layer (SOL) by creating a strong radial E x B flow (Cohen and Ryutov 1997 Nucl. Fusion 37 621). These electrodes were located near the outer midplane and were biased at up to +/-90V with respect to the local vacuum vessel ground. This biasing caused large changes in the local SOL profiles as measured by an array of Langmuir probes between the electrodes. A theory is presented which at least partially describes the experimental results.

The high-speed rotation diagnostic developed for Columbia’s HBT-EP tokamak requires a high quantum efficiency, very low drift detector/amplifier combination. An updated version of the circuit developed originally for the beam emission spectroscopy experiment on TFTR is being used. A low dark current (2nA at 15V bias), low input source capacitance (2pF) FFD-040 N-type Si photodiode is operated in photoconductive mode. It has a quantum efficiency of 40% at the 468.6nm (HeII line that is being observed). A low-noise field-effect transistor (InterFET IFN152 with eNa=1.2nV∕√Hz) is used to reduce the noise in the transimpedance preamplifier (A250 AMPTEK op-amp) and a very high speed (unity-gain bandwidth=200MHz) voltage feedback amplifier (LM7171) is used to restore the frequency response up to 100kHz. This type of detector/amplifier is photon-noise limited at this bandwidth for incident light with a power of >∼2nW. The circuit has been optimized using SIMETRIX 4.0 SPICE software and a prototype circuit has bee...

The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including the stability of global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Improved shaping capability has been crucial to achieving Βt ∼40%. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved elongation κ∼2.8 and triangularity δ∼0.8. Ideal MHD theory predicts increased stability at high values of shaping factor S≡ q95 Ip (a Bt), which has been observed at large values of the S∼37 [MA (m·T)] on NSTX. The behavior of ELMs is observed to depend on plasma shape. A description of the ELM regimes attained as shape is varied will be presented. Increased shaping is predicted to increase the bootstrap fraction at fixed Ip. The achievement of strong shaping has enabled operation with 1 s pulses with Ip =1 MA, and for 1.6 s for Ip =700 kA. Analysis of the noninductive current fraction as well as empirical analysis of the achievable plasma pulse length as elongation is varied will be presented. Data are presented showing a reduction in peak divertor heat load due to increasing in flux expansion. © 2006 American Institute of Physics.

In D‐T plasmas, the understanding of the physics of confined α particles is extremely valuable for the future fusion plasma device. Among the various proposed α diagnostics, an X‐mode collective Thomson scattering system employing a high‐power gyrotron source (P≂200 kW, f=60 GHz, pulse length ≂0.5 s, and modulation frequency=10–25 kHz) is being designed for TFTR. The detailed description of the gyrotron source, transmission lines, optical designs, beam and viewing dump design, and receiver system will be presented in this paper. In particular, the test results of the beam and viewing dump indicate that the stray light can be reduced by 60 dB. The background emission level (∼20 eV) near 60‐GHz range during high Q discharge may also be reduced with beam and viewing dump further. The optical system is designed to measure the radial profile of α particles and to orient the incident wavevector (k0) to test the electromagnetic effects of the scattered spectrum. Prior to the study of α physics in D‐T plasmas, th...