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Use of a large-area liquid lithium surface as a limiter has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter.

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.

Various perfluorinated materials are used at the Princeton Plasma Physics Laboratory in support of the Deuterium-Tritium experimental program on the Tokamak Fusion Test Reactor (TFTR). For example, SF{sub 6} is used as a high dielectric gaseous insulator in the Neutral Beam sources, and Krytox, a perfluorinated polyether, is used as a lubricant in vacuum pumping systems. Each of these materials is robust and stable in the applications for which they are designed but may be a source of trouble when used in tritium systems. This paper reports on the observations made and experience gained operating tritium systems under conditions which degrade these perfluorinated materials. The possible degradation mechanisms and products are described, and the effect on the equipment and instrumentation is described. These observations have led to the conclusion that under certain circumstances perfluorinated materials are not suited for tritium service because of the degradation products from tritium decay and/or process conditions. 11 refs., 1 fig., 1 tab.

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...

During lower hybrid current drive on tokamaks, suprathermal electrons are generated that emit hard x‐ray bremsstrahlung. A pinhole camera has been installed on the PBX‐M tokamak that records 128×128 pixel images of the hard x‐ray plasma bremsstrahlung emitted in the 30–200 keV photon energy range. The camera has a temporal resolution of 3 ms and a spatial resolution of 3–4 cm in the plasma. The detector is a 230‐mm‐diam hard x‐ray image intensifier. This instrument has identified hollow radiation profiles on PBX‐M, indicating off‐axis current drive. A detailed account of the construction of the hard x‐ray camera, its operation, and its performance is given.

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 requirements. A test program was conducted to verify the design. The tests included measurements of flatness of the spacers, determination of contact area, torque vs. tension of the studs and jack nuts, friction coefficient tests on the alumina and G-10 insulators, electrical tests, and tension relaxation tests due to temperature excursions from room temperature to liquid nitrogen temperatures. This paper will describe the design and the results of the test program.

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.

A conceptual design has been developed for the recovery of un-expended fuel, ash, and associated post-detonation products from a ~ 2 GW IFE power reactor. The conceptual design incorporates systems for the safe, efficient collection, processing, and purification of IFE plasma exhaust fuel components. The system has been designed and sized such that tritium bred within blankets can also be collected, processed, and introduced into the fuel cycle. The system is nominally sized to process ~2 kg of tritium per day and is designed to link directly to the target chamber mechanical pumping system. The plasma exhaust can be directly processed from the exhaust of the vacuum pumping system or can be processed in batch mode from buffer vessels in the receiving and analysis system (RAS). Systems for the accurate measurement of material in-process (MIP) have been included. Design emphasis is on safety, reliability, redundancy, and efficiency in order to maximize availability. The primary goal of the fuel recovery system (FRS) design is to economically recycle components of IFE fuel back to the target manufacturers in a fashion by which fuel components are rapidly made available for re-use thus lowering the total active inventory. The FRS design is presented as a facility sub-system in the context of supporting the safe and efficient operation of the IFE target chamber.

A conceptual design for a magnetic intervention system is presented in support of a 2 GW IFE direct drive fusion power reactor. The system is designed employing a cusp field to deflect ions generated by an IFE implosion away from the first wall of the reactor core and into specifically designed ion dumps. The magnetic coil system will employ liquid helium cooled 5083 Aluminum alloy casing on a Rutherford NbTi cable. The cables are configured as four double pancakes with a 5083 Aluminum alloy case for structural support. The conceptual design and corresponding preliminary load and field calculations will be presented.

The lithium pellet injector (LPI) is being developed for the National Spherical Torus Experiment (NSTX). The LPI will inject 'pellets' of various composition into the plasma in order to study wall conditioning, edge impurity transport, liquid limiter simulations, and other areas of research. The control system for the NSTX LPI has incorporated widely used advanced technologies, such as LabVIEW/spl trade/ and PCI bus I/O boards, to create a low-cost control system which is fully integrated into the NSTX computing environment. This paper will present the hardware and software design of the computer control system for the LPI.

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 National Laboratory (ORNL) to test the physics of a high-beta compact stellarator with a low-ripple, tokamak-like magnetic configuration. The engineering challenges of NCSX stem from its complex geometry requirements. These issues are addressed in the construction project through manufacturing R&D and system engineering. As a result, the fabrication of the coil winding forms and vacuum vessel are proceeding in industry without significant technical issues, and preparations for winding the coils at PPPL are in place. Design integration, analysis, and dimensional control are functions provided by system engineering to ensure that the finished product will satisfy the physics requirements, especially accurate realization of the specified coil geometries. After completion of construction in 2009, a research program to test the expected physics benefits will start

A supersonic gas injector (SGI) has been developed for fueling and diagnostic applications on the National Spherical Torus Experiment (NSTX). It is comprised of a graphite converging-diverging Laval nozzle and a commercial piezoelectric gas valve mounted on a movable probe at a low field side midplane port location. Also mounted on the probe is a diagnostic package: a Langmuir probe, two thermocouples and five pick-up coils for measuring toroidal, radial, vertical magnetic field components and magnetic fluctuations at the location of the SGI tip. The SGI flow rate is up to 4times1021 particles/s, comparable to conventional NSTX gas injectors. The nozzle operates in a pulsed regime at room temperature and a reservoir gas pressure up to 0.33 MPa. The deuterium jet Mach number of about 4, and the divergence half-angle of 5deg - 25deg have been measured in laboratory experiments simulating NSTX environment. In initial NSTX experiments reliable operation of the SGI and all mounted diagnostics at distances 1-20 cm from the plasma separatrix has been demonstrated. The SGI has been used for fueling of ohmic and 2-4 MW NBI heated L- and H-mode plasmas. Fueling efficiency in the range 0.1-0.3 has been obtained from the plasma electron inventory analysis

Use of a large-area liquid lithium surface as a first wall has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter.

In advanced tokamak research involving complex magnetic geometries, midplane visible plasma TV images can indicate the need for tokamak control corrections not evident from other diagnostics. Suitable midplane viewports are often unavailable. The PBX‐M plasma TV system uses four tangential midplane vacuum viewports. One tangential vacuum viewport is on a standard re‐entrant midplane port flange. The other three vacuum viewports are tangentially aligned along the inside vessel wall and connected to port flanges 40 cm from the midplane via 60‐cm‐long re‐entrant bellows. Fiber optic bundles transmit images from the re‐entrant viewports through the off‐midplane ports and out to TV cameras positioned beyond the TF coils. The VCR recorded images from each discharge were used to obtain information not available from other diagnostics.

Summary form only given, as follows. This report describes the engineering design of diagnostic cassette and its optical configuration for bay J for KSTAR. Bay J dedicates to a visible optical diagnostic including visible survey spectrometer, H/spl alpha/gonitor, visible Bremsstrahlung array and visible/H/spl alpha/ TV. This cassette design will be a model case as prototype for some other similar port and can be easily modified to a dedicated diagnostics. This report also describes the principal features of the optical lens design and provides some of the calculations for optical transmission made in defining the main components of the optical design.

A pellet injector suitable for the injection of lithium and other low-Z pellets of varying mass into plasmas at precise velocities from 5 to 500 m/s is being developed for use on NSTX. The ability to inject low-Z impurities will significantly expand NSTX experimental capability for a broad range of diagnostic and operational applications. The architecture employs a pellet-carrying cartridge propelled through a guide tube by deuterium gas. Abrupt deceleration of the cartridge at the end of the guide tube results in the pellet continuing along its intended path, thereby giving controlled reproducible velocities for a variety of pellets materials and a reduced gas load to the torus. The planned injector assembly has four hundred guide tubes contained in a rotating magazine with eight tubes provided for injection into plasmas. A PC-based control system is being developed as well and will be described elsewhere in these Proceedings. The development path and mechanical performance of the injector will be described.

The Tokamak Fusion Test Reactor (TFTR) operated for 15 years, from 1982 to 1997. From November 1993 to April 1997 a mixture of deuterium-tritium (D-T) was used to fuel experiments, leaving the reactor system in an activated and contaminated state. In this condition, the removal of the TFTR and its subsystems is greatly complicated. In 1999, PPPL commenced the TFTR Decontamination and Decommissioning (D&D) Project. The objectives of the D&D projects are to completely remove the TFTR in a safe, efficient and cost effective manner. To date all diagnostics, structural components and ancillary systems have been removed, and the tokamak is in the process of being sectioned into smaller pieces. This paper will describe the criteria, the methods employed and the successes of the D&D Project.

The Tokamak Fusion Test Reactor (TFTR) has twenty toroidal-field (TF) coils with a maximum current rating of 73 kA for 3.6 s when cooled with water. Operation of the TF array at this maximal current produces a 5.2-T magnetic field at a major radius of 2.48 m. Repetition rate is dependent upon the efficiency of the cooling system for the array. An alternative approach to the water cooling would be to circulate a silicate ester (oil) in the coils. This technique would improve the electrical insulation of a leaking coil because, unlike water, the oil will not dissolve electrolytes and, therefore, retains a high dielectric strength as it wets the insulation. The oil cooling system is composed of a pump, heat exchanger and piping, and associated control and protective action circuits. Due to higher viscosity and lower heat capacity than water, oil removes 0.89 MW (maximum) at 12 L/s for each coil (1.4 MW at 12 L/s for water). This decreases the repetition rate to one maximum power pulse every 900 s, as opposed to one every 300 s for water. A thermal analysis is presented, and the equipment and controls are described. A flow diagram for the system is provided. >

Recent experimental results in the fusion community have pointed to possible directions for improvement of the tokamak concept. A critical issue for many advanced tokamak scenarios is the optimization of the spatial distribution of plasma current density. Recent experiments on the Tokamak Fusion Test Reactor and Doublet III-D have demonstrated enhanced confinement and improved stability using transient techniques for the modification of the plasma current profile. The Princeton Beta Experiment-Modification (PBX-M) has begun experiments using profile modification techniques that are extrapolable to long-pulse or steady-state operation. PBX-M uses lower hybrid current drive for current profile modification, ion Bernstein wave heating for pressure profile variation, and plasma shaping to alter the magnetic geometry. This paper is an overview of the PBX-M program, its engineering facilities, and discussion of how this program is relevant to future advanced tokamak programs such as the Tokamak Physics Experiment.

The PBX-M passive plate stabilization system provides significant stabilization of long-wavelength external kink modes, the slowing of vertical instability growth rates, and the amelioration of disruption characteristics. The passive plate stabilization system has allowed the use of LHCD and IBW to induce current density and pressure profile modifications, and m=1 divertor biasing for modifying edge plasma transport. Improvements in the passive plate system insulators and support structures have provided reliable operation. Impurity influxes with the close-fitting passive plates are low. Solid target boronization is applied routinely to reduce conditioning time and maintain clean conditions.

PBX-M upgrades have been in progress for enhanced capability and reliability. These include upgrades to the vessel O-ring system for impurity control, the Ion Bernstein Wave and Lower Hybrid Current Drive Systems for plasma profile control, a double viewing upper and lower CHERS based Poloidal Rotation Diagnostic, and a new digital plasma control system.

Observation of core transport barrier formation (for particles, ion and electron energies, and toroidal momentum) by ion Bernstein wave heating (IBWH) in PBX-M plasma is reported. The formation of a transport barrier leads to a strong peaking and significant increase of the core pressure (70%) and toroidal momentum (20%), and has been termed the core-high confinement mode (CH-Mode). This formation of a transport barrier is consistent, in terms of the expected barrier location as well as the required threshold power, with a theoretical model based on the poloidal sheared flow generation by the ion Bernstein wave power. The use of ion Bernstein wave (IBW) induced sheared flow as a tool to control plasma pressure and bootstrap current profiles shows a favorable scaling for the use in future reactor grade tokamak plasmas.

Lower hybrid current drive (LHCD) is being explored as a means to control the current density profile on PBX‐M with the goal of raising the central safety factor q(0) to values of 1.5–2 to facilitate access to a full‐volume second stable regime. Initial experiments have been conducted with up to 400 kW of 4.6 GHz LH power in circular and indented plasmas with modest parameters. A tangential‐viewing two‐dimensional hard x‐ray imaging diagnostic has been used to observe the bremsstrahlung emission from the suprathermal electrons generated during LHCD. Hollow hard x‐ray images have indicated off‐axis localization of the driven current. A serious obstacle to the control of the current density profile with LHCD is the concomitant generation of MHD activity, which can seriously degrade the confinement of suprathermal electrons. By combining neutral beam injection with LHCD, an MHD‐free condition has been obtained where q(0) is raised above 1.

Understanding the wave damping mechanism in the presence of a `spectral gap` is an important issue for the current profile control using Lower Hybrid Current Drive (LHCD). The authors examine a traditional explanation based upon upshifting of the wave parallel refractive index (n{sub {parallel}}) and find that there can be an upper bound in the n{sub {parallel}} upshift. The amount of upshift is not sufficient to bridge the spectral gap completely under some PBX-M LHCD conditions. There is experimental evidence, however, that current was driven even under such conditions. Another mechanism is also considered, based upon the 2-D velocity space dynamics coupled with a compound wave spectrum, here consisting of forward- and backward-running waves. The runaway critical speed relative to the phase speeds of these waves plays an important role in this model.

A diborane gas handling system has been constructed for the purposes of boronizing TFTR. Numerous safety features have been incorporated to mitigate the toxic and inflammable nature of diborane, including many features adopted from the semiconductor industry. The carborane films deposited in TFTR have significantly improved the operational performance of the machine.

It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: (1) Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure, and (2) aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a){approx}3 at {approx}250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H{sub 2}O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of X-rays from runaway electrons, is an indication that TDC is no longer necessary after {approx}10{sup 5} pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150 {degree}Cmore » to about 250 {degree}C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H{sub 2}O is removed from the graphite limiter.« less

The modification and control of the plasma current profile by lower hybrid current drive is considered one of the most promising avenues to control MHD instabilities and improve the high beta performance of tokamaks. The Hard X‐ray Camera on PBX‐M records images of the bremsstrahlung emitted by suprathermal electrons that carry part of the plasma current during lower hybrid current drive. The Hard X‐ray Camera is a pinhole camera that uses a 9 in. diam x‐ray intensifier as detector. It provides a tangential view of the tokamak plasma and delivers 64 images with 128×128 pixels per plasma shot. The spatial resolution in the plasma is approximately 2 cm, and the time resolution 3 ms. The camera is sensitive to x rays between 30 and 200 keV. The photon energy (and the energy of the suprathermal electrons) is measured with an absorber foil method. Modeling of the x‐ray emission has reproduced the experimental images with a better than 10% accuracy. In addition, two inversion techniques have been developed: one...

It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure and aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a) /approx/ 3 at /approx/ 250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H/sub 2/O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of x-rays from runaway electrons, is an indication that TDC is no longer necessary after /approx/ 10/sup 5/ pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature ofmore » 150/degree/C to about 250/degree/C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H/sub 2/O is removed from the graphite limiter. 14 refs., 3 figs., 1 tab.« less