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Magnetic perturbation measurements will be invaluable for characterizing Lithium Tokamak Experiment Beta (LTX-β) plasmas due to the time-evolving 3D nature of the magnetic fields generated by eddy currents in the vessel and copper shell segments, as well as enhanced MHD instability drive due to newly introduced neutral beam heating. The LTX-β upgrade includes two new arrays of Mirnov coils: a shell eddy sensor array of two-axis coils distributed over the back surface of one shell segment and a toroidal array of poloidal field coils at the low-field side midplane gap. Evaporative lithium wall-coating and the high temperatures required for liquid lithium wall operation both complicate the implementation of in-vessel diagnostics. While the shell array is protected from lithium exposure, the shell segment to which it is mounted will at times exceed 300 °C. The toroidal array, however, will experience direct line-of-sight exposure to the lithium evaporator as well as close proximity to the hot shell and may also be subject to poorly confined beam-driven fast ions. We describe how the two new Mirnov coil arrays meet these environmental challenges and enhance the LTX-β diagnostic suite.

Recent experiments on the National Spherical Torus eXperiment (NSTX) have shown the benefits of solid lithium coatings on carbon PFC's to diverted plasma performance, in both L- and H-mode confinement regimes. Better particle control, with decreased inductive flux consumption, and increased electron temperature, ion temperature, energy confinement time, and DD neutron rate were observed. Successive increases in lithium coverage resulted in the complete suppression of ELM activity in H-mode discharges. A liquid lithium divertor (LLD), which will employ the porous molybdenum surface developed for the LTX shell, is being installed on NSTX for the 2010 run period, and will provide comparisons between liquid walls in the Lithium Tokamak eXperiment (LTX) and liquid divertor targets in NSTX. LTX, which recently began operations at the Princeton Plasma Physics Laboratory, is the world's first confinement experiment with full liquid metal plasma-facing components (PFCs). All materials and construction techniques in LTX are compatible with liquid lithium. LTX employs an inner, heated, stainless steel-faced liner or shell, which will be lithium-coated. In order to ensure that lithium adheres to the shell, it is designed to operate at up to 500–600 °C to promote wetting of the stainless by the lithium, providing the first hot wall in a tokamak to operate at reactor-relevant temperatures. The engineering of LTX will be discussed.

The radiated-power-density diagnostic on the equatorial midplane for the NSTX-U tokamak will be upgraded to measure the radial structure of the photon emissivity profile with an improved radial resolution. This diagnostic will enhance the characterization and studies of power balance, impurity transport, and MHD. The layout and response expected of the new system is shown for different plasma conditions and impurity concentrations. The effect of toroidal rotation driving poloidal asymmetries in the core radiation from high-Z impurities is also addressed.

The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107 kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9 MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments.

TFTR performance has surpassed many of the previous tokamak records. This has been made possible by the use of tritium as fuel for DT plasma discharges. Stable operations of tritium systems provide for safe, routine DT operation of TFTR. In the preparation for DT operation, in the commissioning of the tritium systems and in the operation of the Nuclear Facility several key lessons have been learned. They include: the facility must take the lead in interpreting the applicable regulations and orders and then seek regulator approval; the use of ultra high vacuum technology in tritium system design and construction simplifies and enhances operations and maintenance; and central facility control under a single supervisory position is crucial to safely orchestrate operational and maintenance activities.

The lithium tokamak experiment (LTX) is a modest-sized spherical tokamak (R(0)=0.4 m and a=0.26 m) designed to investigate the low-recycling lithium wall operating regime for magnetically confined plasmas. LTX will reach this regime through a lithium-coated shell internal to the vacuum vessel, conformal to the plasma last-closed-flux surface, and heated to 300-400 °C. This structure is highly conductive and not axisymmetric. The three-dimensional nature of the shell causes the eddy currents and magnetic fields to be three-dimensional as well. In order to analyze the plasma equilibrium in the presence of three-dimensional eddy currents, an extensive array of unique magnetic diagnostics has been implemented. Sensors are designed to survive high temperatures and incidental contact with lithium and provide data on toroidal asymmetries as well as full coverage of the poloidal cross-section. The magnetic array has been utilized to determine the effects of nonaxisymmetric eddy currents and to model the start-up phase of LTX. Measurements from the magnetic array, coupled with two-dimensional field component modeling, have allowed a suitable field null and initial plasma current to be produced. For full magnetic reconstructions, a three-dimensional electromagnetic model of the vacuum vessel and shell is under development.

Use of a large-area liquid lithium limiter in the CDX-U tokamak produced the largest relative increase (an enhancement factor of 5–10) in Ohmic tokamak confinement ever observed. The confinement results from CDX-U do not agree with existing scaling laws, and cannot easily be projected to the new lithium tokamak experiment (LTX). Numerical simulations of CDX-U low recycling discharges have now been performed with the ASTRA-ESC code with a special reference transport model suitable for a diffusion-based confinement regime, incorporating boundary conditions for nonrecycling walls, with fuelling via edge gas puffing. This model has been successful at reproducing the experimental values of the energy confinement (4–6 ms), loop voltage (

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

A new set of radiated-power-density diagnostics for the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak have been designed to measure the two-dimensional poloidal structure of the total photon emissivity profile in order to perform power balance, impurity transport, and magnetohydrodynamic studies. Multiple AXUV-diode based pinhole cameras will be installed in the same toroidal angle at various poloidal locations. The local emissivity will be obtained from several types of tomographic reconstructions. The layout and response expected for the new radially viewing poloidal arrays will be shown for different impurity concentrations to characterize the diagnostic sensitivity. The radiated power profile inverted from the array data will also be used for estimates of power losses during transitions from various divertor configurations in NSTX-U. The effect of in-out and top/bottom asymmetries in the core radiation from high-Z impurities will be addressed.

The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980`s The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America`s first operational D-T fusion reactor. This paper will discuss negative pressure employing `elephant trunks` in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and {Delta} pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs.

An important issue for future tokamaks is real‐time plasma control for the avoidance of magnetohydrodynamic instabilities and other applications that require detailed plasma profile and fluctuation data. Although measurements from diagnostics providing this information require significantly more processing than magnetic flux data, recent advancements could make them practical for adjusting operational settings for plasma heating and current drive systems as well as field coil currents. On the Princeton Beta Experiment‐Modification (PBX‐M), the lower hybrid current drive phasing can be varied during a plasma shot using digitally programmable ferrite phase shifters, and neural beam functions can be fully computer controlled. PBX‐M diagnostics that may be used for control purposes include motional Stark‐effect polarimetry for magnetic field pitch angle profiles, soft x‐ray arrays for plasma position control and the separation of βp from li, hard x‐ray detectors for energetic electron distributions, a multich...

The Lithium Tokamak eXperiment is a small, low aspect ratio tokamak [Majeski et al., Nucl. Fusion 49, 055014 (2009)], which is fitted with a stainless steel-clad copper liner, conformal to the last closed flux surface. The liner can be heated to 350 °C. Several gas fueling systems, including supersonic gas injection and molecular cluster injection, have been studied and produce fueling efficiencies up to 35%. Discharges are strongly affected by wall conditioning. Discharges without lithium wall coatings are limited to plasma currents of order 10 kA, and discharge durations of order 5 ms. With solid lithium coatings discharge currents exceed 70 kA, and discharge durations exceed 30 ms. Heating the lithium wall coating, however, results in a prompt degradation of the discharge, at the melting point of lithium. These results suggest that the simplest approach to implementing liquid lithium walls in a tokamak—thin, evaporated, liquefied coatings of lithium—does not produce an adequately clean surface.

The lithium tokamak experiment (LTX) (R 0 = 0.4 m, a = 0.26 m) is designed to explore the low-recycling, lithium wall operating regime for magnetically confined plasmas. A set of shell quadrants internal to the vacuum vessel and conformal to the plasma last closed flux surface is designed to be coated with lithium to produce the lithium plasma-facing surface. The shell quadrants are highly conductive in order to maintain an even thermal distribution, but this conductivity also permits eddy currents to flow that can be larger in magnitude than the plasma current. Due to this effect, plasma start-up is greatly complicated as is the development of an applicable equilibrium solver code. A suitable code, LTX LRDFIT, has been developed and benchmarked and has now been used to compare plasma flux surface reconstructions for discharges before and after initial lithium wall conditioning. Dramatic improvements in plasma performance and shaping have been noted with the introduction of lithium.

Neutral-beam heating of plasmas in the Tokamak Fusion Test Reactor at low preinjection densities (n/sub e/(0)approx. =10/sup 19/ m/sup -3/) were characterized by T/sub e/(0) = 6.5 keV, T/sub i/(0) = 20 keV, n/sub e/(0) = 7 x 10/sup 19/ m/sup -3/, tau/sub E/ = 170 msec, ..beta../sub t//sub h//sub e//sub t//sub a/ = 2, and a d(d,n)/sup 3/He neutron emission rate of 10/sup 16/ sec/sup -1/. The ion temperature and the deuterium-fusion neutron yields were significantly higher than for previous tokamak experiments. The low initial densities were achieved by operation of the Tokamak Fusion Test Reactor with low plasma currents (less than or equal to1 MA) and by extensive limiter conditioning.

During 2 yr of experimental operations, the Poloidal Divertor Experiment (PDX) inner wall neutral beam graphite armor provided protection for perpendicular heating injections into normal and disruptive plasmas as well as injection in the absence of plasma for special experiments, calibrations, and tests involving the optimization and development of the PDX neutral beam injection system. About 80 to 100 heating injections occurred per operating day, at a 360-s duty cycle, into plasmas of various densities, and typically approx. =5 to 50% of the injected neutral beam power was transmitted to the armor. More than 10/sup 3/ neutral beam pulses of 100- to 300-ms duration were injected in the absence of plasma at peak power densities of 1.5 to 3 kW/cm/sup 2/, yielding peak surface temperatures of 950 to 1550/sup 0/C. There was no significant impurity production attributable to beam heating of the armor, and no observed beam-induced, macroscopic surface damage. Many of the design constraints and performance issues encountered in this work are relevant to the design of larger fusion devices.

The PBX-M Neutral Probe Beam Project was initiated to design, fabricate, acceptance-test, install, and commission an 80 kV, 2.7 A neutral probe beam for use in PBX-M q-profile measurements. The 100 ms, H° output beam is modulated at 1 kHz and has a pulse repetition rate of 3 min. The beam has a divergence of 0.5° and a focal length of 400 cm. The injection angle can be varied about 25°, from near perpendicular to outboard tangential. The beamline is shielded from magnetic fields and interfaces to the existing mechanical, electrical, and control systems of PBX-M.

Although the region of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel wall which is susceptible to damage by neutral beam strike is armored with a mosaic of TiC‐clad POCO graphite tiles, at power deposition levels above 2.5 kW/cm2 the armor surface temperature exceeds 1200 °C within 250 ms, and itself becomes susceptible to damage. In order to protect the wall armor, a neutral beam interlock system based on infrared pyrometry measurement of the armor surface temperature was installed on TFTR. For each beamline, a three‐fiber‐optic telescope views three areas of ∼30 cm diameter centered on the armor hot spots for the three ion sources. Each signal is fiber‐optic coupled to a remote 900‐nm pyrometer which feeds analog signals to the neutral beam interrupt circuits. The pyrometer interlock system is designed to interrupt each of the 12 ion sources independently within 10 ms of the temperature exceeding a threshold which can be set in the range of 500–2300 °C. A description of the pyrometer interlock sy...

In 1981, the joint ORNL/PPPL PDX neutral beam heating project succeeded in reliably injecting 7.2 MW of D/sup 0/ into the PDX plasma, at nearly perpendicular angles, and achieved ion temperatures up to 6.5 keV. The expeditious achievement of this result was due to the thorough conditioning and qualification of the PDX neutral beam ion sources at ORNL prior to delivery coupled with several field design changes and improvements in the injection system made at PPPL as a result of neutral beam operating experience with the PLT tokamak. It has been found that the operation of high power neutral beam injection systems in a tokamak-neutral beam environment requires procedures and performance different from those required for development operation on test stands. In this paper, we review the installatin of the PDX neutral beam injection system, and its operation and performance during the initial high power plasma heating experiments with the PDX tokamak.