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LTX-β, the upgraded Lithium Tokamak Experiment, recently began its first campaign with a goal to study the transport properties of gradient-free electron temperature profile equilibria with increased toroidal field and neutral beam injection. The temperature-gradient-free equilibria in LTX were enabled by the lithium plasma-facing surface. To make the lithium surface conditions more consistent for LTX-β, a new, faster lithium evaporator system has been developed. The toroidal magnetic field has been nearly doubled to ≥0.3 T. Auxiliary heating and fueling with a high flux (35 A, 20 kV) neutral beam has begun. The beam provides core fueling without cold edge neutrals, to test whether flat temperature profiles can be sustained, and makes possible core ion temperature measurements via charge exchange recombination spectroscopy. Neutral beam fueling has been demonstrated with increases in line-integrated density. The coupling of the beam matches well with NUBEAM predictions for shine through. Improved performance has been observed following lithium evaporative coatings and an increase in field strength.

By the presented paper I tried to express my deep appreciation to Vitaly D. Shafranov who was my supervisor, friend and mentor. The great value of his equilibrium theory is widely recognized as one of the pillar of tokamaks. At the same time, it may look like the importance of contribution of founders of tokamak physics is diminishing in favor on new research results and models. The paper shows that this is not the case: after six decades of tokamak fusion, the program is incapable to demonstrate fusion power exceeding the applied plasma heating. In contrast to this situation, the return to fundamentals of tokamak physics suggests much more efficient approach with order of magnitude better confinement, consistent with burning plasma.

Iron core transformer tokamaks have a distinct element, namely the iron core which can affect the plasma stability of the plasma. As a consequence of the high non-linear dependence of the magneto-hydrodynamic solutions on the iron permeability μ Fe in JET tokamak for example, Vertical Displacement Events (VDEs), equilibrium and stability calculations are more complicated and more time consuming than in air core transformer tokamaks. By considering the ferromagnetic components as a linear, isotropic and homogeneous media on subdomains, it is known that these media can be replaced by a homogeneous one (vacuum) and a surface-current density distribution i Fe(l) on the separation surfaces between subdomains, with l the curve taken along the curve separating two different magnetic media. For the case of geometry with rotational symmetry, this surface-current density distribution along a curve, in a meridian plane, is given by a Fredholm integral equation of second kind. Practically, the advantage of this method is more obvious for the inverse formulation of the VDEs and stability problems by moving the non-linear term (due to the iron permeability μ Fe ) from the differential operator to the r.h.s. of the equations. In this paper, we are presenting how the influence of ferromagnetic components in the equations of the surface currents developed in the vessel structures during Wall Touching Kink Modes (WTKMs) can be taken into account and are reviewing the equations to be solved in order to simulate the influence of the ferromagnetic components in VDEs and equilibrium stability calculations. The numerical results of these simulations for a real JET tokamak structure and plasma parameters will be reported in a future paper.

Lithium coating technology and liquid lithium limiter have been applied on HT-7 tokamak and many significant results were obtained. The study of exploring lithium as potential plasma facing material is being carried out on EAST tokamak. A flowing liquid lithium (FLiLi) limiter has been successfully tested in the EAST in 2014, and the in-vessel electromagnetic (EM) pump was validated to make liquid lithium circulate from the bottom collector to the distributor of the limiter, which realized the FLiLi operated under steady-state and also drastically reduced the total amount of lithium used in the experiment. To adapt to the plasma shape during discharge, the limiter can be moved along the guide rail into vacuum vessel to meet various plasma scenarios by the driving system. During the operation, the lithium can be easily controlled by adjusting the current of EM pump and the

International audience; The paper presents an analysis of disruptions occurring during JET-ILW plasma operations covering the period from the start of ILW (ITER-like wall) operation up to completion of JET operation in 2016. The total number of disruptions was 1951 including 466 with deliberately induced disruptions. The average rate of unintended disruptions was 16.1 %, which is significantly above the ITER target at 15 MA. The pre-disruptive plasma parameters are: plasma current Ip = (0.82-3.38) MA, toroidal field BT = (0.98-3.4) T, safety factor q95 = (1.52-9.05), plasma internal inductance li = (0.58-1.86), Greenwald density limit fraction FGWL = (0.04-1.61), with 720 X-point plasma pulses from a subset of 1420 unintended disruption shots. Massive gas injection (MGI) has been routinely used in protection mode both to terminate pulses when the plasma is at risk of disruption and to mitigate against disruption effects. The MGI was mainly triggered by the n = 1 locked mode (LM) amplitude exceeding a threshold or by the disruption itself, namely, either dIp/dt (specifically, a fast drop in Ip) or the toroidal loop voltage exceeding threshold values. For mitigation purposes, only the LM was used as a physics precursor and threshold on the LM signal was used to trigger the MGI prior to disruption. Long lasting LM (? 100 ms) do exist prior to disruption in 75% of cases. However, 10% of non-disruptive pulses have a LM which eventually vanished without disruption. The plasma current quench (CQ) may result in 3D configurations, termed as asymmetrical disruptions, which are accompanied by sideways forces. Unmitigated vertical displacement events (VDEs) generally have significant plasma current toroidal asymmetries. Unmitigated non-VDE disruptions also have large plasma current asymmetries presumably because there is no plasma vertical position control during the CQ and so they too are subject to large vertical displacements. MGI is a reliable tool to mitigate 3D effects and correspondingly sideways forces during the CQ. The vessel structure loads depend on the force impulse and force time behaviour, including their rotation. The toroidal rotation of 3D configuration may cause resonance with the natural frequencies of the vessel components in large tokamaks such as ITER. The JET-ILW amplitude-frequency interdependence of toroidal rotation of 3D configurations is presented.

Two different designs of flowing liquid lithium limiter were first tested for power exhaust and particle removal in HT-7 in 2012 autumn campaign. During the experiments, the reliability and compatibility of the limiters within Tokamak were experimentally demonstrated, and some positive results were achieved. It was found that the flowing liquid lithium limiter was effective for suppressing H concentration and led to a low ratio of H/(H + D). O impurity was slightly decreased by using limiters as well as when using a Li coating. A significant increase of the wall retention ratio was also observed which resulted from the outstanding D particles pumping ability of flowing liquid lithium limiters. The strong interaction between plasma and lithium surface could cause lithium ejection into plasma and lead to disruptions. The stable plasmas produced by uniform Li flow were in favor of lithium control. While the limiters were applied with a uniform Li flow, the normal plasma was easy to be obtained, and the energy confinement time increased from ∼0.025 s to 0.04 s. Furthermore, it was encouraging to note that the application of flowing liquid lithium limiters could further improve the confinement of plasma by ∼10% on the basis of Li coating. These remarkable results will help for the following design of flowing liquid lithium limiter in EAST to improve the plasma operation.

In 2012, lithium coating with an upgraded system on EAST, the first application of lithium granules injection for ELMs pacing on EAST, and the first flowing lithium limiter experiments on HT-7 have successfully been carried out and several new results were obtained. On EAST, it was found that both the Mo first walls and the C divertors were well coated by lithium and the lithium film coverage was increased up to 85%, which greatly contributed to the new achievements of EAST, especially stationary H-mode plasma over 30 s and long pulse plasma over 400 s. And at the same time, ELMs suppression by active lithium conditioning and ELMs pacing using lithium granules injection were demonstrated and reported for the first time on EAST. On HT-7, flowing liquid lithium limiters using the TEMHD concept and using a thin flowing film concept were also initially tested and some references were obtained for the future development. Those experiments show that lithium should be an important material for fusion devices. It could be used for wall conditioning, ELMs mitigation and also provide a self-recovery plasma facing components in future fusion devices.

Experiments with liquid lithium limiters (LLLs) have been successfully performed in HT-7 since 2009 and the effects of different limiter surface structures on the ejection of Li droplets have been studied and compared. The experiments have demonstrated that strong interaction between the plasma and the liquid surface can cause intense Li efflux in the form of ejected Li droplets – which can, in turn, lead to plasma disruptions. The details of the LLL plasma-facing surface were observed to be extremely important in determining performance. Five different LLLs were evaluated in this work: two types of static free-surface limiters and three types of flowing liquid Li (FLLL) structures. It has been demonstrated that a FLLL with a slowly flowing thin liquid Li film on vertical flow plate which was pre-treated with evaporated Li was much less susceptible to Li droplet ejection than any of the other structures tested in this work. It was further observed that the plasmas run against this type of limiter were reproducibly well-behaved. These results provide technical references for the design of FLLLs in future tokamaks so as to avoid strong Li ejection and to decrease disruptive plasmas.

The paper describes two major non-linear properties of the vertical instability of a tokamak plasma, which has a vertical elongation: (a) inductive excitation of surface (edge) currents stabilizing the instability and converting it into fast equilibrium evolution, and (b) creation of a wetting zone without the normal component of the magnetic field when the plasma has contact with material surfaces. Two major disruption effects for both mitigated and non-mitigated disruptions, important for JET and ITER, were considered: (a) excitation of vertical disruption during the current quench (i.e., abnormal plasma current ramp down) and (b) related to the wetting zone, potential leading edge effect in plasma energy deposition to the in-vessel tiles during disruptions. Our considerations together with a 2-D version of the VDE (Vertical Disruption Event) code are based on a new mathematical model, called Tokamak MHD (TMHD), as a replacement for the conventional model, a model that cannot solve numerical problems related to extreme plasma anisotropy and negligible mass. The code includes a 3-D model of surface currents on a thin conductive wall and has a well-specified algorithm for extension to vertical disruptions that excite asymmetric kink modes.

Lithium is a very attractive element due to its very low radiation power, strong H retention as well as strong O getter activity. Flowing liquid lithium (FLiLi) device, to be used as a plasma-facing limiters, has been designed and will be tested in HT-7 tokamak. It is mainly composed of distributor, guide plate, collector, and heater as well as cooling loop. The heater uses heater strip and cooling loop design, to control the temperature of lithium on the guide plate ranging from 200 °C to 400 °C. The distributor attached to feeding pipe, distributes liquid lithium (LiLi) flowing on the guide plate. The collector was designed to reclaim the superfluous LiLi and transport it out of device. The paper focuses on the design of flowing liquid lithium device. In addition to the process of design, thermal analysis has been carried out using finite element method (FEM) for optimizing the structure of heater and cooling loop and results of analysis are presented.

Deuterium fueling experiments were conducted with the NSTX Liquid Lithium Divertor (LLD). Lithium evaporation recoated the LLD surface to approximate flowing liquid Li to sustain D retention. In the first experiment with the diverted outer strike point on the LLD, the difference between the applied D gas input and the plasma D content reached very high values without disrupting the plasma, as would normally occur in the absence of Li pumping, and there was also little change in plasma D content. In the second experiment, constant fueling was applied, as the LLD temperature was varied to change the surface from solid to liquid. The D retention was relatively constant, and about the same as that for solid Li coatings on graphite, or twice that achieved without Li PFC coatings. Contamination of the LLD surface was also possible due to compound formation and erosion and redeposition from carbon PFCs.

The Lithium Tokamak eXperiment (LTX) is a spherical tokamak with the unique capability of studying the low-recycling regime by coating nearly 90% of the first wall with lithium in either solid or liquid form. Several grams of lithium are evaporated onto the plasma-facing side of the first wall. Without lithium coatings, the plasma discharge is limited to less than 5 ms and only 10 kA of plasma current, and the first wall acts as a particle source. With cold lithium coatings, plasma discharges last up to 20 ms with plasma currents up to 70 kA. The lithium coating provides a low-recycling first wall condition for the plasma and higher fueling rates are required to realize plasma densities similar to that of pre-lithium walls. Traditional puff fueling, supersonic gas injection, and molecular cluster injection (MCI) are used. Liquid lithium experiments will begin in 2012.

Movable liquid lithium limiter (LLL) experiments with both free-surface and capillary-pore system (CPS) configurations were successively utilized on HT-7 in 2009. In the campaign of 2011, experiments with a new lithium (Li) limiter, which used a CPS configuration with a pore size of about 100 μm and active liquid Li injection from outside of HT-7, were performed. It was found that liquid Li could flow freely driven by only gravity. Confinement of the liquid Li was improved by using the CPS configuration. It was also found that plasma performance was improved due to low recycling and significantly reduced impurity radiation. However, when the CPS LLL is employed as the primary limiter the plasma disruptivity rate increases from ∼15% to ∼90% possibly due to Li emission.

Recently, routine coatings of plasma facing materials with lithium were carried out on EAST using both upgraded evaporative ovens and real-time injection of lithium powder. Employing daily lithium coatings of 10–30 g, the H/(H + D) ratio has been decreased below 10% and both impurity levels and MHD activity have been suppressed. Using these coating technologies, plasma performance has been improved significantly. For example, a 10 s H-mode plasma was achieved at the beginning of the 2012 EAST campaign. Techniques for removing Li coatings from the vacuum vessel have been developed in EAST and rapid recovery of plasma performance following air vents has been documented.

During plasma disruptions in tokamaks, electric currents are excited in the three-dimensional vessel structures by a plasma Wall Touching Kink Mode (WTKM). These modes are frequently excited during a Vertical Displacement Event (VDE) and cause big sideways forces on the vacuum vessel which are difficult to confront. To understand the plasma disruptions in tokamaks and to predict their effects, realistic simulations of these electric currents are required. In the present paper a flat triangle Finite Element (FE) representation of these surface currents excited in a thin conducting wall of arbitrary three-dimensional geometry is described. Our wall model covers both eddy currents, excited inductively, and source/sink currents due to current sharing between the plasma and the wall.

NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the LLD, i.e., about two times that of no-lithium conditions. The role of lithium impurities in this result is discussed. Following the 2010 experimental campaign, inspection of the LLD found mechanical damage to the plate supports, and other hardware resulting from forces following plasma current disruptions. The LLD was removed, upgraded, and reinstalled. A row of molybdenum tiles was installed inboard of the LLD for 2011 experiments with both inner and outer strike points on lithiated molybdenum to allow investigation of lithium plasma facing issues encountered in the first testing of the LLD.

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up to ∼160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium, 1300 g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1] , we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R&D required for use of lithium in future magnetic fusion facilities including ITER.

NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma-facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core

A model for simulating the diffusive evaporation of lithium into a helium filled NSTX vacuum vessel is described and validated against an initial set of deposition experiments. The DEGAS 2 based model consists of a three-dimensional representation of the vacuum vessel, the elastic scattering process, and a kinetic description of the evaporated atoms. Additional assumptions are required to account for deuterium out-gassing during the validation experiments. The model agrees with the data over a range of pressures to within the estimated uncertainties. Suggestions are made for more discriminating experiments that will lead to an improved model.

Lithium as first wall materials was successively performed on EAST and HT-7 superconducting tokamaks. In the last 2 years, lithium coating were carried out by means of ICRF, DC-GDC and HF-GDC on EAST and HT-7, and liquid lithium limiters with free lithium surface and CPS configuration were successively applied on HT-7. Both techniques of lithium coating and liquid lithium limiter were useful for the improvement of plasma performances. This paper will give the first results of lithium experiments on EAST and HT-7.

Major developments on the Mega Amp Spherical Tokamak (MAST) have enabled important advances in support of ITER and the physics basis of a spherical tokamak (ST) based component test facility (CTF), as well as providing new insight into underlying tokamak physics. For example, L–H transition studies benefit from high spatial and temporal resolution measurements of pedestal profile evolution (temperature, density and radial electric field) and in support of pedestal stability studies the edge current density profile has been inferred from motional Stark effect measurements. The influence of the q-profile and E × B flow shear on transport has been studied in MAST and equilibrium flow shear has been included in gyro-kinetic codes, improving comparisons with the experimental data. H-modes exhibit a weaker q and stronger collisionality dependence of heat diffusivity than implied by IPB98(y,2) scaling, which may have important implications for the design of an ST-based CTF. ELM mitigation, an important issue for ITER, has been demonstrated by applying resonant magnetic perturbations (RMPs) using both internal and external coils, but full stabilization of type-I ELMs has not been observed. Modelling shows the importance of including the plasma response to the RMP fields. MAST plasmas with q > 1 and weak central magnetic shear regularly exhibit a long-lived saturated ideal internal mode. Measured plasma braking in the presence of this mode compares well with neo-classical toroidal viscosity theory. In support of basic physics understanding, high resolution Thomson scattering measurements are providing new insight into sawtooth crash dynamics and neo-classical tearing mode critical island widths. Retarding field analyser measurements show elevated ion temperatures in the scrape-off layer of L-mode plasmas and, in the presence of type-I ELMs, ions with energy greater than 500 eV are detected 20 cm outside the separatrix. Disruption mitigation by massive gas injection has reduced divertor heat loads by up to 70%.

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.

NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.

First experiment of liquid lithium limiter was successfully carried out on HT-7 tokamak and a few positive results were obtained. The results showed that by using lithium limiter, specially liquid lithium limiter, Hα intensity reduced 20–30%, the emission of CIII and OV decreased about 10–20%, loop voltage had a slight decline, the core electron temperature slightly increased, the particle confinement time increased by a factor of 2, and the energy confinement time increased 20%. After lithium coating, the hydrogen recycling decreased, and core electron temperature increased significantly by a factor of 2. At the same time, after lithium coating, electron density of edge plasmas obviously decreased while electron temperature slightly increased. These encouraging results are very useful for further research of long tray lithium limiter on HT-7 and liquid divertor on EAST.

A strategy to develop self-consistent simulations of the behavior of lithium in the Liquid Lithium Divertor (LLD) module to be installed in NSTX is described. In this initial stage of the plan, the UEDGE edge plasma transport code is used to simulate an existing NSTX shot, with UEDGE's transport coefficients set using midplane and divertor diagnostic data. The LLD is incorporated into the simulations as a reduction in the recycling coefficient over the outer divertor. Heat transfer calculations performed using the resulting heat flux profiles provide preliminary estimates on operating limits for the LLD as well as input data for subsequent steps in the LLD modeling effort.

Two lithium evaporators were used to evaporate more than 100 g of lithium on to the NSTX lower divertor region. Prior to each discharge, the evaporators were withdrawn behind shutters, where they also remained during the subsequent HeGDC applied for periods up to 9.5 min. After the HeGDC, the shutters were opened and the LITERs were reinserted to deposit lithium on the lower divertor target for 10 min, at rates of 10–70 mg/min, prior to the next discharge. The major improvements in plasma performance from these lithium depositions include: (1) plasma density reduction as a result of lithium deposition; (2) suppression of ELMs; (3) improvement of energy confinement in a low-triangularity shape; (4) improvement in plasma performance for standard, high-triangularity discharges; (5) reduction of the required HeGDC time between discharges; (6) increased pedestal electron and ion temperature; (7) reduced SOL plasma density; and (8) reduced edge neutral density.

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n e / n GW ∼ 1), to enable n e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n e / n GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m 2 ) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Lithium evaporated onto plasma facing components in the NSTX lower divertor has made dramatic improvements in discharge performance. As lithium accumulated, plasmas previously exhibiting robust Type 1 ELMs gradually transformed into discharges with intermittent ELMs and finally into continuously evolving ELM-free discharges. During this sequence, other discharge parameters changed in a complicated manner. As the ELMs disappeared, energy confinement improved and remarkable changes in edge and scrape-off layer plasma properties were observed. These results demonstrate that active modification of plasma surface interactions can preempt large ELMs.

The liquid lithium divertor (LLD) to be installed in NSTX has four toroidal panels, each a conical section inclined at 22° like the previous graphite divertor tiles. Each LLD panel is a copper plate clad with ∼0.25 mm of stainless steel (SS) and a surface layer of flame sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. LITER (evaporators) already used in NSTX will be upgraded for the LLD. Each has twelve 500 W cartridge heaters with thermocouples, 16 other thermocouples, and a channel for helium cooling. During LLD experiments, the LLD will be heated so that the lithium is just above its melting temperature. The length of each shot will be preset to prevent excessive evaporation of lithium from the LLD. This duration depends on the heat load and is likely to be in the range of less than a second to several seconds. Careful thermal control of the LLD is important to maximize the shot times and to guide operation of the LLD. This paper describes the layout of the LLD, its expected thermal performance, the control system, and supporting experiments and analysis. A companion paper in this conference, “Physics design requirements for the national spherical torus experiment liquid lithium divertor,” provides other information.

A method is developed for calculating uncertainties in reconstructing the equilibrium profiles of the safety factor q and plasma pressure p in the ITER device from external magnetic measurements and from motional Stark effect line polarization (MSE-LP) and motional Stark effect line shift (MSE-LS) signals from excited NBI atoms inside the plasma core. It is shown that, with MSE-LP signals, as well as with MSE-LS signals (the use of which was recently proposed by Nova Photonics, Inc.), it is possible to substantially improve the reconstruction of the profiles that determine the plasma magnetic configuration.

The mission of the Current Drive eXperiment-Upgrade (CDX-U) spherical tokamak is to investigate lithium as a plasma-facing component (PFC). The latest CDX-U experiments used a combination of a toroidal liquid lithium limiter and lithium wall coatings applied between plasma shots. Recycling coefficients for these plasmas were deduced to be 30% or below, and are the lowest ever observed in magnetically-confined plasmas. The corresponding energy confinement times showed nearly a factor of six improvement over discharges without lithium PFC’s. An electron beam (e-beam) for evaporating lithium from the toroidal limiter was one of the techniques used to create lithium wall coatings in CDX-U. The evaporation was not localized to the e-beam spot, but occurred only after the entire volume of lithium in toroidal limiter was liquefied. This demonstration of the ability of lithium to handle high heat loads can have significant consequences for PFC’s in future burning plasma devices. � 2007 Elsevier B.V. All rights reserved. PACS: 28.52.Fa; 52.25.Vy; 52.40.Hf; 52.55.Fa

The low recycling regime, although never considered as an option for ITER, may suggest a solution to its important issues, such as edge localized modes, plasma and particle control, tritium inventory, damage of plasma facing components and dust accumulation, in a way consistent with both the ITER mission (including the ignition) and its baseline design and safety. Such a regime can be approached using liquid lithium surfaces efficiently pumping hydrogen isotopes. An active area of about 40 m 2 , covered by ’0.1 mm thick lithium, which is replenished with the rate of 10 kg/h would be capable of absorption of plasma D and T particles and at the same time consistent with the ITER limitations regarding lithium. For low recycling conditions, a new consideration is outlined for the helium ash pumping problem. � 2007 Elsevier B.V. All rights reserved. PACS: 52.35.Py; 52.40.Hf; 52.55.Fa; 28.52.Fa

Lithium coatings on the graphite plasma facing components (PFCs) in NSTX are being investigated as a tool for density profile control and reducing the recycling of hydrogen isotopes. Repeated lithium pellet injection into Center Stack Limited and Lower Single Null ohmic helium discharges were used to coat graphite surfaces that had been pre-conditioned with ohmic helium discharges of the same shape to reduce their contribution to hydrogen isotope recycling. The following deuterium NBI reference discharges exhibited a reduction in density by a factor of about 3 for limited and 2 for diverted plasmas, respectively, and peaked density profiles. Recently, a lithium evaporator has been used to apply thin coatings on conditioned and unconditioned PFCs. Effects on the plasma density and the impurities were obtained by pre-conditioning the PFCs with ohmic helium discharges, and performing the first deuterium NBI discharge as soon as possible after applying the lithium coating.

A program involving the extensive and systematic use of lithium (Li) as a “first,” or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak—both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.

Recent experiments in the Current Drive Experiment-Upgrade (CDX-U) provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall to gain engineering experience with a liquid metal first wall and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, Btoroidal = 2 kG, IP = 100 kA, Te(0) ∼ 100 eV, ne(0) ∼ 5 × 10 19 m −3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium pool limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium pool limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5–8 increase in gas fuelling to achieve a comparable density, indicating that recycling is strongly reduced. Modelling of the discharges demonstrated that the lithium limited discharges are consistent with Zeffective < 1.2 (compared with 2.4 for the pre-lithium discharges), a broadened current channel and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced.

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.

A generalized global particle-in-cell (PIC) code has been developed to systematically study neoclassical physics and equilibrium electric field dynamics in general toroidal geometry. This capability enables realistic assessment of the irreducible minimum transport level and the bootstrap current in toroidal systems. The associated analysis takes into account the comprehensive influences of large orbits, toroidal geometry, and self-consistent electric field, for more meaningful experimental comparisons. The simulation model and δf algorithm are described, and an interesting new result of non-local ion thermal transport is presented.

Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma–liquid metal interactions in a functioning tokamak. Most of the interest in liquid metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm 2 , subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now been performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments, the liquid lithium plasma-facing area was increased to 2000 cm 2 . Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described.

The use of flowing liquid lithium as a first wall for a reactor has potentially attractive physics and engineering features. The current drive experiment-upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has begun experiments with a fully toroidal liquid lithium limiter. CDX-U is a compact (R = 34 cm, a = 22 cm, B toroidal = 2 kG, J P = 100 kA, T e (0) ∼ 100 eV, n e (0) ∼ 5 × 10 19 m -3 ) short-pulse ( < 25 ms) spherical tokamak with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, can be filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium prior to the experiment. The total area of the tray is approximately 2000 cm 2 . The tokamak edge plasma, when operated in contact with the lithium-filled tray, shows evidence of reduced impurities and recycling. The reduction in recycling and impurities is largest when the lithium is liquefied by heating to 250 °C. Discharges which are limited by the liquid lithium tray show evidence of performance enhancement. Radiated power is reduced and there is spectroscopic evidence for increases in the core electron temperature. Furthermore, the use of a liquid lithium limiter reduces the need for conditioning discharges prior to high current operation. The future development path for liquid lithium limiter systems in CDX-U is also discussed.

Guiding center equations for particle motion in a toroidal magnetic configuration are derived using general magnetic coordinates. For the case of axisymmetry, the explicit transformation to exact Hamiltonian canonical variables is presented for the first time. Approximate canonical coordinates are introduced also for three-dimensional configurations with strong toroidal magnetic field. Previous derivations made use of so-called Boozer equilibrium coordinates, which are highly nonuniform and are canonical only in the exceptional case of low beta, up–down symmetric configurations. The present formalism is valid for arbitrary, spatially well distributed magnetic coordinates, greatly increasing the accuracy of calculations. Magnetostatic equilibrium is not assumed in the present formalism, the analysis holds for any configuration with nested flux surfaces.

The National Spherical Torus Experiment (NSTX) is currently being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive operation in high-performance plasmas, assessing plasma–material interface (PMI) solutions to handle the high heat loads expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection (CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet–Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Toroidal Alfven eigenmode avalanches and higher frequency Alfven eigenmodes can affect NB-driven current through energy loss and redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium–oxygen affinity, which increases deuterium pumping and reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding. Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing unstable regime. Two Alfven eigenmode-driven fast ion transport models have been developed and successfully benchmarked against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned for mid-2015.

The current drive experiment-upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego (UCSD). Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance.

The theory of stabilizing free-boundary magnetohydrodynamic instabilities in tokamaks by liquid lithium streams driven by magnetic propulsion is formulated. It is shown that even for uniform current distribution lithium streams open stability windows. Also, it was found that while the conventional, wall-locked, resistive wall mode can be well suppressed by the flow, a new, stream-locked mode determines the limits of the flow stabilization.