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Net erosion/deposition rates of carbon and tungsten were measured at the outer strike point of the divertor plasma on the floor of the DIII-D tokamak during deuterium H-mode operation at an average power deposition of about 45 W/cm 2 . For carbon, net erosion rates of up to 4 nm/s were found. For a tungsten film, no appreciable erosion was detected. However, measurements of deposited tungsten on adjacent carbon surfaces indicated a net W erosion rate of 0.06 nm/s.

In this paper, we examine a denser and broader scrape-off layer (SOL) plasma, one class of which was first seen in VH mode, in the DIII-D tokamak. An enhanced SOL appears in many types of double-null (DN) discharges and is not a property of VH mode only. The L and H mode DN enhanced SOL density and temperature profiles exhibit a 5–6 cm broad profile outside the separatrix. The VH mode enhanced SOL is associated with a higher separatrix density and temperature and is only 1.5–2 cm wide. Both types of SOL enhancement are only observed in high triangularity discharges. The origin of the enhanced SOL is, however, not yet understood.

Exposure of tungsten to low energy (

The open divertor in DIII-D has been modified by the installation of a baffled plenum with a narrow entrance aperture. Neutral particle pressures in the range of 2-20 mtorr have been measured in this new plenum during experiments in L and H mode plasmas in which the neutral beam power was varied and the divertor target location was scanned with respect to the entrance aperture. These pressures exceed the predicted minimum required to provide adequate particle exhaust for controlling the density in H mode discharges when pumping experiments are conducted in the future. The pressure measurements presented here, which were carried out in the absence of a pump, were modelled with the aid of two transport codes: an edge plasma code and a neutrals code. It was found in this modelling process that the divertor flux amplification factor required to explain the measured pressures is of the order of 30 to 40, much higher than the factors used in the modelling of data in the absence of the baffle. This higher flux amplification factor results from increased recycling at the divertor owing to the presence of the baffle

D{sub 2} gas injected into ELMing H-mode discharges in DIII-D reduced total integrated heat flux to the divertor by {approximately}2{times} and peak heat flux by {approximately}5{times}, with only modest degradation to plasma stored energy. Steady gas injection without particle pumping results in eventual degradation in stored energy. The initial reduction in peak heat flux at the divertor tiles may be primarily due to the increase in radiated power from the X-point/divertor region. The eventual formation of a high density region near the X-point appears to play a role in momentum (and energy) transfer from the flux surfaces near the outboard strike point to flux surfaces farther out into the scrapeoff. This may also contribute to further reduction in peak heat flux.

In this paper, scrape-off layer measurements in DIII-D are presented as a function of the main discharge plasma parameters. A systematic study is under way to understand and predict the behavior of the edge and divertor plasma in DIII-D and this scaling behavior will be crucial for the design of ITER. To facilitate the studies, a fast reciprocating Langmuir probe incorporating five graphite tips was installed at the midplane of DIII-D which has the capability of performing multiple plunges 1 cm inside the separatrix during 5 MW of NBI. Density and temperature profiles in the midplane (reciprocating probe), near the top (Thomson scattering) and at the lower divertor plate (fixed Langmuir probe array) are compared by mapping the measurements into magnetic flux coordinates. Local pressure measurements are compared on different parts of a flux surface. The three different local measurements also indicate the spatial evolution of plasma conditions as plasma approaches the divertor plate. Ohmic and L-mode discharges exhibit similar (exponential) density and temperature decay in the scrape-off layer. H-mode discharges, however, display a faster spatial decay reflecting at least a factor of 3 decrease in the perpendicular diffusion coefficient. Consistency of the magnitude and scaling behavior of the edge profile parameters with models of the scrape-off layer is examined.

The divertor material evaluation system (DIMES) at DIII-D is a collaborative program between General Atomics, Sandia National Laboratories (SNL), and Argonne National Laboratory (ANL). This program was initiated in response to the need for understanding the interaction between the plasma and divertor surface materials in tokamaks. Material erosion, tritium retention, disruption effects and material transport are very important topics for the design of ITER. The first phase of the DIMES study is integral material exposure measurements. The second phase of the study is the installation of the DIMES sample changer mechanism. The mechanical design goal for the second phase is to allow the insertion of test materials and instrumented samples into the bottom divertor plate region of DIII-D without venting the tokamak. Various samples can then be exchanged overnight after exposures as short as one plasma shot. This paper presents the results of the integral experiments, the design of the DIMES sample changer mechanism and the planning of material evaluation experiments at DIII-D using the DIMES mechanism.

An array of fixed Langmuir probes spanning the inner and outer divertor strike points has been used to measure the electron temperature, plasma density, and particle/heat flux profiles of the divertor plasma in DIII-D. We report here on observations made during the steady state phase of high-power neutral-beam heated H-mode plasmas. These data are important for predicting the expected impurity sources and target plate material lifetimes in future high power tokamaks such as ITER. H-mode plasmas formed with a high divertor neutral pressure in the absence of external gas feed in DIII-D were found to have beneficial consequences in reducing the peak values of divertor electron temperature without loss of core energy confinement. We have observed a steady state, high density ( n pl,d ≈10 14 cm −3 ), moderate temperature ( 2 . A better understanding of the importance of collisional effects may be needed to interpret these results.

We have developed an edge physics data base containing data for the plasma in the divertor region and the scrape-off layer (SOL) for the DIII-D tokamak. The data base provides many of the parameters necessary to model the power flow to the divertor and other plasma processes in the plasma edge. It will also facilitate the analysis of DIII-D data for comparison with other divertor tokamaks. In addition to the core plasma parameters, edge-specific data are included in this data base. Initial results using the data base show good agreement between the pressure profiles measured by the Langmuir probes and those determined from the Thomson data for the inner strike point, but not for the outer strike point region. We also find that the ratio of separatrix density to average core density, as well as the in/out asymmetry in the SOL power at the divertor in DIII-D do not agree with values currently assumed in modeling the international thermonuclear experimental reactor (ITER).

In this paper we report results of scaling studies aimed at determining how the divertor conditions vary with plasma current, toroidal field, and neutral beam heating power in H-mode discharges with ELMs in the DIII-D tokamak. We find that ELMs produce relatively more direct particle losses (50% or more of the total) than energy losses (≤20%). The time-average peak divertor heat flux in these plasmas is found to scale as d α ( P NBI I p )( B p,mp / B p,div ). The linear power dependence suggests that the plasma sheath at the targets is primarily responsible for limiting the parallel energy flow, while the I p variation may mean that the radial energy transport in the SOL decreases with increasing plasma current, just as it does in the core.

We report here measurements of the divertor electron density and temperature profiles for ohmic, NBI heated L-mode, and H-mode discharges in DIII-D. The scaling of peak divertor plasma parameters with discharge current, line average density, and input power has also been obtained. The measurements were made using two fixed, single-tipped Langmuir probes mounted at the outboard (larger major radius) divertor strike point. Complete profile information is obtained by scanning the separatrix intercept across the probes on either a shot-by-shot basis, or during a single discharge. Electron temperatures as high as 50 eV have been observed during the L-mode; densities are in the range of a few times 10 13 cm 3 . Large negative floating potentials, on the order of kT e , have been observed at the strike point. The temperature and density profiles peak inside the magnetically determined strike point location for L- and H-mode and are more peaked during H-mode, consistent with the improvement in radial transport.

We report here on a study of the DIII-D divertor plasma which combines experimental measurements and 2D plasma modeling using the Braams 82 code. New data from a set of Langmuir probes recently installed in the divertor are presented. These data complement divertor heat flux measurements and help to better characterize divertor operation in neutral-beam heated discharges. In L-mode plasmas we find that the electron temperature rises with beam power (up to 45 eV with Pbeam = 5 MW) and is a strong function of n¯e. The peak divertor heat flux rises linearly with power and develops a strong inboard/outboard asymmetry which depends upon the toroidal field direction. Such behavior is not seen in the B2 simulation results. Following the H-mode transition the divertor heat flux, plasma density and temperature remain low even as the main plasma density rises.

A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, Ip ≲ 3 MA, βT ≲ 10.7%, P(auxiliary)≲ 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions.

The DIII‐D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII‐D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and Hα measurements with TV cameras and fiber optics coupled to a high‐resolution spectrometer.

A quiescent plasma is sustained for 80 energy confinement times by only gas fuelling and neutral beam heating in an axisymmetric region of the Tandem Mirror Experiment Upgrade (TMX-U). This plasma should be unstable because of the bad magnetic curvature and the absence of ion cyclotron heating which previously provided ponderomotive stabilization to sustain plasmas in bad-curvature regions of other axisymmetric mirror experiments. The TMX-U data are consistent with stabilization by a symbiosis between two mechanisms – line tying, which reduces the growth rate, and finite Larmor radius edge stabilization, which can result in quiescent operation.

In this work, we examine how deuterium becomes trapped in plasma-exposed tungsten and forms near-surface platelet-shaped precipitates. How these bubbles nucleate and grow, as well as the amount of deuterium trapped within, is crucial for interpreting the experimental database. Here, we use a combined experimental/theoretical approach to provide further insight into the underlying physics. With the Tritium Plasma Experiment, we exposed a series of ITER-grade tungsten samples to high flux D plasmas (up to 1.5 × 1022 m−2 s−1) at temperatures ranging between 103 and 554 °C. Retention of deuterium trapped in the bulk, assessed through thermal desorption spectrometry, reached a maximum at 230 °C and diminished rapidly thereafter for T > 300 °C. Post-mortem examination of the surfaces revealed non-uniform growth of bubbles ranging in diameter between 1 and 10 μm over the surface with a clear correlation with grain boundaries. Electron back-scattering diffraction maps over a large area of the surface confirmed this dependence; grains containing bubbles were aligned with a preferred slip vector along the ⟨111⟩ directions. Focused ion beam profiles suggest that these bubbles nucleated as platelets at depths of 200 nm–1 μm beneath the surface and grew as a result of expansion of sub-surface cracks. To estimate the amount of deuterium trapped in these defects relative to other sites within the material, we applied a continuum-scale treatment of hydrogen isotope precipitation. In addition, we propose a straightforward model of near-surface platelet expansion that reproduces bubble sizes consistent with our measurements. For the tungsten microstructure considered here, we find that bubbles would only weakly affect migration of D into the material, perhaps explaining why deep trapping was observed in prior studies with plasma-exposed neutron-irradiated specimens. We foresee no insurmountable issues that would prevent the theoretical framework developed here from being extended to a broader range of systems where precipitation of insoluble gases in ion beam or plasma-exposed metals is of interest.

Experiments are presented that investigate the mechanistic cause of multilayer erosion observed from condenser optics exposed to EUV laser-produced plasma (LPP) sources. Using a Xe filament jet source excited with Nd-YAG laser radiation (300 mJ/pulse), measurements were made of material erosion from Au, Mo, Si and C using coated quartz microbalances located 127 mm from the plasma. The observed erosion rates were as follows: Au=99nm/106 shots, Mo= 26nm/106 shots, Si=19nm/106 shots, and C=6nm/106 shots. The relative ratio Au:Mo:Si:C of erosion rates observed experimentally, 16:4:3:1 compares favorably with that predicted from an atomic sputtering model assuming 20 kV Xe ions, 16:6:4:1. The relative agreement indicates that Xe-substrate sputtering is largely responsible for the erosion of Mo/Si multilayers on condenser optics that directly face the plasma. Time-of-flight Faraday cup measurements reveal the emission of high energy Xe ions from the Xe-filament jet plasma. The erosion rate does not depend on the repetition rate of the laser, suggesting a thermal mechanism is not operative. The Xe-filament jet erosion is ~20x that observed from a Xe spray jet. Since the long-lived (millisecond time scale) plasma emanating from these two sources are the same to within ~30%, sputtering from this long-lived plasma can be ruled out as an erosion agent.

A critical issue in the realization of EUV lithography (EUVL) as a production technology is the lifetime of the condenser, the optic in closest proximity to any compact, high-power EUV source. During operation of the Engineering Test Stand (ETS), a full-field, high-power EUVL alpha tool, the silicon/molybdenum multilayer mirrors used as a condenser were eroded by extended exposure to the LPP source. The erosion rate varied considerably, and diagnostic instrumentation on the ETS was not intended to address this issue, so the cause of this erosion was not determined at the time. We present here the results of experiments in which samples of gold, molybdenum, and silicon were exposed to an LPP using a liquid xenon jet as the target. The measured erosion rates suggest a sputtering mechanism. Observations of the plasma environment at the condenser position show the presence of fast ions, which, if they are xenon, have kinetic energies of tens of keV. Such ions would contribute significantly to condenser erosion.

A set of 19 dome‐shaped divertor Langmuir probes similar in design to probes used in JET and JT‐60 has been used successfully in DIII‐D to measure the divertor electron temperature, particle flux, and floating potential. A comparison of the power flux using IR cameras with the particle flux and electron temperature from the probes using collisionless sheath theory has indicated that the particle flux to the divertor surface may be strongly modified by collisions within the magnetic sheath. In order to study this effect, a set of probes have been designed that can be inserted into the divertor plasma using DIMES. Two dome‐shaped probes compare fluxes intercepted both above and within the magnetic sheath of the divertor surface. In addition, a third probe oriented normal to the magnetic field is used to verify the projected area of the probe surface. The probe design accommodates parallel power fluxes up to 50 MW/m2 for 30 ms, allowing for the study of beam‐heated plasmas in DIII‐D.

Full-field imaging with a developmental projection optic box (POB 1) was successfully demonstrated in the alpha tool Engineering Test Stand (ETS) last year. Since then, numerous improvements, including laser power for the laser-produced plasma (LPP) source, stages, sensors, and control system have been made. The LPP has been upgraded from the 40 W LPP cluster jet source used for initial demonstration of full-field imaging to a high-power (1500 W) LPP source with a liquid Xe spray jet. Scanned lithography at various laser drive powers of >500 W has been demonstrated with virtually identical lithographic performance.

Recent development work on the EUV electric capillary discharge source been has focused on two areas: increasing EUV power generation and minimizing debris deposition on plasma facing optics. To achieve high-power operation, a pulser capable of driving the source up to 1.7 kHz and a new high-power lamp have been integrated. An EUV flux of 9 W into p-sr and a 2 percent bandwidth has been generated in burst mode at 1000 Hz. Three additional parametric studies are discussed. The first compares the EUV power generation and spectral output for three different capillary materials. The second study compares the source efficiency for 3 mm and 6 mm length capillaries. And the third parametric study measures the EUV output stability over a one million pulse run. The second focus area has been to increase mirror reflectance lifetimes through the further development of the gas curtain debris mitigation approach. A new gas curtain laboratory has been built with more than a 10x increase in flow capability and a 10x reduction in chamber background pressure. Measurements of the gas curtain efficiency have demonstrated a reduction in particulate deposition rate of at least a factor of eighty.

The advantages of using of table top x‐ray lasers (XRLs) for different applications have been described. Examples of the first successful use of XRLs, the current efforts in applying them and the potential applications where an XRL can be used in future have been discussed. Modeling results showing the possibility of 3–4 times shorter wavelength capillary discharge x‐ray lasers and calculated spectrum of Xe capillary EUV source are presented.

It is long ago recognized that Z-pinches represent very natural medium for x-ray lasers (XRL) due to its favorable geometry and achievable high densities and temperatures. They also are very efficient x-ray sources. One of their variants, the capillary discharges, attracted attention of plasma physics researchers for almost two decades. It has been used for hot dense plasma formation and x-ray lasers [1,2], for transportation of laser beams and XUV radiation generation in x-ray lithography[3,4], for basic Z-pinch research and some others. The combination of efficiency, simplicity and low cost of capillary electrical discharges allowed to scale capillary x-ray lasers to table-top dimensions. In this paper we show the modeling results for next, 3–4 times shorter wavelength x-ray lasers. As an efficient x-ray source of line and continuum radiation it can be used for many practically important application in science and technology. In particular, the capillary discharge can appear as powerful potential candidate for emerging XUV microlithography. We present here the results of numerical modeling of spectra and density of Xe EUV source which involved plasma heating and dynamics, detailed atomic kinetics and radiation transport and material ablation physics.

Experimental results are presented from the three-Langmuir probe (LP) diagnostic head of the divertor material evaluation system (DiMES) on DIII-D that confirm the size of the projected current collection area of the LPs, which is essential for properly measuring ion saturation current density (Jsat) and the sheath power transmission factor (SPTF). Also using the 3-LP DiMES head, the hypothesis that collisional effects on plasma density occurring in the magnetic sheath of the tile are responsible for a lower than expected SPTF is tested and deemed not to have a significant impact on the SPTF. Three-dimensional thermal modeling of wall tiles is presented that accounts for lateral heat conduction, temperature dependence of tile material properties and radiative heat loss from the tile surface. This modeling was developed to be used in the analysis of temperature profiles of the divertor embedded thermocouple (TC) array to obtain more accurate interpretations of TC temperature profiles to infer divertor surface heat flux than have previously been accomplished using more basic one-dimensional methods.

Progress continues on the development of the EUV Electric Capillary Discharge Source. Over the past year we have studied the high average power operational characteristics, used interferometry to measure the in-capillary electron density, and further reduced the debris deposition rate on plasma facing optics. A pulser capable of driving the source at up to 1 kHz was acquired and preliminary testing is in progress. EUV flux and spectral emission measurements were made for pure xenon and xenon-helium mixtures using a new electrode designed to prevent debris from depositing on multilayer optics. Finally, through improvements in capillary and electrode design and material properties coupled with the assistance of mitigation approaches, we have been able to significantly reduce the amount of debris generated by the source thereby increasing the reflectance lifetime of plasma-facing optics.

Two prototype Langmuir probes (based on a JET design) have been installed near the outer strike point on the divertor floor in the DIII‐D tokamak. Results from these graphite probes were used to design a full array of Langmuir probes for the DIII‐D divertor and an additional set of probes for the bias ring of the Advanced Divertor program. The design allows for the projected area of the probes to be relatively insensitive to the magnetic field angle in the divertor region, while projecting the intercepted heat flux over a much larger area. The prototype probes have survived the full separatrix plasma flux during both L‐ and H‐mode operation. I/V characteristics during high power neutral beam operation exhibit electron temperatures approaching 50 eV and large negative floating potentials near the outer separatrix strike point. Edge localized modes cause an increase in the ion flux as large as a factor of 10 above the H‐mode level. As a result, a new set of faster, higher voltage electronics will be used for the full array. The electronics will be capable of floating either with the divertor floor or with the bias ring.

Under appropriate conditions, exposing tungsten to a high flux D plasma creates near-surface blisters and other changes in surface morphology. We have characterized the sizes of blisters formed at different temperatures (147 °C≤Tsurface≤704 °C) and performed a surface analysis to elucidate factors that influence blister formation. Tungsten targets that were exposed to low energy (70 eV) D ions at a flux of 1.1×1022 m−2 s−1 in the tritium plasma experiment (TPE) were considered. We used AES to analyze the surface for evidence of implanted impurities. Blister diameters and heights were quantified using SEM imagery and vertical scanning interferometry. Given the likelihood of D precipitation in blisters, we expect that the data obtained here could be incorporated into a computational model to better simulate the diffusion and desorption of D in W. With this in mind, we present an analysis of thermal desorption profiles showing the release of D from the surface.

The Tokamak Fusion Test Reactor (TFTR) is scheduled to begin DT operation in 1990 with the on-site tritium inventory limited to 5 grams. The physics and chemistry of the in-vessel tritium inventory will impact safety concerns, and also the entire operating schedule of the tokamak. We have investigated plasma-material interaction processes that will affect this first tritium-fueled tokamak. Tritium inventory estimates for TFTR are derived from: (1) laboratory simulation, (2) in-situ plasma measurements, (3) post-run surface analysis, and (4) modeling. This paper presents the results of these investigations, the derivation of a tritium inventory estimate and its uncertainties, and a discussion of its impact. A particular discharge-by-discharge operating schedule has been developed and evaluated. The major source of in-vessel tritium inventory will be codeposition of tritium and eroded carbon onto surfaces. We find that the on-site limit may be approached unless specific inventory reduction techniques are invoked, e.g., discharge cleaning.

The design and fabrication of suitable plasma-facing components for fusion devices is a difficult task with often conflicting requirements. Most operating devices utilize stainless steel as the first wall material and graphite for the main limiters. There are potential problems with the extrapolation of graphite to power producing reactors and, hence, the evaluation of metallic alternatives is important. Prime candidate metals are beryllium and tungsten for plasma interactive components. There are several potential advantages to metallic components over graphite or ceramics. In general, metals have superior fracture toughness, are inert chemically to the hydrogen atmosphere, have a better developed fabrication and joining technology, and better resistance to neutron damage. Metals are limited, however, in their ability to withstand disruption damage and, with the exception of beryllium, require a very low edge temperature to avoid sputtering problems. The current status of research in the areas of erosion, tritium permeation and inventory, thermal stress and fatigue, off-normal events, neutron irradiation and fabrication on these and other metals is reviewed and compared to the prime non-metal choice of graphite.

Following a two‐year operational period the Tokamak Fusion Test Reactor (TFTR) graphite fixed bumper limiter has been examined by a variety of methods. The areal density of metals was mapped in situ by beta backscattering. Several tiles were examined in detail by nuclear‐reaction analysis, Rutherford backscattering, and proton‐induced x‐ray emission to measure areal densities of deuterium and impurities. Some areas of the limiter were found to be covered by deposited material several microns thick. Other areas where the incident plasma flux is higher were much cleaner. Long‐term collection coupons were also examined to characterize deposition on the wall. From these results the total amount of deuterium on the limiter and wall in TFTR is estimated to be ∼1.3×1024 atoms. The implications of this for wall‐pumping effects and future tritium inventory are discussed.

To understand material transport by the plasma in the TFTR tokamak, graphite bumper limiter tiles and metal surfaces have been studied. Detailed measurements of the TFTR inner bumper limiter POCO ™ AXF-5Q graphite tiles indicate areas of net erosion and areas of net deposition. These areas are poloidally asymmetric and on the scale of a bay (l/20th of the torus) repeat regularly toroidally. Finer scale measurements indicate that there are subtle variations in the interaction of the plasma with the wall. Furthermore, a study of the correlation of limiter deposits with the type of discharges in TFTR indicates that the material composition depth distribution was determined by the tokamak operational history. In particular, the relative amounts of carbon, hydrogen, oxygen, and metal in the deposits changed over time, reflecting plasma impurity levels. The outer wall of the vessel was not exposed to direct plasma flux and does not show evidence of erosion.

New and more accurate techniques for finding poloidal and toroidal mode numbers of MHD oscillations from Mirnov coils are presented. It is found that the data for the MHD mode structure are best fit by a model which can accommodate toroidal effects on the mode geometry. The experimentally observed toroidal effects are compared with theoretical expectations. Examples of other physical MHD phenomena occurring during startup on TFTR are also illustrated. An example of toroidally coupled 4,1 and 3,1 modes is presented. An efficient algorithm is described that computes the best fit to Mirnov coil phase data by determining both the m, n numbers and a free toroidal correction parameter (λ', where λ' = 0 → cylindrical). The possibility of the mode being poloidally asymmetric owing to toroidal effects and an off-centre location of the mode justify leaving the λ' parameter free. Since the optimal value of λ' can be obtained automatically with a computer, the need for 'guesswork' is eliminated.

The methods for finding poloidal and toroidal numbers of MHD oscillations from Mirnov coils are reviewed and modified. Examples of various MHD phenomena occurring during start-up on TFTR are illustrated. It is found that the MHD mode structure best fits a model with the toroidal correction included. A new algorithm which finds m,n numbers can accommodate toroidal effects which are manifested in the phase data. The algorithm can find m,n numbers with a given toroidal correction parameter lambda', (lambda' = 0 cylindrical). This algorithm is also used to find the optimal value of lambda' automatically, eliminating the need for ''guesswork.'' The algorithm finds the best parameters to the fit much faster than more conventional computational techniques. 9 refs., 21 figs., 2 tabs.

We are in the process of assembling a database of edge and divertor plasma parameters suitable for use in benchmarking tious 2D models of the scrape-off- layer (SOL) plasma. Also, we are using the Braams B2 code to derive transport coefficients for the edge plssma. In parallel, work is starting on an upgrade to the B2 code that includes padlel current flow and EXB drifts. These efforts are directed at increasing the confidence level of models of the tokamak edge plasma so that we can predict the effect of planned upgrades to DIII-D (e.g., the Advanced Divertor Program) and the performance of next generation machines such as CIT or ITER, where initial design studies show that plasma conditions at the divertor targets can have a large impact on the lifetime and cost of the machine. This report summarizes our recent progress in characterizing the DIII-D SOL plasma and in modeling these data with the the B2 code. Section I contains a brief description of the diagnostics available for characterizing the SOL plasma. In Section II we present our measurements of the SOL parameters for H-mode plasmas. This includes data showing how the divertor plasma parameters (n{sub e}(r), T{sub e}(r), and Q(r)) vary from ohmic to L-mode to H-mode, and power balance for quasi-stationary H-mode plasmas. Section III covers divertor-target heat-flux asymmetries for double and single null operation with forward and reversed toroidal field. In Section IV we show the scaling of L-mode parameters with neutral beam power, and Section V concludes with a summary of the results obtained from the Braams B2 SOL simulation code.

Attention has focused recently on flowing liquid lithium as a first wall for a reactor because of its potentially attractive physics and engineering features. In order to test the suitability of liquid lithium as a plasma facing component, the Current Drive eXperiment - Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has recently installed a fully toroidal liquid lithium limiter. CDX-U is a compact (R = 34 cm, a = 22 cm, B/sub toroidal/ = 2 kG, I/sub p/ =100 kA, T/sub e/(O) /spl sim/ 100 eV, n/sub e/(0) /spl sim/ 5 /spl times/ 10/sup 19/ m/sup -3/ short-pulse (< 25 msec) spherical torus (ST) with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, is 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 (melting point = 180.5/spl deg/C) prior to the experiment. The total area of liquid lithium exposed to the plasma is approximately 2000 cm/sup 2/. The design of the limiter, modifications to CDX-U to accommodate in-vessel inventories of approximately 1 liter of liquid lithium, techniques for loading lithium onto the limiter, and other preparations will be described. CDX-U has previously been successfully operated with a smaller area cm/sup 2/) liquid lithium rail limiter. Diagnostics specific to lithium operations include multichord spectrometry of the 135 /spl Aring/ LiIII line in the core plasma, monitors for neutral lithium light at the lithium limiter, and a fast (10,000 frame per second) camera which monitors motion of the liquid during the discharge. First results of plasma operations with the toroidal liquid lithium limiter will also be given.