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We modeled tungsten–carbon mixed surface evolution, sputtering erosion, and transport for the tungsten coated region of a small angle slot (SAS) divertor design for the DIII-D tokamak. This divertor concept aims to achieve a closed slot dissipative plasma to minimize heat load and surface erosion, and to study high-Z material performance. Our advanced simulations use coupled ITMC-DYN material mixing/response and 3D full kinetic REDEP/WBC erosion/redeposition code packages, with divertor plasma solution from the SOLPS-ITER package with 4 MW power input. The SAS design geometry and resulting in-slot plasma parameters cause significant differences in predicted sputter/transport from a conventional divertor. For 2% C/D incident plasma ratio, an equilibrium mixed C/W surface is attained at ∼30 s of discharge, from wall sputtered carbon transported to the 10 cm long tungsten divertor region. Tungsten remains exposed to the plasma, but the evolved surface composition varies with different C/D ratios. Tungsten is primarily sputtered from the mixed surface by impinging carbon ions in the +1 to +4 charge states, with some self-sputtering. Redeposition of sputtered tungsten to the divertor is significant, ∼80% along the higher plasma temperature attached plasma SAS entrance region, but this is less than the typically near-unity values for a conventional divertor. Plasma-incident carbon is highly backscattered (∼50%) from the mixed surface, with little redeposition (+ ions, also with low redeposition (∼10%–30%). Finally, the modeling shows non-zero but low sputtered tungsten current from the divertor to the core plasma direction. These results appear favorable for effective testing of a tungsten-containing SAS divertor in DIII-D, and extrapolation of mixed-material evolution/response findings to the analogous low-Z/high-Z, Be/W, ITER plasma facing system.

An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based on experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. These techniques may find applications in the future devices.

We analyzed recent DIII-D tokamak tungsten divertor probe experiments using advanced, coupled, sputter erosion/redeposition, plasma, and surface response code packages. Modeling is done for ELMing H-mode, and L-mode plasmas, impinging on various size tungsten deposits on Divertor Material Evaluation System (DiMES) carbon probes. The simulations compute 3D, full kinetic, sub-gyromotion, impurity sputtering and transport, including changes in tungsten surface composition and response due to mixed deuterium and carbon ions irradiation. Per our analysis, ELM (edge localized mode) plasma sputtering in DIII-D mostly involves free-streaming high energy (∼500–1000 eV) D+ and C+6 ions, with high near-surface plasma density. L-Mode sputtering is due to impurity sputtering (C, W) only, with lower density. All cases show complete redeposition of tungsten on the divertor, with significant redeposition on the tungsten spots themselves, and low self-sputtering. Comparison of ELM plasma gross tungsten erosion simulation results with in-situ spectroscopic data is good, as are code/data comparisons of net erosion using post-exposure Rutherford backscattering (RBS) data for the L-mode probes. The analysis, extrapolated to a full tungsten divertor, implies low net erosion and negligible plasma contamination from sputtering. These results support the use of high-Z plasma facing surfaces in ITER and beyond.

We analyze a DIII-D tokamak experiment where two tungsten spots on the removable DiMES divertor probe were exposed to 12 s of attached plasma conditions, with moderate strike point temperature and density (∼20 eV, ∼4.5 × 10 19 m −3 ), and 3% carbon impurity content. Both very small (1 mm diameter) and small (1 cm diameter) deposited samples were used for assessing gross and net tungsten sputtering erosion. The analysis uses a 3-D erosion/redeposition code package (REDEP/WBC), with input from a diagnostic-calibrated near-surface plasma code (OEDGE), and with focus on charge state resolved impinging carbon ion flux and energy. The tungsten surfaces are primarily sputtered by the carbon, in charge states +1 to +4. We predict high redeposition (∼75%) of sputtered tungsten on the 1 cm spot—with consequent reduced net erosion—and this agrees well with post-exposure DiMES probe RBS analysis data. This study and recent related work is encouraging for erosion lifetime and non-contamination performance of tokamak reactor high-Z plasma facing components.

Cold atmospheric pressure plasmas (CAPPs) 1 and electric pulses (EPs) 2 are effective techniques for sterilization, wound treatment, and cancer treatment, whose synergistic effects have only recently been studied 3. While these experiments demonstrated this synergy for different combinations of EPs and CAPPs, the physical phenomena that drive this behavior remain poorly understood. To elucidate these results, we combine models of EP interactions with the cell membrane to model pore formation and ion transport 4 with a model of CAPP generation by a plasma jet 5. The EP/membrane simulation couples the asymptotic Smoluchowski equation for pore formation and the Nernst-Planck equation for ion motion 4. The CAPP model uses coupled sets of governing equations for charged and neutral species continuity, electron energy transport, and self-consistent electrostatic potential 5. We report the results of each model individually and then couple the two models. The membrane voltage induced by the CAPP induced ion buildup at the membrane couples with the EP induced membrane voltage to form membrane pores. The motion of the charged particles from the CAPP model couples with the force from the Nernst-Planck model in the EP model to describe the resulting charged particle motion. We will report the impact of different CAPP and EP parameters and treatment orders on efficacy.

Key wordsFusion plasma/material interaction modelling, plasma simulation/supercomputing. Plasma/material interactions is a critical scientific issue for fusion power, with major potential limitations on plasma core and edge operating parameters. Gaining understanding and predictive capability in this area will require simultaneously addressing complex and diverse physics occurring over a wide range of lengths (angstroms to meters) and times (femtoseconds to days). This requires further development and validation of detailed physics models and computational strategies at each of these scales. The overriding numerical need is for petascale real-time coupling between 3-D material response/evolution/trapping codes, plasma edge/SOL codes, and impurity transport codes. We discuss selected science and modeling/computational challenges in this area and our ideas for achieving these goals; this from the standpoint of our existing simulation codes. c

Plasma interactions with mixed-material surfaces are being analyzed using advanced modeling of time-dependent surface evolution/erosion. Simulations use the REDEP/WBC erosion/redeposition code package coupled to the HEIGHTS package ITMC-DYN mixed-material formation/response code, with plasma parameter input from codes and data. We report here on analysis for a DIII-D Mo/C containing tokamak divertor. A DIII-D/DiMES probe experiment simulation predicts that sputtered molybdenum from a 1 cm diameter central spot quickly saturates (∼4 s) in the 5 cm diameter surrounding carbon probe surface, with subsequent re-sputtering and transport to off-probe divertor regions, and with high (∼50%) redeposition on the Mo spot. Predicted Mo content in the carbon agrees well with post-exposure probe data. We discuss implications and mixed-material analysis issues for Be/W mixing at the ITER outer divertor, and Li, C, Mo mixing at an NSTX divertor.

Experimental observation of net erosion of molybdenum being significantly reduced compared to gross erosion in the divertor of DIII-D is reported for well-controlled plasma conditions. For the first time, gross erosion rates were measured by both spectroscopic and non-spectroscopic methods. In one experiment a net erosion rate of 0.73 ± 0.03 nm/s was measured using ion beam analysis (IBA) of a 1 cm diameter Mo-coated sample. For a 1 mm diameter Mo sample exposed at the same time the net erosion rate was higher at 1.31 nm/s. For the small sample redeposition is expected to be negligible in comparison with the larger sample yielding a net to gross erosion estimate of 0.56 ± 12%. The gross rate was also measured spectroscopically (386 nm MoI line) giving 2.45 nm/s ± factor 2. The experiment was modeled with the REDEP/WBC erosion/redeposition code package coupled to the ITMC–DYN mixed-material code, with plasma conditions supplied by the OEDGE code using Langmuir probe data input. The code-calculated net/gross ratio is =0.46, in good agreement with experiment.

We are studying lithium sputtering, evaporation, transport, material mixing, and surface evolution for the National Spherical Torus Experiment (NSTX) for various surfaces and plasma conditions. Lithium modeling is complex, particularly for NSTX short pulse, multiple material, variable plasma conditions. Cases examined include: (1) liquid lithium divertor (LLD) with planned high heating power/low- D -recycle plasma, (2) non-pumping/high-recycle solid or liquid divertor surface, (3) Li and C impingement on a molybdenum surface. An impurity erosion/redeposition code package is the overall integration tool, with sputter yield and velocity distributions from binary collision mixed-material codes, sheath code input for NSTX boundary conditions, and inputs of plasma edge solutions from external data-calibrated plasma fluid codes. Analysis predictions are generally favorable, showing non-runaway lithium self-sputtering, acceptable net erosion (∼5 nm/s), and moderate edge (∼10%) and core plasma (∼0.1–1%) Li contamination, for the cases studied. A Mo divertor surface is significantly affected by C and Li impingement but with low core plasma contamination predicted for a high-recycle edge plasma.

We analyze an important 15 year old Alcator C-MOD study of campaign-integrated molybdenum divertor erosion in which the measured net erosion was significantly higher (∼X3) than originally predicted by a simple model [1] , [2] . We perform full process sputtering erosion/redeposition computational analysis including the effect of a possible RF induced sheath. The simulations show that most sputtered Mo atoms are ionized close to the surface and strongly redeposited, via Lorentz force motion and collisional friction with the high density incoming plasma. The predicted gross erosion profile is a reasonable match to MoI influx data, however, the critically important net erosion comparison with post-exposure Mo tile analysis is poor, with ∼X10 higher peak erosion measured than computed. An RF sheath increases predicted erosion by ∼45%, thus being significant but not fundamental. We plan future analysis.

Evolution of the ITER outer divertor target surface is analyzed for the case of wall-sputtered beryllium transported to an initially tungsten divertor. Coupled codes for the convective edge plasma, impurity transport, and mixed-surface sputtering, give the wall-source beryllium flux to the divertor, the sputtering erosion/redeposition response, and the resulting beryllium surface content/growth. The analysis shows zero net Be growth over most (∼80%) of the divertor, due to high re-sputtering and reflection—with an equilibrium Be/W surface quickly formed (∼10 s)—but with a region near the strike point having substantial (∼1 nm/s) growth.

Plasma fluxes to the divertor region in ITER near the magnetic separatrix have been modeled extensively in the past. The smaller, but potentially very important fluxes to the main chamber and outer divertor regions are the focus of the present paper. Two main additions to the usual transport modeling are investigated: namely, convective radial transport from intermittent, rapidly propagating ‘blob’ events, and inclusion of the magnetic flux-surface region beyond the second X-point that actually contacts the main-chamber wall. The two-dimensional fluid transport code UEDGE is use to model the plasma, while the energy spectrum of charge-exchange neutrals to the main chamber wall is calculated by the DEGAS 2 Monte Carlo code. Additionally, the spatial distribution of beryllium sputtered from the main chamber wall is determined in the fluid limit.

The US Advanced Limiter-divertor Plasma-facing Systems (ALPS) program is developing the science of liquid metal surface divertors for near and long term tokamaks. These systems may help solve the d...

We analyzed plasma surface interaction issues for the planned Module-A static liquid lithium divertor for NSTX using coupled codes/models describing the plasma edge, divertor temperature, and erosion/redeposition, with input data from tokamak and laboratory experiments. A 300 nm lithium pre-shot deposited coating will strongly pump impinging D + ions. This yields a low-recycle SOL plasma with high plasma temperature, Te 200–400 eV, low density, Ne 1– 3 · 10 18 m 3 , and peak heat loads of 8–20 MW/m 2 , for 2–4 MW core plasma heating power. This regime has advantages for the NSTX physics mission. Peak surface temperature can be held to an acceptable 6470 � C with moderate strike point sweeping (10 cm/s) using a carbon (for 2 MW) or Mo/Cu or W/Cu substrate (2–4 MW). Erosion/redeposition analysis shows acceptable coating lifetime for a 2 s pulse and low core plasma contamination by sputtered lithium. � 2004 Elsevier B.V. All rights reserved.

We developed and applied models for overall (atom + ion) sputtering and sputtered Li + transport of liquid lithium tokamak divertor surfaces. The model/analysis has four parts: (1) a temperature-dependent data-calibrated empirical/code (TRIM) model of lithium sputtering by D + and Li + as a function of incident particle energy and angle; (2) temperature and energy-dependent molecular dynamics (MolDyn) modeling using an effective interionic pair potential of surface-reflected redeposited Li + ; (3) analytical model of reflected lithium charge state; and (4) analytic model of Li + near-surface emission/redeposition cascade. We predict: (1) strong temperature dependence of sputter yields, (2) reflection coefficients of order 50% (thermal energies) and 10% (hyperthermal energies), (3) reflected lithium charge fractions of 10–30% near 1 eV incidence, and (4) enhanced but non-runaway Li emission for the studied surface temperature range between 473 and 653 K.

The Advanced Limiter Test (ALT) project is the focus of a fruitful and intense International Energy Agreement collaboration on TEXTOR. The pump limiter is a mechanical boundary that is laid out for taking the full heat load of TEXTOR, namely 8 MW (assuming 2 MW radiated power) for 10 s, and provides a pumping efficiency of at least 5% of the working gas. This layout is adopted from the requirements of a fusion reactor: It is mandatory to remove both the full power that is convected to the limiter or divertor and the helium ash that is generated in the fusion process. In order to obtain pumping for all gases, the ALT-II is equipped with turbomolecular pumps. A short description of ALT-II is given, and the power and particle fluxes to the limiter surface and into the exhaust scoops are discussed. Requirements of the helium removal rate for a reactor and relevant measurements are discussed, and particle removal and the power distribution to the limiters are treated. Related topics of the ALT-II program were hydrogen recycling and the measurement of turbulence-induced anomalous particle transport in the plasma edge.

We analyzed beryllium first wall sputtering erosion, sputtered material transport, and T/Be codeposition for a typical nextgeneration tokamak design—the fusion ignition research experiment (FIRE). The results should be broadly applicable to any future tokamak with a beryllium first wall. Starting with a fluid code scrapeoff layer attached plasma solution, plasma D 0 neutral fluxes to the wall and divertor are obtained from the DEGAS2 neutral transport code. The D + ion flux to the wall is computed using both a diffusive term and a simple convective transport model. Sputtering coefficients for the beryllium wall are given by the VFTRIM-3D binary-collision code. Transport of beryllium to the divertor, plasma, and back to the wall is calculated with the WBC+ code, which tracks sputtered atom ionization and subsequent ion transport along the SOL magnetic field lines. Then, using results from a study of Be/W mixing/sputtering on the divertor, and using REDEP/WBC impurity transport code results, we estimate the divertor surface response. Finally, we compute tritium codeposition rates in Be growth regions on the wall and divertor for D–T plasma shots using surface temperature dependent D–T/Be rates and with different assumed oxygen contents. Key results are: (1) peak wall net erosion rates vary from about 0.3 nm s −1 for diffusion-only transport to 3 nm s −1 for diffusion plus convection, (2) T/Be codeposition rates vary from about 0.1 to 10.0 mg T s −1 depending on the model, and (3) core plasma contamination from wall-sputtered beryllium is low in all cases (< 0.02%). Thus, based on the erosion and codeposition results, the performance of a beryllium first wall is very dependent on the plasma response, and varies from acceptable to unacceptable. © 2004 Elsevier B.V. All rights reserved.

Several experiments exposing a 5 cm 2 solid and liquid lithium to the divertor plasma of the DIII-D are described. The divertor plasma strikepoint cleans and conditions the initially solid lithium surface. The effective sputtering rate and transport of lithium was found to be acceptable. Lithium has a sputtering yield of solid lithium T e ∼20 eV. The sputtered lithium is ionized in a short distance from the divertor and promptly redeposited. Experiments and modeling show the sputtered lithium is well shielded by the divertor plasma. The behavior of the liquefied lithium was dominated by its macroscopic movement and injection into the plasma caused by J × B magnetohydrodynamic (MHD) forces. Plasma MHD events, such as edge localized modes (ELMs) and locked modes, are found to provide simultaneously the energy to melt the lithium and the transiently high scrape-off layer (SOL) currents to cause the J × B motion. The macroscopic removal of lithium from the small sample, comprising

Lithium-sputtering erosion and transport has been measured in the outer divertor of the DIII-D tokamak. The Divertor Materials Evaluation System (DiMES) mechanism places a 2.5 cm lithium spot as a plasma-facing surface in the divertor. Plasma diagnostics and atomic lithium visible spectroscopy are used to measure the lithium erosion yield near the outer strikepoint (OSP) with swept-plasma parameters of electron temperature 5–25 eV and electron density (0.03–1.80) × 1019 m−3. Solid-phase lithium physical sputtering is measured to be less than 10% (Li/D+). The yield increases with incident energy. Physical sputtering models confirm measurements of sputtered energy and spatial distributions showing skewed angular distributions when the sample is exposed near the OSP and isotropic angular distributions when the sample is exposed to the private flux region. REDEP/WBC modelling of near-surface impurity transport agrees well with experimental measurements showing sputtered lithium neutral atoms effectively ionized about a centimetre away from the Li–DiMES probe surface. A reduction in lithium physical sputtering by a factor of 4–5 is measured when the lithium surface forms an oxide, consistent with the physical sputtering behaviour of most metal-oxides. Although of less significance than lithium atom transport, there is a modelling/data discrepancy regarding lithium ion transport with, e.g. the data showing more asymmetric ion transport than predicted.

New Monte Carlo modelling with the ERO-TEXTOR code has been done for methane transport in the scrape off layer of TEXTOR using recently published rate coefficients of the methane family from Brooks/Janev. Compared to former used rate coefficients from Ehrhardt–Langer the new data result in significantly larger D/XB-values at plasma temperatures above ∼20 eV. The computed amount of re-deposition is at high plasma temperatures up to a factor of 100 smaller with the new rate coefficients. Parameter studies of the transport of locally injected methane and C 2 D 4 in the divertor MkIIa of JET with ERO-JET (adapted version of ERO-TEXTOR) show a strong influence of the location of injection on D/XB. Calculated D/XB-values can increase up to a factor of 1000 if hydrocarbons are puffed at a location inside the SOL instead at the strike point.

Some of the remaining crucial plasma edge physics and plasma–material interaction issues of the ITER tokamak are discussed in this paper, using either modelling or projections of experimental results from existing tokamak operation or relevant laboratory simulations. The paper covers the following subject areas at issue in the design of the ITER device: (1) plasma thermal loads during Type I ELMs and disruptions, ensuing erosion effects and prospects for mitigating measures, (2) control of co-deposited tritium inventory when carbon is used even on small areas in the divertor near the strike points, (3) efficiency of edge and core fuelling for expected pedestal densities in ITER, and (4) erosion and impurity transport with a full tungsten divertor. Directions and priorities of future research are proposed to narrow remaining uncertainties in the above areas.

We have improved the modeling of chemically eroded carbon transport and applied this to JET and ITER. New features are: (1) coupled REDEP and ERO-JET impurity transport code calculations for sputtered wall/divertor carbon, (2) MolDyn molecular dynamics calculations of carbon/hydrocarbon particle reflection at hydrogen-saturated carbon surfaces, (3) ADAS full collisional radiative carbon ion recombination rate coefficients. At low incident particle energies relevant to chemical-erosion (� 0.1–15 eV), we predict high reflection coefficients (� 20–100%), implying more net erosion and T/C codeposition than for full-sticking models. Calculated tritium codeposition rates for the JET MkIIA divertor, using reference chemical erosion yields of order 1% – while higher than previously estimated are well short (X � 1=40) of published data. Possible explanations include much higher chemical erosion yields. � 2003 Elsevier Science B.V. All rights reserved.

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.

Future fusion devices will use tritium and deuterium fuel. Because tritium is both radioactive and expensive, it is absolutely necessary that there be an understanding of the tritium retention characteristics of the materials used in these devices as well as how to recover the tritium. There are three materials that are strong candidates for plasma-facing-material use in next-step fusion devices. These are beryllium, tungsten, and carbon. While beryllium has the disadvantage of high sputtering and low melting point (which limits its power handling capabilities in divertor areas), it has the advantages of being a low-Z material with a good thermal conductivity and the ability to get oxygen from the plasma. Due to beryllium's very low solubility for hydrogen, implantation of beryllium with deuterium and tritium results in a saturated layer in the very near-surface with limited inventory (J. Nucl. Mater. 273 (1999) 1). Unfortunately, there are nuclear reactions generated by neutrons that will breed tritium and helium in the material bulk (J. Nucl. Mater. 179 (1991) 329). This process will lead to a substantial tritium inventory in the bulk of the beryllium after long-term neutron exposure (i.e. well beyond the operation life time of a next-step reactor like ITER). Tungsten is a high-Z material that will be used in the divertor region of next-step devices (e.g. ITER) and possibly as a first wall material in later devices. The divertor is the preferred location for tungsten use because net erosion is very low there due to low sputtering and high redeposition. While experiments are still continuing on tritium retention in tungsten, present data suggest that relatively low tritium inventories will result with this material (J. Nucl. Mater. 290–293 (2001) 505). For tritium inventories, carbon is the problem material. Neutron damage to the graphite can result in substantial bulk tritium retention (J. Nucl. Mater. 191–194 (1992) 368), and codeposition of the sputtered carbon with the tritium from the plasma will produce a layer of carbonaceous material potentially containing kilograms of tritium in the cooler areas of the tokamak (J. Vac. Sci. Technol. A5 (1987) 2286). This paper reviews the tritium retention mechanisms for the three materials discussed above. Tritium removal techniques, including those used in situ to minimize in-vessel inventories as well as those used to reduce contamination prior to waste disposal, are discussed.

This paper reviews sputtering erosion/redeposition modeling for plasma facing surfaces. Basics of the WBC Monte Carlo impurity transport code are described. An example analysis is shown for erosion of a liquid tin fusion reactor divertor. Multiyear erosion studies of candidate divertor and first wall coating materials (Li, Be, C, W, etc.) are summarized—showing generally serious erosion concerns for low-Z solid materials, and encouraging results for high-Z materials and liquid divertor surfaces. Future goals are discussed, e.g. for supercomputer modeling.

Plasma/material (PMI) interactions is the most critical engineering issue for fusion power development. Key concerns include the lifetime of plasma facing components (PFC's) due to steady state sputter erosion, plasma contamination by eroded material, tritium codeposition in redeposited material, and plasma operating limits due to these factors.

The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

A design study of a fusion ignition research experiment (FIRE) is underway to investigate and assess near term opportunities for advancing the scientific understanding of self-heated fusion plasmas. The emphasis for the FIRE program is on understanding the behavior of plasmas dominated by alpha heating ( Q ≥5). Study activities have focused on the technical evaluation of a compact, high field, highly shaped tokamak. One of the key issues for the design is to find suitable plasma facing components (PFCs). We have investigated a variety of plasma edge and divertor conditions ranging from reduced recycling high heat flux conditions (attached) to reduced heat flux detached operation. The inner divertor detaches easily while impurities must be added to the outer divertor to achieve detachment. The outer divertor and private space baffle will have to be actively cooled. The plasma-facing surface of the divertor is tungsten bonded to a CuCrZr heat sink. The remainder of the PFCs are beryllium coated copper attached to the vacuum vessel. Plasma current disruptions impose strong constraints on the design. Appreciable PFC surface melting and evaporation and onset of ‘plasma shielding’ are expected. The forces induced on the PFC due to disruptions determine the size of the attachment of the PFC to the vacuum vessel.

An operational scenario has been demonstrated on the DIII-D tokamak where the graphite covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplished using a low temperature (detached) divertor plasma that eliminates physical sputtering. Likewise, the carbon influx arising from chemical erosion is found to be very low in the detached divertor, although uncertainties exist concerning chemical erosion yield due to the unknown effect of detachment on hydrocarbon transport. Near strike point regions, the rate of carbon deposition is ≈ 3 cm/burn-year, with a corresponding hydrogenic co-deposition rate greater than 1 kg/(m2burn-year); rates which are problematic for steady state fusion reactors. The carbon net deposition rate in the divertor is consistent with carbon arriving from the core plasma region. Carbon ion influx from the main wall is measured to be relatively large in the high density detached regime and is of sufficient magnitude to account for the deposition rate in the divertor. Divertor redeposition is, therefore, determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D with detachment, no significant reduction is found in the core plasma carbon density, illustrating the importance of non-divertor erosion and the complex coupling between erosion/re-deposition and impurity plasma transport.

The phenomenon of runaway surfaces overheating in fusion experiments due to plasma-sheath/surface-emission coupling has been analyzed. A semi-analytic model has been used as well as the BPHI-3D kinetic sheath code (J.N. Brooks, D. Naujoks, Phys. Plasmas 7 (2000) 2565) coupled with the THERM code which solves the non-stationary heat conduction equation also in 3D geometry. Runaway heating due to initial overheating and subsequent sheath breakdown and superheat has been analyzed for two materials – lithium and carbon. For typical liquid lithium divertor conditions the critical exposure time for thermal runaway is of order 10 ms – generally greater than plasma transient periods (e.g., ELMS) or flowing liquid exposure times. Critical exposure times for carbon are much longer (∼1–2 s), as expected due to thermal property differences, and this may explain various `hot spot' formations in carbon systems. It is also shown that, especially for carbon materials, effects such as flake formation and deterioration of heat conductivity can play a crucial role.

A sputtering erosion/redeposition analysis was performed for three candidate tokamak fusion reactor liquid divertor surfaces–lithium, tin–lithium (Sn80Li20), and flibe (LiF+BeF2 salt). The analysis uses coupled edge-plasma, impurity-transport, and sputtering codes (UEDGE/WBC/VFTRIM), and available sputtering data. A pure-lithium surface strongly absorbs impinging D–T ions-this results in a high temperature, low density, (∼200 eV, ∼ 1×10 19 m −3 ) low-recycle plasma edge regime. Lithium appears to perform well in this regime. Although overall sputtering is high, self-sputtering is finite. Most (∼95%) of the sputtered lithium is confined to the near-surface region and redeposited on the divertor with the remainder (∼5%) also being redeposited after transport in the scrape-off layer. Lithium core plasma contamination is low (∼10 −4 Li / D – T ) . Tin–lithium and flibe would likely operate in a high-recycle regime (e.g., 30 eV, 3×10 20 m −3 ). Erosion/redeposition performance of these materials is also good, with finite self-sputtering and negligible core plasma contamination predicted, but with some concern about changing surface composition due to different constituent element redeposition distances.

A design study of a Fusion Ignition Research Experiment (FIRE) is underway to investigate and assess near term opportunities for advancing the scientific understanding of self-heated fusion plasmas. The emphasis for the FIRE program is on understanding the behavior of plasmas dominated by alpha heating (Q ≥ 5). Study activities have focused on the technical evaluation of a compact, high field, highly shaped tokamak. One of the key issues for the design is to find suitable plasma facing components (PFCs). We have investigated a variety of plasma edge and divertor conditions ranging from reduced recycling high heat flux conditions (attached) to reduced heat flux detached operation. The inner divertor detaches easily while impurities must be added to the outer divertor to achieve detachment. The outer divertor and private space baffle will have to be actively cooled. The plasma-facing surface of the divertor is tungsten bonded to a CuCrZr heat sink. The remainder of the PFCs are beryllium coated copper attached to the vacuum vessel. Plasma current disruptions impose strong constraints on the design. Appreciable PFC surface melting and evaporation and onset of plasma shielding are expected. The forces induced on the PFC due to disruptions determine the size of the attachment of the PFC to the vacuum vessel.

Erosion of the first-wall and divertor, and distribution of eroded material in combination with tritium codeposition (primarily with eroded carbon) over many pulses, remain critical issues for the design, operation, and safety of a long-pulse next-step fusion device, such as ITER. These issues are currently being investigated by experiments in tokamaks and in laboratories, as well as by modelling. In this study, we analyse erosion (e.g., by sputtering, ELMs, and off-normal transients) and codeposition effects in the reduced-size ITER device, called `ITER-FEAT', with a strike-point carbon divertor target and metallic walls, for a `semi-detached' edge plasma regime using two-dimensional profiles of plasma edge parameters, modelled by the code B2-EIRENE. This paper accompanies the overview paper given by G. Janeschitz et al. [Plasma wall interactions in ITER-FEAT, these Proceedings]. Tritium codeposition with chemically eroded carbon still presents removal/control challenges, albeit to a somewhat lesser extent than in the 1998 ITER design, and demands efficient tritium inventory removal/control techniques. Due to numerous model uncertainties, not the least of which are the plasma solutions themselves, our intent is to provide a scoping analysis, defining trends and suggesting further research needs.

The possible reduction of the chemical erosion yield of carbon Ychem with increasing hydrogenic ion flux is a critical issue for the use of carbon based plasma facing materials in future high flux confinement devices. The PISCES linear plasma device is used to assess carbon Ychem through nearly an order of magnitude scan in flux (≤ 1023 s-1 m-2) and plasma density, while controlling other exposure parameters that are known to affect Ychem; namely, incident ion energy and surface temperature. Two independent techniques are used to measure Ychem, CD band emission spectroscopy and sample mass loss. Both techniques are carefully benchmarked using a combination of methane injection experiments and modelling. The key results of this study are that the CD molecular band photon efficiency is sensitive to the plasma density at fixed electron temperature and that the redeposition efficiency of hydrocarbons increases with density, thereby decreasing the measured net erosion. After taking these trends into account, no measurable flux dependence of Ychem is found from either of the two measurement techniques, with Ychem ~ 3-5% for an incident deuterium ion energy of 30 eV. Independent scans of incident energy and surface temperature confirm previously measured (by ion beams with relatively low fluxes) Ychem dependences on these parameters. These results together suggest that apparent reductions of Ychem in plasma devices with increasing flux are more likely to be attributed to either increasing redeposition efficiency with density, changing dissociation/excitation/transport properties of the hydrocarbons near the surface, decreasing incident energy or some combination thereof.

The advanced limiter-divertor plasma-facing systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter:divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and divertors are a peak heat flux of \ 50 MW:m 2 , elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (40%). The evaluation of various options is being conducted through a combination of laboratory experiments, www.elsevier.com:locate:fusengdes

A model of collisional processes of hydrocarbons in hydrogen plasmas has been developed to aid in computer modeling efforts relevant to plasma–surface interactions. It includes 16 molecules (CH up to CH4, C2H to C2H6, and C3H to C3H6) and four reaction types (electron impact ionization/dissociative ionization, electron impact dissociation, proton impact charge exchange, and dissociative recombination). Experimental reaction rates or cross sections have been compiled, and estimates have been made for cases where these are not available. The proton impact charge exchange reaction rates are calculated from a theoretical model using molecular polarizabilities. Dissociative recombination rates are described by the equation A/TB where parameter A is fit using polarizabilities and B is estimated from known reaction rates. The electron impact ionization and dissociation cross sections are fit to known graphs using four parameters: threshold energy, maximum value of the cross section, energy at the maximum, and a ...

Recent progress towards an increased understanding of the physical processes in the divertor and scrape-off layer (SOL) plasmas in DIII-D has been made possible by a combination of new diagnostics, improved computational models and changes in divertor geometry. The work focused primarily on ELMing H mode discharges. The physics of partially detached divertor plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, was studied in two dimensions, and a model of the heat and particle transport was developed that includes conduction, convection, ionization, recombination and flows. Plasma and impurity particle flows were measured with Mach probes and spectroscopy and compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation was increased in the divertor and SOL with `puff and pump' techniques using SOL D2 puffing, divertor cryopumping and argon puffing. The important physical processes in plasma-wall interactions were examined with a DiMES (divertor material evaluation system) probe, plasma characterization near the divertor plate and the REDEP code. Experiments comparing single null plasma operation in baffled and open divertors demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H mode with pumping and baffling resulted in a reduction in H mode core densities to ne/nGr ≈ 0.25 (with nGr the Greenwald density). Divertor particle exhaust and heat flux were studied as the plasma shape was varied from a lower single null to a balanced double null, and finally to an upper single null.

A comprehensive modelling effort to analyse sputtering erosion and redeposition in the DIII-D/DiMES 70, 71 and 79 experiments using metal films (beryllium, vanadium, molybdenum, tungsten) on a carbon divertor probe has been performed. These materials were exposed at the outer strike point of an attached H mode plasma with peak at Te ~75 eV. The analysis uses coupled impurity transport (REDEP, WBC) and related codes with inputs of measured plasma parameters. The code output was compared with measured erosion and redeposition profiles, and other data. The predicted redeposition profiles for beryllium, carbon, molybdenum and tungsten agree well with the data. Beryllium and carbon exhibit longer transport distances compared with those of high-Z metals - primarily due to longer mean free paths for sputtered atom ionization. Photon emission calculations for beryllium also compare well with the data, thereby tending to validate coupled models for plasma parameters, impurity transport and atomic data. For vanadium the comparison of the code with the data varies from poor to fair depending on the ionization model used. For most metal films, the absolute erosion is less than that predicted for the `pure' metal, a fact we attribute to the effect of a carbon overlay or mixture. For carbon the predicted peak net erosion rate (~4 nm/s) is approximately 5 times less than the gross rate, and this is confirmed by the data. Carbon net erosion and core contamination result primarily from the quiescent (intra-ELM) period oblique incidence deuterium physical sputtering and self-sputtering. ELM period sputtering and chemical erosion appear to play a small role in net erosion in these plasmas.

Tritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER. We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation. We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER – and more generically in any other next-step experimental fusion facility fuelled with tritium – the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

The Tokamak Fusion Test Reactor (TFTR) tritium retention experience is reviewed and the data related to models of plasma surface interactions. Over 3.5 years of TFTR deuterium-tritium operations, approximately 51% of the tritium injected into TFTR was retained in the torus. Most of this was subsequently recovered by glow discharges and air ventilation. Co-deposition rates for representative conditions in tritium operation were modeled with the BBQ code. The calculations indicate that known erosion mechanisms and subsequent co-deposition are sufficient to account for the order of magnitude of retention.

We are analyzing erosion and tritium codeposition for ITER, DIII-D, and other devices with a focus on carbon divertor and metallic wall sputtering, for detached and semi-detached edge plasmas. Carbon chemical-sputtering hydrocarbon-transport is computed in detail using upgraded models for sputtering yields, species, and atomic and molecular processes. For the DIII-D analysis this includes proton impact and dissociative recombination for the full methane and higher hydrocarbon chains. Several mixed material (Si-C doping and Be/C) effects on erosion are examined. A semi-detached reactor plasma regime yields peak net wall erosion rates of {approximately}1.0 (Be), {approximately}0.3 (Fe), and {approximately}0.01 (W) cm/burn-yr, and {approximately}50 cm/burn-yr for a carbon divertor. Net carbon erosion is dominated by chemical sputtering in the {approximately}1-3 eV detached plasma zone. Tritium codeposition in divertor-sputtered redeposited carbon is high ({approximately}10-20 g-T/1000 s ). Silicon and beryllium mixing tends to reduce carbon erosion. Initial hydrocarbon transport calculations for the DIII-D DiMES-73 detached plasma experiment show a broad spectrum of redeposited molecules with {approximately}90% redeposition fraction.

Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. For attached divertors (Te > 40 eV), outer strike point (OSP) erosion rates exceed 10 cm/exposure-year, physical sputtering is dominant and the effective sputtering yield, Y, is greater than 10%. In detached divertors (Te

The purpose of the ALPS program is to identify and evaluate advanced limiter/divertor systems that will enhance the attractiveness of fusion power. The highest priority goals at present are achievi...

The mission of the Divertor Materials Evaluation System (DiMES) in DIII-D is to establish an integrated data base from measurements in the divertor of a tokamak in order to address some of the ITER and fusion power reactor plasma material interaction issues. Carbon and metal coatings of Be, W, V, and Mo were exposed to the steady-state outer strike point on DIII-D for 4--18 s. These short exposure times ensure controlled exposure conditions, and the extensive arrays of DIII-D divertor diagnostics provide a well-characterized plasma for modeling efforts. Post-exposure analysis provides a direct measure of surface material erosion rates and the amount of retained deuterium. For carbon, these results match closely with the results of accumulated carbon deposition and erosion, and the corresponding deuterium retention of long term exposure tiles in DIII-D. Under the carbon-contaminated background plasma of DIII-D, metal coatings of Be, V, Mo, and W were exposed to the steady-state outer strike point under ELMing and ELM-free H-mode discharges. The rate of material erosion and deuterium retention were measured. As expected, W shows the lowest erosion rate at 0.1 mm/s and the lowest deuterium uptake of 2 {times} 10{sup 20}/m{sup 2}.

New data on hydrogen plasma isotopes retention in beryllium and tungsten are now becoming available from various laboratories for conditions similar to those expected in the International Thermonuclear Experimental Reactor (ITER) where previous data were either missing or largely scattered. Together with a significant advancement in understanding, they have warranted a revisitation of the previous estimates of tritium inventory in ITER, with beryllium as the plasma facing material for the first-wall components, and tungsten in the divertor with some carbon-fibre-composites clad areas, near the strike points. Based on these analyses, it is shown that the area of primary concern, with respect to tritium inventory, remains codeposition with carbon and possibly beryllium on the divertor surfaces. Here, modelling of ITER divertor conditions continues to show potentially large codeposition rates which are confirmed by tokamak findings. Contrary to the tritium residing deep in the bulk of materials, this surface tritium represents a safety hazard as it can be easily mobilised in the event of an accident. It could, however, be possibly removed and recovered. It is concluded that active and efficient methods to remove the codeposited layers are needed in ITER and periodic conditioning/cleaning would be required to control the tritium inventory and avoid exhausting the available fuel supply. Some methods which could possibly be used for in-situ cleaning are briefly discussed in conjunction with the research and development work required to extrapolate their applicability to ITER.

Sputtering erosion is a key performance issue for beryllium coatings in future fusion reactors. Past analysis of this subject for various edge plasma regimes—as briefly reviewed here—shows acceptable erosion only for low duty factor devices. Erosion analysis is now focusing on a very low boundary temperature (5 2 eV) ‘detached plasma’ regime for the International Thermonuclear Experimental Reactor (ITER) Vertical Director design. Erosion and tritium codeposition calculations are made for this regime using a coupled WBC:REDEP:DEGAS code analysis. The analysis includes calculations of low energy, oblique incidence, sputtering yields using the VFTRIM-3D code calibrated to fit recent data on D‐Be sputtering. We find that beryllium performs better than carbon for detached conditions due to the absence of chemical sputtering. Beryllium performs worse than vanadium or tungsten. Net erosion rates of beryllium coated first wall components, due mainly to physical sputtering by charge exchange D‐T neutrals, vary from 0 to 2 cm per burn-year along the wall. Tritium co-deposition in redeposited beryllium surface layers may be high. © 1997 Elsevier Science S.A.

The authors analyzed sputtering erosion and tritium codeposition for the ITER vertical target divertor design using erosion and plasma codes (WBC/REDEP/DEGAS+) coupled to available materials data. Computations were made for a beryllium, carbon, and tungsten coated divertor plate, and for three edged plasma regimes. New data on tritium codeposition in beryllium was obtained with the TPE facility. This shows codeposited H/Be ratios of the order of 10% for surface temperatures {le} 300 C, beryllium thereby being similar to carbon in this respect. Hydrocarbon transport calculations show significant loss (10--20%) of chemically sputtered carbon for detached conditions (T{sub e} {approx} 1 eV at the divertor), compared to essentially no loss (100% redeposition) for higher temperature plasmas. Calculations also show a high, non-thermal, D-T molecular flux for detached conditions. Tritium codeposition rates for carbon are very high for detached conditions ({approximately} 20g-T/1000 s discharge), due to buildup of chemically sputtered carbon on relatively cold surfaces of the divertor cassette. Codeposition is lower ({approximately} 10X) for higher edge temperatures ({approximately} 8--30 eV) and is primarily due to divertor plate buildup of physically sputtered carbon. Peak net erosion rates for carbon are of order 30 cm/burn-yr. Erosion and codeposition rates for beryllium are much lower than for carbon at detached conditions, but are similar to carbon for the higher temperatures. Both erosion and tritium codeposition are essentially nil for tungsten for the regimes studied.

The DiMES (Divertor Material Evaluation Studies) mechanism allows insertion of material samples to the lower divertor floor of the DIII-D tokamak. The main purpose of these studies is to measure erosion rates and redeposition mechanisms under tokamak divertor plasma conditions in order to obtain a physical understanding of the erosion/redeposition processes and to determine its implications for fusion power plant plasma facing components. Thin metal films of Be, W, V, and Mo, were deposited on a Si depth-marked graphite sample and exposed to the steady-state outer strike point on DIII-D. A variety of surface analysis techniques are used to determine the erosion/redeposition of the metals and the carbon after 5--15 seconds of exposure. These short exposure times ensure controlled exposure conditions and the extensive array of DIII-D divertor diagnostics provide a well characterized plasma for modeling efforts. Erosion rates and redeposition lengths are found to decrease with the atomic number of the metallic species, as expected. Under these conditions, the peak net erosion rate for carbon is {approximately} 4 nm/s, with the erosion following the ion flux profile. Comparisons of the measured carbon erosion with REDEP code calculations show good agreement for both the absolute net erosion rate and its spatial variation. Measured erosion rates of the metals are smaller than predicted for sputtering from a bare metal surface, apparently due to effects of carbon deposition on the metal surface. Visible spectroscopic measurements of singly ionized Be have determined that the erosion process reaches steady-state during the exposure.

In this experiment, a molybdenum marker embedded at the center of a graphite substrate is used to study the transport of Mo over the surface under redeposition conditions. The sputtered Mo atoms undergo a number of events (excitation, ionization, de-excitation) in front of the sample before being redeposited. These events can be linked to observable quantities measured by plasma diagnostics. Plasma density and electron temperature are measured by Langmuir probes. Optical emission spectroscopy is used to obtain neutral and ionized Mo and Ar emission profiles in the plasma column. New techniques are introduced to evaluate both excitation and ionization rate coefficients from the measured Mo I line intensity profiles. The measured coefficients are compared to theoretical calculations and a good agreement is found. After plasma exposure, energy dispersive X-ray analysis (EDX) is used to measure the Mo impurity concentration on the graphite surface. These concentrations are then compared to simulated redeposition flux obtained by the WBC code.

Near‐surface recycling of neon and argon atoms and ions at a divertor has been studied using impurity transport and surface interaction codes. A fixed background deuterium–tritium plasma model is used corresponding to the International Thermonuclear Experimental Reactor (ITER) [ITER EDA Agreement and Protocol 2, ITER EDA Documentation Series No. 5 (International Atomic Energy Agency, Vienna, 1994)] radiative plasma conditions (Te≤10 eV). The noble gas transport depends critically on the divertor surface material. For low‐Z materials (Be and C) both neon and argon recycle many (e.g., ∼100) times before leaving the near‐surface region. This is also true for an argon on tungsten combination. For neon on tungsten, however, there is low recycling. These variations are due to differences in particle and energy reflection coefficients, mass, and ionization rates. In some cases a high flux of recycling atoms is ionized within the magnetic sheath and this can change local sheath parameters. Due to inhibited backfl...