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The Prototype-Material Plasma Exposure eXperiment (Proto-MPEX) is a linear plasma device to test the plasma source concepts for the future Material Plasma Exposure eXperiment (MPEX). MPEX is planned to address plasma material interactions under fusion reactor divertor conditions as will occur in devices like DEMO, including the plasma exposure of a-priori neutron irradiated materials. This requires a novel linear plasma device with a plasma source that can produce high electron densities along with heating systems to obtain high electron and ion temperatures. In Proto-MPEX a high-power helicon source was tested together with electron heating (28 GHz microwaves) and ion cyclotron heating (8.5–12 MHz). Electron densities in excess of 1 × 1020 m−3 were produced with the helicon antenna. Electron temperatures can reach 20 eV and are typically below 6 eV at high densities (>5 × 1019 m−3). Ion heating has been demonstrated at low electron densities ( 2 × 1019 m−3).

The availability of future fusion devices, such as a fusion nuclear science facility or demonstration fusion power station, greatly depends on long operating lifetimes of plasma facing components in their divertors. ORNL is designing the Material Plasma Exposure eXperiment (MPEX), a superconducting magnet, steady-state device to address the plasma material interactions of fusion reactors. MPEX will utilize a new high-intensity plasma source concept based on RF technology. This source concept will allow the experiment to cover the entire expected plasma conditions in the divertor of a future fusion reactor. It will be able to study erosion and redeposition for relevant geometries with relevant electric and magnetic fields in-front of the target. MPEX is being designed to allow for the exposure of a priori neutron-irradiated samples. The target exchange chamber has been designed to undock from the linear plasma generator such that it can be transferred to diagnostics stations for more detailed surface analysis. MPEX is being developed in a staged approach with successively increased capabilities. After the initial development step of the helicon source and electron cyclotron heating system, the source concept is being tested in the Proto-MPEX device. Proto-MPEX has achieved electron densities of more than $4 \times 10^{19}~\text{m}^{-3}$ with a large diameter (13 cm) helicon antenna at 100 kW power. First heating with microwaves resulted in a higher ionization represented by higher electron densities on axis, when compared with the helicon plasma only without microwave heating.

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. The codes B2.5-Eirene and EMC3-Eirene were extensively used for design studies of the planned Material Plasma Exposure eXperiment (MPEX). Effects on the target plasma of the gas fueling and pumping locations, heating power, device length, magnetic configuration and transport model were studied with B2.5-Eirene. Effects of tilted or vertical targets were calculated with EMC3-Eirene and showed that spreading the incident flux over a larger area leads to lower density, higher temperature and off-axis profile peaking in front of the target. The simulations indicate that with sufficient heating power MPEX can reach target plasma conditions that are similar to those expected in the ITER divertor. B2.5-Eirene simulations of the MAGPIE experiment have been carried out in order to establish an additional benchmark with experimental data from a linear device with helicon wave heating.

High density (≥6 × 1019 m−3), low temperature (2–6 eV) helicon discharges in the Prototype Material Plasma Exposure eXperiment (Proto-MPEX) are analyzed with the coupled multifluid plasma, kinetic neutrals code B2.5-Eirene. The interpretative analyses are constrained by data from multiple diagnostics, including Langmuir probes, Mach probes, filterscopes, infrared TV system, Thomson scattering, and baratrons. The objectives of the transport simulations include: investigation of the effects of heating, fueling, and plasma production; pumping, and assumed radial transport models on the calculated density and temperature distributions; plasma flow profiles and power balance. The primary objective in this report is to investigate the effects of the radial transport model in full plasma (the entire length of the plasma column in Proto-MPEX) data-constrained simulations. Results from three assumed forms of the radial transport coefficients are presented, including spatially constant, radially decreasing, and Bohm (D,χ ∼ Te/|B|). The results from each of the three transport coefficient sets agree qualitatively with the core (near axis) data. With the implicit Te dependence, the Bohm coefficients tend to decrease as functions of radius, although not as strongly as the centrally peaked set. The axial variation in the Bohm coefficients is largely due to the axial structure of the magnetic field. The agreement of the simulations and the diagnostic data with the Bohm set indicates that transport in the plasma column of Proto-MPEX is dominated by Bohm diffusion.High density (≥6 × 1019 m−3), low temperature (2–6 eV) helicon discharges in the Prototype Material Plasma Exposure eXperiment (Proto-MPEX) are analyzed with the coupled multifluid plasma, kinetic neutrals code B2.5-Eirene. The interpretative analyses are constrained by data from multiple diagnostics, including Langmuir probes, Mach probes, filterscopes, infrared TV system, Thomson scattering, and baratrons. The objectives of the transport simulations include: investigation of the effects of heating, fueling, and plasma production; pumping, and assumed radial transport models on the calculated density and temperature distributions; plasma flow profiles and power balance. The primary objective in this report is to investigate the effects of the radial transport model in full plasma (the entire length of the plasma column in Proto-MPEX) data-constrained simulations. Results from three assumed forms of the radial transport coefficients are presented, including spatially constant, radially decreasing, and Boh...

Coupled fluid plasma and kinetic Monte Carlo neutrals simulations in a linear configuration are reported. The configuration mimics the tokamak divertor plasma channel contacting a target surface with nearby wall. We calculate the parameters of the source plasmas, 3–5 m from the target, required to produce high recycling target plasmas recently simulated for ITER. It is shown that the source plasma needs to deliver heat fluxes of 10–20 MW/m 2 , ion fluxes of 10 23 /m 2 /s, densities of 2–6 × 10 19 /m 3 , and electron and ion temperatures of 15–30 eV over a plasma radius of 5–6 cm. The neutral H and H 2 fluxes to the vessel wall are calculated to be comparable to those measured in the divertor regions of today’s tokamaks. These results identify some design features for a prospective plasma material test station and the research required for this plasma source.

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

The Prototype Material Plasma Exposure eXperiment (Proto-MPEX) at Oak Ridge National Laboratory is a linear plasma device that combines a helicon plasma source with additional microwave and radio frequency heating to deliver high plasma heat and particle fluxes to a target. Double Langmuir probes and Thomson scattering are being used to measure local electron temperature and density at various radial and axial locations. A recently constructed Mach-double probe provides the added capability of simultaneously measuring electron temperatures ( T e), electron densities ( n e), and Mach numbers (M). With this diagnostic, it is possible to infer the plasma flow, particle flux, and heat flux at different locations along the plasma column in Proto-MPEX. Preliminary results show Mach numbers of 0.5 (towards the dump plate) and 1.0 (towards the target plate) downstream from the helicon source, and a stagnation point (no flow) near the source for the case where the peak magnetic field was 1.3 T. Measurements of par...

Plasma material interaction (PMI) studies are crucial to the successful development of future fusion reactors. Prototype Material Plasma Exposure eXperiment (Proto-MPEX) is a prototype design for the MPEX, a steady-state linear device being developed to study PMI. The primary purpose of Proto-MPEX is developing the plasma heating source concepts for MPEX. A power accounting study of Proto-MPEX works to identify machine operating parameters that could improve its performance, thereby increasing its PMI research capabilities, potentially impacting the MPEX design concept. To build a comprehensive power balance, an analysis of the helicon region has been performed implementing a diagnostic suite and software modeling to identify mechanisms and locations of heat loss from the main plasma. Of the 106.3 kW of input power, up to 90.5% of the power has been accounted for in the helicon region. When the analysis was extended to encompass the device to its end plates, 49.2% of the input power was accounted for and verified diagnostically. Areas requiring further diagnostic analysis are identified. The required improvements will be implemented in future work. The data acquisition and analysis processes will be streamlined to form a working model for future power balance studies of Proto-MPEX.

Data-constrained interpretative analyses of plasma transport in convection dominated helicon discharges in the Proto-MPEX linear device, and predictive calculations with additional Electron Cyclotron Heating/Electron Bernstein Wave (ECH/EBW) heating, are reported. The B2.5-Eirene code, in which the multi-fluid plasma code B2.5 is coupled to the kinetic Monte Carlo neutrals code Eirene, is used to fit double Langmuir probe measurements and fast camera data in front of a stainless-steel target. The absorbed helicon and ECH power (11 kW) and spatially constant anomalous transport coefficients that are deduced from fitting of the probe and optical data are additionally used for predictive simulations of complete axial distributions of the densities, temperatures, plasma flow velocities, particle and energy fluxes, and possible effects of alternate fueling and pumping scenarios. The somewhat hollow electron density and temperature radial profiles from the probe data suggest that Trivelpiece-Gould wave absorpti...

The national high-power advanced torus experiment is a concept for a new facility to address the FESAC theme of ‘taming the plasma-material interface’. This concept exploits the compactness and excellent access provided by low aspect ratio to achieve a high ratio of exhaust power to major radius in order to study the integration of high-performance, long-pulse plasmas with a reactor-relevant high heat flux plasma boundary. Predictions of the scrape-off-layer plasma characteristics are presented, as calculated with the 2D edge modeling code SOLPS. Calculations in a variety of magnetic geometries indicate that very high levels of divertor heat flux can be expected, with peak values far in excess of the power handling capabilities of presently-used materials. Possible methods to reduce the heat flux to acceptable levels are discussed.

Plasma boundary control in stellarators has been shown to be very effective in improving plasma performance and, accordingly, is an important element from the very beginning of the National Compact Stellarator Experiment (NCSX) design. Studies of the magnetic field topology outside the last closed magnetic surface (LCMS) indicate the possibility of many toroidal revolutions of field lines launched within a couple of centimeters of the LCMS. Field line connection lengths, typically in the order of 100 m, should be sufficient to allow for the necessary separation of divertor and separatrix temperatures. In the top and bottom of the bean-shaped cross section (toroidal angle {psi} = 0), afield expansion of >5 is observed, which will help to spread out the heat flux on limiters and divertor plates. Plasma-facing components (PFCs) will be developed systematically according to our respective understanding of the NCSX boundary; the phased PFC development will start out with a set of limiters and has the eventual goal to develop a divertor with all the benefits of impurity and neutrals control. Neutrals calculations have been started to investigate the effect of neutrals penetration at various plasma cross sections, especially at the location of {psi} = 0 deg. Advanced wallmore » conditioning techniques, as employed in other major fusion devices, will be incorporated in the NCSX operation.« less

The availability of future fusion devices such as a Fusion Nuclear Science Facility (FNSF) or DEMO greatly depends on long operating lifetimes of plasma facing components in their divertors. ORNL is designing the Material-Plasma Exposure eXperiment (MPEX), a superconducting magnet, steady-state device to address the plasma material interactions of fusion reactors. MPEX will utilize a new high-intensity plasma source concept based on RF technology. This source concept will allow the experiment to cover the entire expected plasma conditions in the divertor of a future fusion reactor. It will be able to study erosion and re-deposition for relevant geometries with relevant electric and magnetic fields in-front of the target. MPEX is being designed to allow for the exposure of a-priori neutron-irradiated samples. The target transfer cask has been designed to undock from the linear plasma generator such that it can be transferred to diagnostics stations for more detailed surface analysis. MPEX is being developed in a staged approach with successively increased capabilities. After the initial development step of the helicon source and ECH system the source concept is being tested in the Proto-MPEX device (100 kW helicon, 200 kW EBW, 30 kW ICRH). Proto-MPEX has achieved electron densities of more than 4×1019m−3 with a large diameter (13cm) helicon antenna at 100 kW power. First heating with microwaves resulted in a higher ionization represented by higher electron densities on axis, when compared to the helicon plasma only without microwave heating.

Studies of the H-mode pedestal in the DIII-D tokamak are presented. The global energy confinement increases as the plasma pressure on top of the pedestal increases. The best empirical description for a pedestal width parameter is {delta}{sub pe} [proportional to] ({beta}{sub pol}{sup PED}){sup 0.4}, where {delta}{sub pe} is the width of the electron pressure pedestal and {beta}{sub pol}{sup PED} is the poloidal beta at the top of the pedestal. The edge profiles of electron density n{sub e}, electron temperature T{sub e}, and ion temperature T{sub i} can all have different shapes. Thus, a simple width scaling for the edge might not exist, and studies of the physics of individual profiles have been initiated. A model for the n{sub e} profile, based on self-consistent treatment of edge particle sources and edge particle transport, agrees with several experimental observations. The steep gradient region for the T{sub e} profile often extends farther into the plasma than the n{sub e} pedestal step. Magnetohydrodynamic stability provides the ultimate limits to the evolution of the pedestal and usually leads to edge instabilities called edge-localized modes (ELMs). However, the absence of ELMs in a regime called the Quiescent H-mode shows that large pedestals can be produced withoutmore » ELMs.« less

A reference equilibrium for the US National Compact Stellarator Experiment is predicted to be sufficiently close to quasi-symmetry to allow the plasma to flow in the toroidal direction with little viscous damping, yet to have sufficiently large deviations from quasi-symmetry that nonambipolarity significantly affects the physics of the shielding of resonant magnetic perturbations by plasma flow. The unperturbed velocity profile is modified by the presence of an ambipolar potential, which produces a broad velocity profile. In the presence of a resonant magnetic field perturbation, nonambipolar transport produces a radial current, and the resulting j × B force resists departures from the ambipolar velocity and enhances the shielding.

Measurements and modeling of the 2D poloidal D α intensity distribution in DIII-D low density L-mode and medium density ELMy H-mode plasmas indicate that the core plasma is predominately fueled near the divertor X-point region. The neutral deuterium and ion carbon emission were measured in the divertor and inner main chamber scrape-off layer (SOL) using a plasma imaging technique, covering 85% of the poloidal cross-section. Typically, the peak emission in the inner main SOL at the tokamak midplane was three orders of magnitude lower than in the divertor. For discharges with the ion B × ∇ B drift direction toward the lower divertor the UEDGE/DEGAS codes predict strong core plasma fueling from the significantly higher density and lower temperature plasma calculated in the inner divertor leg. The concomitant carbon ion flow reversal in the inner divertor leg enhances the leakage of carbon from the divertor into the main SOL, and hence into the core.

A series of hydrogen and deuterium discharges are analyzed with fluid plasma and Monte Carlo neutrals codes. Comparison of poloidally averaged radial distributions of core neutral density and ionization with analytic solutions of 1-D plasma and neutrals continuity equations support the hypothesis that the width of the density pedestal is largely determined by the neutral source. The increased neutral penetration depth that arises from multiple charge exchange can be included in the analytic model with radially dependent scale lengths. The scale length in the analytic model depends on the neutral fluid velocity which increases across the divertor and pedestal as the neutral atoms charge exchange with the higher temperature background ions. The neutral penetration depth and corresponding density pedestal width depend sensitively on the neutral temperature and the degree of ion-neutral temperature equilibration.

A system of absolutely calibrated Hα detectors for neutral hydrogen monitoring has been designed and implemented for the Helically Symmetric eXperiment. An array of detectors at fixed poloidal angle but many toroidal angles yields a measurement of the toroidal asymmetry in Hα emission and is used for monitoring hydrogen recycling. A multichord array at the toroidal angle of the gas puff allows an accurate determination of neutral gas penetration. Preliminary neutral gas modeling allows for the estimation of the density of atomic and molecular hydrogen from the Hα emission, which are used for particle transport and flow damping studies.

Pedestal studies in DIII-D find that the width of the region of steep gradient in the H-mode density is comparable with the neutral penetration length, as computed from a simple analytic model. This model has analytic solutions for the edge plasma and neutral density profiles, which are obtained from the coupled particle continuity equations for electrons and deuterium atoms. In its range of validity (edge temperature between 40 and 500 eV), the analytic model quantitatively predicts the observed decrease in the width as the pedestal density increases and the observed strong increase in the gradient of the density as the pedestal density increases. The model successfully predicts that L-mode and H-mode profiles with the same pedestal density have gradients that differ by less than a factor of 2. The width of the density barrier, measured from the edge of the electron temperature barrier, is the lower limit for the observed width of the temperature barrier. These results support the hypothesis that particle fuelling is an important part of the physics that determines the structure of the H-mode transport barrier.

Recent measurements of the neutral densities both inside and outside the separatrix near the X-point of the DIII-D tokamak, in both L- and H-mode plasmas, are analysed with the two-dimensional transmission/escape probability neutral transport code GTNEUT and with the two-dimensional Monte Carlo code DEGAS. The predictions of the two codes are in good agreement with each other and agree with the experiment to within the experimental error bars in most cases.

Measurements from balanced or slightly unbalanced, double-null, ELMing H-mode plasmas show that the upper, more closed, divertor strike point particle fluxes are larger than the lower, more open, divertor strike point particle fluxes for comparable conditions. Comparison of the upper and lower outer strike point peak particle fluxes measured with Langmuir probes during scans of the magnetic balance shows an asymmetry around the magnetic balance point which is consistent with both the enhanced upper recycling and the expected direction of private region E × B drifts. Comparison of the inner and outer peak particle fluxes in the upper divertor shows a dependence on magnetic field direction that is also consistent with the private region E × B drift direction.

Fueling with gas and pellet injection from several different locations has been used on the DIII-D tokamak to study core fueling and transport in H-mode and L-mode plasmas. Specific experiments have been carried out to examine the fueling efficiency into DIII-D H-mode plasmas that have periodic edge localized modes (ELMs). The fueling efficiency, defined as the total increase in number of plasma electrons divided by the number of input fuel atoms, is determined by measurements of plasma electron density profiles before and after a fueling pulse. We have found previously that there is a significantly higher fueling efficiency for pellets injected from the inner wall [J. Nucl. Mater. 290 (2001) 398] versus outside midplane pellet injection or gas puffing. In this study we extend this work to include the investigation of gas puffing from the same inner wall injection locations as the pellets to determine if a similar effect may exist. The mechanism for the improved pellet fueling from the inner wall is hypothesized to be a ∇B induced polarization of the pellet cloud leading to an E×B drift [ Phys. Plasmas 7 (2000) 1968 ] in the major radius direction. The possibility of similar physics playing a role in gas puff fueling is examined in these experiments and does not appear to be a significant effect.

The national compact stellarator experiment (NCSX) [EPS 2001, Madeira, Portugal, 18–22 June 2001] is a new fusion project located at Princeton Plasma Physics Laboratory, Princeton, NJ. Plasma boundary control in stellarators has been shown to be very effective in improving plasma performance [EPS 2001, Madeira, Portugal, 18–22 June 2001] and, accordingly, will be an important element from the very beginning of the NCSX design. Plasma-facing components will be developed systematically according to our understanding of the NCSX boundary, with the eventual goal to develop a divertor with all the benefits for impurity and neutrals control. Neutrals calculations have been started to investigate the effect of neutrals penetration at various cross-sections.

Data from an array of eight tangential Dα viewchords, located at the outer midplane of DIII-D, are inverted to obtain atomic neutral density profiles across the separatrix and are compared to results of B2.5 and DEGAS simulations. In H-mode and (low density) L-mode plasma conditions, charge-exchange neutrals reflected and desorbed from the outer midplane region of the vacuum vessel wall can account for the midplane neutral density that is deduced from the Dα data, making it unnecessary to invoke main chamber ion recycling. The midplane measurements and DEGAS calculations are used to validate a slab-model of neutrals in the edge. In this model, cold neutrals from the wall fuel the edge and are attenuated by charge-exchange and ionization, and the results are in agreement with measured neutral profiles.

An analytic model, derived from coupled continuity equations for the electron and neutral deuterium densities, is consistent with many features of edge electron density profiles in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159]. For an assumed constant particle diffusion coefficient, the model shows that particle transport and neutral fueling produce electron and neutral density profiles that have the same characteristic scale lengths at the plasma edge. For systematic variations of density in high-mode (H-mode) discharges, the model predicts that the width of the electron density transport barrier decreases and the maximum gradient increases, as observed in the experiments. The widths computed from the model agree quantitatively with the experimental widths for conditions in which the model is valid. These results support models of transport barrier formation in which the H-mode particle barri...

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1–10 eV and electron densities of . The resulting heat fluxes are about 10 MW . MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to at high magnetic fields of 1.0 T at the target. The experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL's High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.

Radial transport in helicon heated hydrogen plasmas in the MAGnetized Plasma Interaction Experiment (MAGPIE) is studied with the B2.5-Eirene (SOLPS5.0) code. Radial distributions of plasma density, temperature and ambipolar potential are computed for several magnetic field configurations and compared to double Langmuir probe measurements. Evidence for an unmagnetized ion population is seen in the requirement for a convective pinch term in the continuity equation in order to fit the centrally peaked density profile data. The measured slightly hollow electron temperature profiles are reproduced with combinations of on-axis and edge heating which can be interpreted as helicon and Trivelpiece–Gould wave absorption, respectively. Pressure gradient driven radial charged particle diffusion is chosen to describe the diffusive particle flux since the hollowness of the temperature profiles assists the establishment of on-axis density peaking.

The recycling coefficient of the first wall of a fusion device is a space- and time-dependent quantity. In present analyses, it is often taken as one single parameter, without taking account of the vast variations across the surface areas of the vessel. The present paper is an analysis of the spatially dependent recycling coefficient as a function of time, using the DIII-D vacuum vessel as an example. In the first step, an analytical formula is constructed to simulate the recycling coefficient as a function of the trapped particle flux, based on computer simulations (W. Eckstein, Calculated Trapping Curves of D in C, Si, Garching Report IPP 9/33, October 1980). In the second step, the vessel walls are subdivided into a computational grid of 123 segments and for each segment the incident particle flux during a discharge is computed with the B2 plasma and DEGAS neutral codes. The incident particle fluxes and the recycling coefficient formula are then used to analyze the local and temporal recycling behavior of the wall.

Analysis of DIII-D discharge data with fluid plasma and Monte Carlo neutrals transport codes reveals that core particle fueling stays relatively constant between the L-mode and the ELM-free H-mode phase immediately following the L–H transition. This indicates that in the ELM-free phase nearly all of the increase in plasma electron density comes from a decrease in the cross-field transport rate and an increase in the impurity influx. This result differs from conclusions of previous work in that the effects of the thinner H-mode scrape-off-layer do not appear to be as important in a plasma that is fueled primarily from divertor recycling as would be expected if the fueling from limiter recycling were dominant. In both L-mode and H-mode the calculated core particle confinement times are less than, but within 50% of, the corresponding energy confinement times.

Neutral particle densities and energy losses have been measured in the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Their effect on the formation and evolution of the edge barrier which accompanies the enhanced confinement regime are discussed. The neutrals can enter the edge dynamics through the particle, momentum, and energy balance. Neutral densities of up to 5×1016 m−3 have been measured in the edge barrier region. Neutrals enter the local dynamics around most of the periphery, not just at the X-point. High resolution measurements of the ionization profile have been obtained for the region near the separatrix. The profile shifts inside the separatrix as the plasma is making a transition from low-to high-mode confinement (H-mode) regimes, partly accounting for the dramatic rise in edge density. The measured neutral density is large enough to affect the bulk ion momentum by charge exchange, and thereby introduces a negative radial electric field at the edge. At the same time, significant edge heat flux, carried by the neutrals, contributes to the measured power loss. At very high edge densities, this loss mechanism could contribute to quenching H-modes.

Theories predict that neutrals play a role in the low to high (L-H) confinement mode transition in tokamak plasmas via charge exchange damping and other effects. Previous estimates of neutral damping have been based on calculations of the edge neutral density. This work introduces a new method of measuring the neutral density near the X point, where simulations predict it to be a maximum. The technique employed uses Dα light from a TV camera reconstructed onto a poloidal plane, along with Thomson scattering measurements of the electron temperature and density. Measured neutral densities span the range 109-1013cm-3. Good agreement, considering the neutral density error bars, is found between the measurements and the 2-D simulations. This work represents the first step in verifying previous 2-D simulations and in corroborating previous conclusions that the neutral damping is large enough to play a role in the L-H transition process.

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 first-flight neutral transport model to describe the dependence of pump plenum neutral pressure on plasma parameters and magnetic geometry is presented. It is shown that the model is in excellent agreement with neutral pressure data from a low recycling DIII-D tokamak discharge. It is also shown that the main contribution to plenum pressure arises from the part of the ion particle flux profile which is closest to the plenum entrance. This work illustrates the sufficiency of a simple model in divertor plenum hardware design studies to maximize particle exhaust for density control.

In order to study possible effects of neutrals on the power threshold for transition from L-mode to H-mode confinement in DIII-D, the problem of reconstructing neutral particle distributions inside the separatrix from available edge, scrape-off layer, and divertor plasma diagnostic data is addressed. Neutral particle profiles in the shear layer are reconstructed primarily with the DEGAS code using plasma parameters obtained by fitting discharge data with the B2.5 code. A series of discharges in which the ion grad B drift direction is toward the X-point is analyzed. For reversed Bt, the L to H transition power threshold is 2–4 times higher than normal Bt discharges at the same density: one discharge of this type is analyzed. The variation of the L–H transition power threshold divided by the density is seen to be correlated with the neutral density scale length in the shear layer for both directions of the magnetic field. Radial and poloidal neutral particle profile results for three low density discharges in which the L–H transition power threshold varies from 1.3 to 5.3 MW are presented.

To study the effect of edge neutrals on the low-to-high confinement transition threshold, a broad range of plasma discharges has been analyzed. These discharges vary by gas puffing and pumping rates, position of the X point, and line-averaged density. It is shown that the determination of the neutral density (or neutral pressure) in the scrape-off layer (SOL) can give a misleading indication of the neutral population inside the separatrix. An increase of neutral density in the SOL creates an increase of plasma density that, in turn, increases the opacity to the neutrals and results in reduced neutral penetration. At a constant magnetic field, the transition power divided by the density appears to be a function of a single parameter measuring the neutrals effect. From this analysis, this parameter cannot be uniquely identified. For instance, it may be the radial decay length of the neutral profile or the charge-exchange damping rate at about r/a≈0.95. A similar correlation exists between these neutral para...

To assess the effect of edge neutrals on the low-to-high confinement transition threshold, a broad range of plasma discharges has been analysed. From this analysis, the transition power divided by the density, at constant magnetic field, appears to be a function of a single parameter measuring the neutrals' effect. This results suggest that there is a missing parameter linked to the neutrals in the power-threshold scaling laws.

One of the most difficult aspects of performing an equilibrium particle transport analysis in a diverted tokamak is the determination of the particle flux which enters the plasma after recycling from the divertor plasma, the divertor target plates or the vessel wall. An approach which has been utilized in the past is to model the edge, scrape-off layer (SOL), and divertor plasma to match measured plasma parameters and then use a neutral transport code to obtain an edge recycling flux while trying to match the measured divertor D(x emissivity. Previous simulations were constrained by electron density (n{sub e}) and temperature (T{sub e}), ion temperature (T{sub i}) data at the outer midplane, divertor heat flux from infrared television cameras, and n{sub e}, T{sub e} and particle flux at the target from fixed Langmuir probes, along with the divertor D{sub {alpha}} emissivity. In this paper, we present results of core fueling calculations from the 2-D modeling for ELM-free discharges, constrained by data from the new divertor diagnostics. In addition, we present a simple technique for estimating the recycling flux just after the L-H transition and demonstrate how this technique is supported by the detailed modeling. We will show the effect which inaccuracies in the recycling flux have on the calculated particle flux in the plasma core. For some specific density profiles, it is possible to separate the convective flux from the conductive flux. The diffusion coefficients obtained show a sharp decrease near a normalized radius of 0.9 indicating the presence of a transport barrier.

Experiments performed with the Tore Supra vented pump limiter (Vented-PL) demonstrate that a vented structure exhibits a significant particle exhaust capability while allowing moderate heat flux density on the whole head surface. An experimental comparison is presented between a classical pump limiter with throats (Throat-PL) and the test Vented-PL for both deuterium and helium ohmic discharges. It is shown that, without active pumping, the deuterium and helium pressures in the Vented-PL plenum are typically 3 times lower than those measured in the Throat-PL plenum. However, when pumping is activated, this pressure is demonstrated to be high enough to allow the Vented-PL to control actively the plasma density. With only one modular Vented-PL, an exhaust efficiency (defined as the global recycling coefficient reduction) as high as 6% has already been measured for a deuterium plasma

DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D{sub 2} and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D{sub 2} gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design.

In Tore Supra, plasma particle exhaust has been studied by means of modular pump limiters, with either throats (ion collection) or vented structures (neutral collection), and with an ergodic divertor. Detailed study of active plasma density control has been performed in these three configurations for ohmic conditions.

A particle exhaust scheme, employing a cryocondensation pump in the outboard divertor region under a baffle, has been installed and operated in the DIII-D tokamak. The cryopump provides toroidally symmetric pumping at a rate of 30000 to 40000 l/s for D{sub 2} in the pressure range of 1 to 4 mTorr. Pressures in the 2 to 3 mTorr range are routinely observed under the baffle. This translates to particle exhaust throughputs of nearly 100 Torr l/s. The exhaust throughput could be controlled by selecting the position of the plasma strike region with respect to the opening to the baffle chamber. The pump has been used quite effectively for plasma density control. 19 refs., 9 figs., 1 tab.

Evaluations of core fueling rates and global particle confinement times are presented for a single-null ELM-free H-mode discharge in the DIII-D tokamak. The edge plasma transport is modeled with the B2.5 code, and the neutral transport with both the DEGAS code and the internal neutrals model in the B2.5 code. Radial transport coefficients and corresponding plasma particle and heat fluxes into the scrape-off layer (SOL) are determined from power balance and by fitting the electron density and temperature profiles measured by Thomson scattering and the ion temperature profiles from charge-exchange/recombination spectroscopy. Divertor plasma parameters are deduced by fitting infrared television camera (IRTV) data, divertor Langmuir probe measurements, and Dα data. Core fueling rates are calculated with DEGAS and compared to those required by particle balance, with the core efflux determined by B2.5. Global particle confinement times, deduced at four time slices during the 1.8 s ELM-free period, are typically 3–4 times the corresponding measured energy confinement times.

Particle control is playing an increasingly important role in tokamak plasma performance. The present paper discusses particle control of hydrogen/deuterium by wall pumping on graphite or carbonized surfaces, as well as by external exhaust with pumped limiters and pumped divertors. Wall pumping is ultimately a transient effect and by itself not suitable for steady state particle exhaust. Therefore, external exhaust techniques with pumped divertors and limiters are being developed. How wall pumping phenomena interact and correlate with these inherently steady state, external exhaust techniques, is not well known to date. In the present paper, the processes involved in wall pumping and in external pumping are investigated in an attempt to evaluate the effect of external exhaust on wall pumping. Some of the key elements of this analysis are: (1) charge-exchange fluxes to the wall play a crucial role in the core-wall particle dynamics, (2) the recycling fluxes of thermal molecules have a high probability of ionization in the scrape-off layer, (3) thermal particles originating from the wall, which are ionized within the scrape-off layer, can be directly exhausted, thus providing a direct path between wall and exhaust which can be used to control the wall inventory. This way, the wall can be kept in a continuous pumping state in the sense that it continuously absorbs energetic particles and releases thermal molecules which are then removed by the external exhaust mechanism. While most of the ingredients of this analysis have been observed individually before, the present evaluation is an attempt to correlate effects of wall recycling and external exhaust.

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

Particle control experiments in Tore Supra have shown that the change in core plasma density with pumping at the outboard pump limiter amounts only to a small fraction of the total number of particles extracted by the outboard pump limiter system. To develop a model for the core-wall particle exchange at the level of individual basic physical processes, the particle exchange between the core plasma and the wall has been modeled in detail for a series of Tore Supra discharges. Core and scrape-off layer particle balance and wall diffusion calculations were performed using a radial transport code and a 1D wall diffusion code. Core-wall particle exchange for the case of the outboard pump limiter is found to be mediated by the localized charge exchange flux in the near limiter region. The wall particle efflux required for detailed balance does not match that calculated to be available from the near limiter wall region unless some local heating of the wall is assumed.

The DIII-D divertor baffle system was designed to facilitate density control in long pulse H-mode discharges by removing a particle flux equal to the neutral beam fueling rate (∼20 Torrl/s) with a ∼1 mTorr neutral pressure under the baffle (p0). Initial measurements of the baffle pressure indicated that p0∼10 mTorr (without pumping or biasing), a value much in excess of that required for long pulse density control. Radial sweeps of the X-point position have been employed to determine the maximum p0, as well as to establish the dependence of this pressure on geometry. An estimate of the particle equilibration time for the baffle system has been made by studying the baffle pressure response to “giant” ELM effects. “Steady state” experiments in which the X-point position was fixed for ∼2.5 s have also been carried out and steady baffle pressures were observed. The scaling of baffle pressure with plasma parameters has been found to be similar under transient and “steady state” conditions. Detailed modeling of these experiments with the B2, DEGAS, and WDIFFUSE (wall model) codes has been performed.

A series of ohmic discharges with active pumping in the Tore Supra outboard pump limiter has been analyzed with the DEGAS neutrals transport code and an analytic scrape-off layer (SOL) plasma model. Pumping speed and plenum pressure measurements indicated 5–10 Torr 1/s throughput with only modest effects on density (d N core /d t + source rate from ionization and dissociation of wall-desorbed molecules is seen to peak very near the radial position of the limiter throat. Consequently, a strong recycling vortex is created in the region of the limiter, with the ion flux amplified by factors of ≈ 2 at the outer limiter surfaces and > 3 within the limiter throat. The calculations indicate that less than 30% of the pump throughput is due to first generation ions from the core efflux, with the balance from local recycling in the strongly ionizing SOL.

In this paper, we investigate the role of wall conditions during the density pump-out observed in Resonant Magnetic Perturbation (RMP) experiments. We use a 2D fluid code, coupled to a neutral Monte-Carlo code (SOLPS5) to model a reference ELMing H-mode. Next we modify either the recycling at the target plates or the pumping efficiency at the pump entrance. This causes a change in upstream profiles, that alone cannot account for the shape change observed in the RMP H-mode. Therefore, the radial transport must be altered to match the RMP H-mode. By comparing the transport model for the ELMing H-mode and the RMP H-mode, we show that inside the separatrix an increase in transport is necessary to model the effects of the applied perturbed field in RMP H-mode.

A critical issue in long-pulse tokamak discharges is the need for density control and power-handling capability in the presence of wall outgassing, neutral beam injection, and pellet fueling. Direct particle and energy exhaust in Tore Supra is obtained with a system of pump limiters, including six located at the bottom of the machine and a large horizontal module at the outer midplane. This paper focuses on two-dimensional (2-D) modelling of the scrape-off-layer (SOL) and outboard pump limiter, using the MHD fluid code b2 and the neutral transport code DEGAS. Temperature, density, and ion flux data from Langmuir probes in the throat of the limiter are used along with estimates of the power scrape-off-length from infrared camera data to obtain a self-consistent description of the SOL plasma/neutral source distribution within the limiter and throughout the SOL. Good agreement with measured quantities is obtained, and three to four iterations of the b2/DEGAS calculation are necessary for convergence.

Hα emission in TEXTOR is measured at several locations in order to study recycling and global particle confinement with the Advanced Limiter Test-II (ALT-II) pump limiter. The measurements are modeled with the neutral transport code DEGAS for ohmic density scans in deuterium plasmas with carbonized vacuum vessel walls. The core efflux, particle confinement time, diffusion coefficient D⊥, boundary flow asymmetry, and profiles of the core fueling rate by particle recycling are determined for line-averaged densities n ¯ e = ( 1.4 − 3.7 ) × 10 13 c m − 3 . Measured distributions of core plasma density and temperature, together with scrape-off layer (SOL) plasma parameters from the scanning probe, the neutral lithium beam diagnostic, and the Langmuir probes in the ALT-II scoops, describe the plasma environment for the DEGAS Hα simulations. In particular, it is found that radial profiles of ion flux in the SOL, derived from scoop Langmuir probe measurements, give excellent fits to the Hα data. A tangency point particle flux asymmetry favoring the ion drift side of the blade of 2.25 to 1.24 for the density range (1.4–3.7)× 1013 cm−3 is indicated by both the Hα simulations and the edge plasma diagnostics. The particle confinement time varies from approximately 50 to 80 ms for this ohmic density scan.

A principal objective of the collaborative DIII-D Divertor Program (ADP) is to achieve density control in H-mode discharges with edge biasing and with continuous particle exhaust at a rate determined by the external fueling sources (typically 20 Torr{center dot}L/s). The divertor baffle-bias ring system has been optimized for pumping speeds {approx}50,000 L/s with the neutral transport code DEGAS. With an entrance slot conductance of 50,000 L/s, a pumping speed of the same order is required to remove half of the {approx}40 Torr{center dot}L/s that enters the baffle chamber for typical H-mode discharges. Increasing the exhaust fraction with higher pumping speed is self-limiting, owing to the attendant reduction of the recycling flux. The effects of pumping on the plasma core, scrape-off layer (SOL), and divertor have been estimated with a model that self-consistently couples the transport in these regions. The required {approx}50,000 L/s pumping speed can be achieved with either titanium getter pumps or cryopumps. Evaluation of both systems has led to the conclusion that cryopumps will be more compatible with the environment of the DIII-D divertor. 8 refs., 7 figs.