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Detailed measurements of axial electron temperature, Te, profiles in the presheath region were carried out using a Langmuir probe and the line intensity ratio technique for both He I (728.1 nm/706.5 nm) and Be II (467.3 nm/313.1 nm). The results show that Te increases toward the material surface, which contradicts the standard picture that Te is constant along the magnetic field in the sheath-limited regime. While no target bias voltage, Vb, dependence is seen, the Te rise becomes more prominent with decreasing neutral pressure. Similarly, the ion temperature, Ti, evaluated from Doppler broadening of a He II line emission at 468.6 nm is found to increase toward the surface, but also does not depend on Vb. Possible mechanisms of the Te and Ti rise as well as validity of the line intensity ratio technique near the material surface are discussed.

A systematic study of the influence of the deposition conditions on the deuterium retention by and release from co-deposited beryllium layers has been carried out in PISCES-B. Experimental parameters such as the beryllium deposition rate, the incident particle energy and the substrate temperature are shown to affect the level of hydrogen isotope retention in the layers. In addition, the pressure during deposition and the presence of Ar in the plasma are monitored since they influence the layer morphology and the ease of fuel removal by thermal desorption. Consequences for both the tritium inventory and effectiveness of tritium removal during bake-out procedures in ITER will be discussed.

Beryllium has been chosen as the first wall material for ITER. In order to understand the issues of material migration and tritium retention associated with the use of beryllium, a largely beryllium first wall will be installed in JET. As part of the JET ITER-like wall, beryllium tiles with marker coatings are proposed as a diagnostic tool for studying the erosion and deposition of beryllium around the vessel. The nominal structure for these coatings is a similar to 10 mu m beryllium surface layer separated from the beryllium tile by a 2-3 mu m metallic inter-layer. Two types of coatings are tested here; one with a nickel inter-layer anti one with a copper/beryllium mixed inter-layer. The coating samples were deposited by DC magnetron Sputtering at General Atomics and were exposed to deuterium plasma in PISCES-B. The results of this testing show that the beryllium/nickel marker coating would be suitable for installation in JET.

A beryllium-seeded deuterium plasma and transient surface heating system is used in PISCES-B to investigate mixed-material erosion and redeposition properties of ITER relevant divertor materials. This heating can be used to investigate the effects of thermal cycling in plasma facing components expected during ELMs in ITER-like devices. An experimental investigation of the effects of thermal cycling on Be films on graphite has been conducted. It has been shown previously that Be film growth on C can form carbide layers that reduce the chemical erosion of C during deuterium ion bombardment. Results from heat cycling on the chemical erosion and on deuterium retention in C targets up to 1200 °C have shown an enhancement in layer formation. In this report, the temperature range was extended to above 2100 °C, the expected dissociation temperature of Be 2 C. It has been found that even heat pulses up 0.1 s long to temperatures above this dissociation temperature, Be 2 C layer formation is enhanced by the thermal cycling. This work was supported by Grant DE-FG02-07ER-54913 from the US DoE.

The characteristic time of protective beryllium layer formation on a graphite target, τBe/C, has been investigated as a function of surface temperature, Ts, ion energy, Ei, ion flux, Γi, and beryllium ion concentration, cBe, in beryllium seeded deuterium plasma. τBe/C is found to be strongly decreased with increasing Ts in the range of 550–970 K. This is thought to be associated with the more efficient formation of beryllium carbide (Be2C). By scanning the parameters, a scaling expression for τBe/C has been derived as τ Be / C [ s ] = 1.0 × 10 - 7 c Be - 1.9 ± 0.1 E i 0.9 ± 0.3 Γ i - 0.6 ± 0.3 exp ( ( 4.8 ± 0.5 ) × 10 3 / T s ) , where cBe is dimensionless, Ei in eV, Γi in 1022 m−2 s−1 and Ts in K. Should this scaling extend to an ITER scenario, carbon erosion of the divertor strike point region may be reduced with characteristic time of ∼6 ms. This is much shorter than the time between predicted ITER type I ELMs (∼1 s), and suggests that protective beryllium layers can be formed in between ELMs, and mitigate carbon erosion.

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

A neutral Be atom source and deposition probe have been installed on PISCES-B to simulate ITER diverter erosion and deposition phenomena where Be impurity concentration in the plasma, up to 10%, is expected due to first wall erosion. Graphite target erosion by deuterium plasma is found to be significantly reduced with as little as ∼0.1% Be impurities in the plasma [K. Schmid, M.J. Baldwin, R. Doerner, these Proceedings]. Deposited Be on the target is re-eroded and leads to deposited layers in the target vicinity, that are sampled with the deposition probe. These layers are Be rich and contain only small C concentrations (

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

A vacuum-compatible-injector device, that allows the extrusion of clean-surface samples of low-melting-point liquid metal into a holder kept in vacuum, is reported. Clean-surface samples of liquid lithium, gallium and 80 at.% tin–20 at.% lithium alloy have been demonstrated for the purpose of their use in plasma/liquid metal interaction experiments. As a result of the use of the injector, subsequent experiments are free of the influences of surface-oxide material. The elimination of surface oxide facilitates wetting and spreading of clean-liquid lithium on a stainless steel substrate holder (AISI 304). Further, the spreading of the liquid is more rapid with increasing holder temperature. At temperatures above 500 °C, evaporative loss of the lithium is substantial under vacuum conditions. However, pressurizing the vacuum system with argon gas to ∼1 atm significantly suppresses evaporation while maintaining chemically-inert conditions that keep the lithium clean, and facilitate a reasonable wetting rate.

Surface modification and helium retention on powder metallurgy tungsten and plasma sprayed tungsten coated CFC exposed by low energy (100 eV) high flux (3.83×10 21 –1.20×10 22 m −2 s −1 ) helium plasma have been investigated. Fluence is about 10 26 He m −2 . It is found that fine surface morphology change occurs by the helium exposure. There is little difference of the surface modification between the powder metallurgy tungsten and the plasma sprayed tungsten. In the case that the surface morphology do not changed, two helium desorption peak appeared. The amount of desorption of helium is larger than that of deuterium from tungsten under the same exposure condition by heating up to 1273 K by TDS. In the case that fine surface morphology change occurs, desorption peak at high temperature newly appears. Surface modification such as blister is not formed on the surface exposed to helium under the same exposure condition which the blister is formed by deuterium exposure.

The injection of carborane (C2B10H12) on the PISCES-B linear plasma device has been used to produce boron containing films on various target species. Film growth rates achieved are extremely high (up to 30 nm/s) compared to those typically found for glow discharges (∼0.01 nm/s). For low-Z target materials (C and Al) the film production is highly efficient, with the boron film growth rate comparable to the incident ion flux and the injection rate of boron atoms. The boron to carbon ratio is 3.0–3.6 for these films. Similarly high growth rates (∼10 nm/s) are obtained with high-Z target (W), but with lower deposition efficiency and higher B/C film ratio. The high film growth rate/efficiency are apparently linked to the high degree of carborane ionization and dissociation caused by the ∼40 eV PISCES-B plasma, compared with T

Plasma sprayed tungsten coated carbon fiber composites (VPS-W/CX-2002U) and powder metallurgy tungsten (PM-W) have been exposed to a high deuterium flux (≃1022 m−2 s−1) with low energy (100 eV) in a range from 708 to 843 K. Surface modification and deuterium retention after the exposure have been investigated to prove the suitability of such materials in fusion devices. Blisters are formed on the PM-W by deuterium irradiation with a fluence of 7.5×1025 m−2. The amount of blisters and their average size increase with an increase of a fluence to 3.00×1026 m−2. On the other hand, no modification is observed on VPS-W/CX-2002U. Desorption of VPS-W/CX-2002U irradiated by deuterium is different from that of PM-W. The peak temperature of D2 and HD release from PM-W is about 703 K. However, the desorption curve of VPS-W/CX-2002U gradually increases with increasing temperature up to 1273 K.

This paper reports on recent studies of high flux deuterium and helium plasma interaction with liquid lithium in the Pisces–B edge plasma simulator facility. Deuterium retention is explored as a function of plasma ion fluence in the range 6×10 19 –4×10 22 atoms cm −2 and exposure temperatures of 523–673 K. The results are consistent with full uptake of the deuterium ions incident on the liquid metal surface, independent of the temperature of the liquid lithium. Full uptake continues until the sample is volumetrically converted to lithium deuteride. Helium retention is not observed for fluences up to 5×10 21 He atoms cm −2 . Measurements of the erosion of lithium are found to be consistent with physical sputtering for the lithium solid phase. However, a mechanism that provides an increased evaporative-like yield and is related to ion impact events on the surface, dominates during the liquid phase leading to an enhanced loss rate for liquid lithium that is greater than the expected loss rate due to evaporation at elevated temperatures. Further, the material loss rate is found to depend linearly on the incident ion flux, even at very high temperature.

Liquid metals such as gallium, tin and lithium are potential plasma-facing materials that may be used to withstand the high heat and particle fluxes in a fusion plasma environment. The interaction of plasma with liquid gallium surfaces has been examined experimentally because of the liquid’s wide temperature range (303–2478 K) and relatively low chemical reactivity. The deuterium retention in liquid gallium samples following plasma exposure in the PISCES experimental plasma device is measured using thermal desorption spectroscopy and is found to be independent of exposure temperature. The retention level saturates at a value of roughly 3 × 10 23 Dm −3 (or about 5 ppm) for sample exposure temperatures ranging from 333 to 800 K and incident ion fluences up to 1.5 × 10 26 Dm −2 . Results from the analyses of the surface after plasma exposure using Auger electron spectroscopy is reported. Micropitting is observed in scanning electron microscope pictures taken after plasma exposure and after resolidification of the gallium surface. Calculations based on the observed retained concentration of deuterium in gallium show that the pumping capabilities of a flowing gallium surface will be small. In addition, measurements of the erosion yield of deuterium-bombarded gallium are presented and compare favourably with results from sputtering yield calculations.

Particle balance measurements have been performed in low-density [Ne(r=0)

PISCES-B is a steady-state linear-plasma simulator facility dedicated to the study of plasma–material interactions. This paper will summarize recent experiments conducted in PISCES in the areas of: plasma–liquid-metal interactions, the flux dependence of the chemical erosion of graphite, and surface modification to tungsten samples exposed to high-fluence plasmas. Enhanced erosion of liquid-metal samples exposed to plasma is observed for both lithium and gallium samples. The surface stratification of liquid-metals is discussed as a possible explanation. The importance of correction terms in the interpretation of the chemical erosion measurements from graphite samples is then emphasized. And finally, surface analysis of tungsten samples, after deuterium plasma bombardment, has revealed microscopic damage in the surface at ion energies below the sputtering threshold energy. This damage appears to be due to pockets of deuterium gas forming within the tungsten and subsequently rupturing.

Visible emission spectroscopy is used to quantify the carbon impurity concentration in the linear plasma divertor simulator of the Plasma Interaction Surface Component Experimental Station (PISCES-B). A technique has been developed to obtain noninvasively the absolute photometric calibration of a visible spectrometer in order to minimize exposure of equipment in the beryllium safety enclosure of PISCES-B. The principal intrinsic source of carbon appears to be chemical sputtering at the vessel walls, with a typical background concentration ∼0.2% with deuterium plasmas. Helium plasmas have a lower carbon contamination ∼0.01%. Methane gas injection is used to increase the carbon contamination in a controlled manner to better simulate tokamak edge plasma conditions. It is found that the spectroscopic method of determining the carbon fraction agrees well with the relative gaseous carbon contamination measured with a residual gas analyzer during plasma operations.

When any plasma confinement device is fabricated from more than a single material which can come into contact with either particle or heat flux, there is the potential for migration of one of these materials to the locations of other materials. This combination of materials, or mixed materials, can have substantially different properties than either of the original materials. The PISCES-B linear plasma device is examining the formation conditions and properties of mixed-material surface layers which can form on plasma-facing components. The PISCES-B device has been modified to incorporate an impurity gas (CD4, CO, O2, etc.) puffing system in the target interaction region. It is, therefore, possible to control the fraction of impurities in the incident plasma and to perform systematic tests on the conditions necessary to form mixed-materials surface layers. The concentration of the species in the plasma column is measured spectroscopically, as well as by a residual gas monitor on the vacuum chamber. Measurements of the rate of growth of the thickness of the mixed material layer are performed. A simple erosion model can adequately describe the growth rate of the mixed-material layer and may allow for growth rate predictions in other plasma environments. It is also important to investigate the role of redeposition of metallic impurities in the formation of mixed material layers. A beryllium evaporator has been independently installed upstream of the target-interaction region to allow seeding of the incident plasma with beryllium. The presence of beryllium on the sample surface is observed to reduce the chemical erosion of the graphite by more than the reduction of the surface carbon concentration. And finally, the hydrogen isotope retention properties of carbon-containing layers on beryllium could have serious implications for tritium accumulation in ITER.

The transport of implanted hydrogen isotopes in the refractory semiconducting ceramic, boron carbide, is investigated using the TMAP4 code. A review of experimental results for the diffusivity and solubility of hydrogen isotopes in boron carbide is presented, which provide Arrhenius expressions for the kinetics of hydrogen isotope transport. These expressions are utilized in the TMAP4 model to provide predictions for the hydrogen isotope implantation and desorption for experiments now in progress at the UCSD PISCES laboratory for the National Ignition Facility (NIF).

Experiments have been conducted in the PISCES-B device to investigate the erosion and deuterium retention characteristics of beryllium. The impurity fractions of both carbon and oxygen in a deuterium plasma have been reduced to 0.2% for carbon and 0.2% or less for oxygen. Under these conditions, the measured sputtering yield of the beryllium samples agrees with beryllium-oxide sputtering. The clean plasma conditions allow for investigation of the surface morphology of the samples under various exposure conditions. During high-temperature exposure the surface develops a porous structure, unlike the smooth surface resulting from low-temperature exposures. The hydrogen isotope retention characteristics of beryllium are measured under conditions which simulate the ITER first wall and baffle plasma interaction regions. The beryllium samples develop a saturated surface layer under these high-flux bombardment conditions. Thermal desorption spectroscopy (TDS) is used to measure the release temperature of the retained deuterium. The TMAP4 code is used to model the deuterium release from the beryllium.

A better understanding of helium transport in the plasma core and edge in enhanced confinement regimes is now emerging from recent experimental studies on DIII-D. Overall, the results are encouraging. Significant helium exhaust ({tau}*{sub He}/{tau}{sub E} {approximately} 11) has been obtained in a diverted, ELMing H-mode plasma simultaneous with a central source of helium. Detailed analysis of the helium profile evolution indicates that the exhaust rate is limited by the exhaust efficiency of the pump ({approximately}5%) and not by the intrinsic helium transport properties of the plasma. Perturbative helium transport studies using gas puffing have shown that D{sub He}/X{sub eff}{approximately}1 in all confinement regimes studied to date (including H-mode and VH-mode). Furthermore, there is no evidence of preferential accumulation of helium in any of these regimes. However, measurements in the core and pumping plenum show a significant dilution of helium as it flows from the plasma core to the pumping plenum. Such dilution could be the limiting factor in the overall removal rate of helium in a reactor system.

Measurements with high spatial resolution in the neighbourhood of the separatrix show edge ion temperatures in the 1-2 keV range, exceeding the electron temperature substantially, in neutral beam heated H mode discharges on DIII-D. Charge exchange recombination (CER) spectroscopy on light ions such as He2+ and C6+ is used to measure ion temperature, ion density and poloidal and toroidal plasma rotation profiles. Near the separatrix the spatial resolution of the measurements is a few millimetres. The scale length for the edge ion temperature gradient is found to be about one third of the poloidal gyroradius, i.e. a factor of 2 to 3 longer than the neoclassical scale length. The electron temperatures at the plasma edge are substantially ( approximately=1 keV) lower than the ion temperatures. Outside the separatrix, T1 is also found to be higher than Te. We investigate the power balance globally and locally at the plasma edge. It is found that the high heat transfer from the ions to the electrons, just inside the separatrix, can be balanced by local ion heating due to neutral beam absorption

Metallic mirrors will be essential components of the optical diagnostic systems in the International Thermonuclear Experimental Reactor (ITER). Reliability of these systems may be affected by mirror reflectivity changes induced by erosion and/or deposition of impurities (carbon, beryllium). The present study aims to assess the effect of beryllium (Be) deposition on the reflectivity of metallic mirrors and to collect data on the optical quality of these layers in terms of morphology, roughness, etc. Mirrors from molybdenum and copper were exposed in the PISCES-B linear plasma device to collect eroded material from graphite and beryllium targets exposed to beryllium-seeded deuterium plasma. After exposure, relative reflectivity of the mirrors was measured and different surface analysis techniques were used to investigate the properties of the deposited layers. Be layers formed in PISCES-B exhibit high levels of porosity which makes the reflectivity of the Be layers much lower than the reflectivity of pure Be. It is found that if Be deposition occurs on ITER first mirrors, the reflectivity of the coated mirrors will strongly depend on the layer morphology, which in turn depends on the deposition conditions. (C) 2007 American Institute of Physics.

The thermionic electron emission current emitted from a laser-produced hot spot on a tungsten target in weakly-ionized deuterium plasma is measured. It is found to be one to two orders of magnitude larger than expected for bipolar space charge limited thermionic emission current assuming an unperturbed background plasma. This difference is attributed to the plasma being modified by ionization of background neutrals by the emitted electrons. This result indicates that the allowable level of emitted thermionic electron current can be significantly enhanced in weakly-ionized plasmas due to the presence of large neutral densities.

Recent experiments in the Current Drive Experiment-Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, B{sub toroidal} = 2 kG, I{sub P} = 100 kA, T{sub e}(0) = 100 eV, n{sub e}(0) {approx} 5 x 10{sup 19} m{sup -3}) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm{sup 2} (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium-limited discharges are consistent with Z{sub effective} < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiationmore » are strongly reduced.« less

Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described.

Research on the Spherical Torus (or Spherical Tokamak) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect-ratio devices, such as the conventional tokamak. The Spherical Tours (ST) experiments are being conducted in various U.S. research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium-size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the U.S., an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high-performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (B), noninductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values bT of up to 35% with the near unity central betaT have been obtained. NSTX will be exploring advanced regimes where bT up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for noninductive sustainment in NSTX is the high beta-poloidal regime, where discharges with a high noninductive fraction ({approx}60% bootstrap current + neutral-beam-injected current drive) were sustained over the resistive skin time. Research on radio-frequency-based heating and current drive utilizing HHFW (High Harmonic Fast Wave) and EBW (Electron Bernstein Wave) is also pursued on NSTX, Pegasus, and CDX-U. For noninductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip {approx} 500 kA. In parallel, start-up using radio-frequency current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

The spectroscopic method is developed to obtain the He+ ion density nHe+ in low electron temperature, Te=5–20eV, plasmas mixed with He. Plasmas were produced in the PISCES-B linear divertor plasma simulator [R. P. Doerner et al., Phys. Scr. T111, 75 (2004)] where the electron densities are ne=(1−15)×1018m−3 and the ionization degree is ∼1–10%. In the method, the He I line intensity IHeI at λ=447.1nm is used, instead of the He II line intensity in the conventional method. The radial confinement time of He+ ions is requisite, and is measured to be at a level of the Bohm confinement time. The He+ ion concentration, nHe+∕ne, is found to be proportional to IHeI, and to weakly depend on ne and Te. Because of the higher ionization energy of He than other species (D2, Ne, and Ar), the measured nHe+∕ne becomes systematically lower than the He gas pressure fraction, and agrees with data from an omegatron mass spectrometer. The omegatron measurement and estimates of the He+ ion loss rates indicate that the influence...

A power switching system has been developed to reverse the voltage polarity on the sample holder of PISCES-B from negative to positive potential in the microsecond time scale. Positive biasing draws electrons from the steady-state plasma through the sample, creating Ohmic heating on the surface. This pulsed biasing is used to replicate the transient heat loads that will be seen on plasma facing components during transient events, such as edge localized modes, in a device such as ITER. Surface temperatures are measured using two pyrometry techniques, a fast two-color system and a slower, more sensitive spectral system.

The interactions of several types of mixed-materials with a bombarding deuterium plasma are described in this paper. The first type of mixed-material surface is a designed, or engineered, surface: a silicon-doped carbon-fiber composite (NS-31). The Si-doped CFC is compared to an identical, but undoped CFC. The net erosion rate, which under these experimental conditions should be dominated by chemical erosion, is reduced by an amount that is about the same as the concentration of the dopant material. Examination of the CFC surface shows that the dopant exists in macroscopic size zones and is not uniformly distributed throughout the CFC. The addition of a more uniformly distributed dopant, in this case beryllium deposited from the plasma on graphite, is shown to reduce the chemical erosion by more than the concentration of the dopant in the surface layer. Finally, the concentration of impurities in the plasma (and therefore the arrival rate of these impurities at the surface) is influential in determining the resultant chemical bonding on the surface. If the arrival rate of carbon at the surface is large, then a surface rich in carbon-carbon bonding develops. If the arrival rate of carbon at the surface is reduced, then the carbon in the surface exhibits preferential carbidic bonding. At low carbon concentration, carbidic bonding is observed in the surface layer regardless of the temperature of the sample during the plasma exposure.

As a result of the vertical elongation of dee-shaped plasmas, natural single-null poloidal-divertor equilibria have been stably obtained in the upper half of the Doublet III vacuum chamber with a plasma current of 320 kA, a major radius of 140 cm, an average minor radius of ~45 cm, and a vertical elongation of 1.3–1.4 in the plasma cross-section. Without employing any particular divertor chamber, this simple divertor reduces the influx of metallic impurities to the main plasma and the re-cycling of charged particles in the periphery of the main plasma. The divertor reduces the radiation loss power and improves the energy confinement time.

Observation of the intensity of the recycling particle flux at the main plasma edge for various limiter and divertor discharges indicates that the gross energy confinement of beam-heated discharges is closely related to the intensity of the edge particle flux. In limiter discharges, the global particle confinement time and the energy confinement time τE show many similarities: 1) linear Ip dependence at Ip < 600 kA, 2) no BT dependence, and 3) deterioration against injection power. Improvement of τE by increasing Ip, for example, is associated with high temperatures at the plasma edge region accompanied by reduced particle recycling. – Divertor discharges with low particle recycling around the main plasma show better energy confinement than limiter discharges at high plasma densities. The improvement of τE is primarily originated in the reduction of heat transport at the main plasma edge region, which is associated with the reduction of recycling particle flux at the main plasma edge. Under certain operation condition, for example, excessive cold-gas puffing, the discharge shows relatively high scrape-off plasma density and strong particle recycling between the main plasma and the limiter. The energy confinement time of these discharges degrades somewhat or reduces completely to that of the limiter discharge. – In low-recycling divertor discharges, the central electron and ion temperature is proportional to the injection power, and the plasma stored energy is proportional to ePabs (scales as INTOR scaling). With ≈ 4 MW beam injection, high-temperature and high-density plasmas were obtained (stored energy up to 280 kJ, Te(0) ≈ Ti(0) ≈ 2.5–3.0 keV at e ≈ (6–7) × 1013 cm−3, τE* ≈ 70 ms).

Stable plasmas with surface elongations of up to 1.8 (aspect ratio 3.4) have been produced in the upper lobe of Doublet III with the use of both passive and active controls. The growth rate of the vertical instability has been measured at various values of elongation by disabling the feedback circuit of the vertical position control power supply. A rigid-shift analysis of growth rates indicates that the passive stabilization effect of the field-shaping coils plays a key role in achieving a high elongation of 1.8. Experiments have demonstrated that the maximum stable elongation is determined by the strength of the passive stabilization effect even with active feedback control. The dee-shape is found to be preferable to an elliptical shape because the triangularity reduces the absolute value of the decay index required to produce a given elongation.

The DIII-D high and low beta stability results have established the basic feasibility of the divertor and H-mode configurations up to elongation 2.0 for next generation tokamaks. The 6.8% beta T achieved has already exceeded projected operating requirements of next generation devices. beta T > 6% has been sustained for 800 ms. Stability calculations and patterns of MHD mode behavior suggest a central expanding zone of ballooning instabilities leads ultimately to unstable m/n=2/1 modes which cause beta collapse or disruption. The pressure gradient at the plasma edge just reaches the first regime ballooning limit prior to ELMs. ECRH has proven effective for generating H-mode, sawtooth suppression, and ELM suppression.

Using a single null divertor configuration, heat flux intensity and its profile on the divertor plates as a function of plasma current and density were measured with an infrared camera and thermocouples. The vertical width of the heat flux on the divertor plates 2λ is ≈ 10 cm at the lower separatrix and is ≈ 5.5 cm at the upper separatrix. A diffusion coefficient D ⊥ which is obtained from the measurement of the diffusion length across the scrape-off field lines is roughly proportional to 1 B t and its magnitude is on the order of Bohm diffusion. The heat flux on the plates decreases by more than a factor of 5 with increasing electron density in the main plasma and is much smaller than that on the limiters in non-diverted plasmas. Only 3% of ohmic input power goes into the divertor plates at high density of the main plasma, while ≈ 20% goes in at low density. The decrease of heat flux is in good agreement with the increase of radiation loss in the divertor region. The heat flux on the divertor plates can be reduced by remote radiative cooling in high density discharges.

Reduced recycling, reduced edge neutral pressure, improved density control, and improved discharge reproducibility have been achieved in the DIII-D tokamak by in situ helium condition of the graphite tiles. An improvement in energy confinement has been observed in hydrogen discharges with hydrogen beam injection after helium preconditioning. After the graphite wall coverage in DIII-D was increased to 40%, helium glow wall conditioning, routinely applied before each tokamak discharge, has been necessary to reduce recycling and obtain H-mode. The utilization of helium glow wall conditioning was an important factor in the achievement of an ohmic H-mode, i.e. no auxiliary heating, with significant improvement in ohmic energy confinement.

Neutral-beam-heated expanded-boundary (XB) divertor discharges have been obtained in Doublet III with high heating efficiency for wide ranges of plasma parameters (Ip: 300–800 kA; BT: 8–24kG; e :(2–10) × 1013cm−3; Pb 3 MW), a mild deterioration of the energy confinement time has been observed.