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Experiments performed in DIII-D demonstrate that higher plasma current and heating power combined with impurity seeding facilitate the achievement of divertor detachment with a higher pedestal pressure and higher plasma performance in H-mode plasmas with a baffled closed divertor compared with an open divertor. Dedicated experiments were carried out to study the impact of power, plasma current and impurity seeding on divertor detachment with ion B × ∇ B directed into the divertor favorable for the L–H transition. With a factor of three variation in heating power and with only D2 puffing, no significant difference in the separatrix density at detachment onset was found. The higher heating power leads to higher impurity concentration and wider scrape-off layer (SOL) width, and reduces the detachment onset density to one similar to that in lower-power plasmas. Higher current requires higher pedestal and line-averaged densities to achieve divertor detachment; however, the increase in separatrix density at increasing plasma current is found to be less pronounced. Initial calculations found that both power scan and plasma current scan datasets are qualitatively consistent with theory after considering the change in impurity concentration and heat flux width. This also motivates the future extensive study of transport and divertor impurity behavior in order to have a quantitative comparison between experiment and theory. Compared with an open divertor, a closed divertor facilitates detachment onset at ∼40% lower line-averaged plasma density. Additional N2 seeding facilitates the achievement of detachment at a lower separatrix density and thus a higher pedestal temperature, which is beneficial for advanced tokamak scenarios. Higher heating power requires a higher N2 puffing rate to achieve the same degree of detachment, while a higher N2 puffing rate leads to lower detachment onset line-averaged density, both of which agree with theory. In contrast to the narrower pedestal in an open divertor approaching detachment, the pedestal density width in a closed divertor increases with density. The density gradient increases with line-averaged density at higher plasma current, but remains nearly unchanged at lower plasma current. In particular, compared with discharges with low power, at high heating power the pedestal density gradient is much weaker, while the SOL density is significantly higher and wider. At the same plasma current, both pedestal pressure gradient and temperature gradient decrease linearly with the line-averaged density but remain similar across different heating powers. Even with different plasma current and heating power, the normalized pressure gradient remains identical. As a result, achievement of divertor detachment with a higher pedestal pressure and higher plasma performance is shown in a closed divertor, which is important for improving core–edge integration as one of the critical issues for future tokamak fusion reactors. [ABSTRACT FROM AUTHOR]

Abstract: The Quiescent H-mode exhibits H-mode levels of confinement and edge pedestal pressures, but does not exhibit ELMs. To date this mode has only been observed in tokamaks during beam heating with some or all of the beams injected counter to the direction of plasma current. During QH-mode, fast ion loss to the low field side plasma facing surfaces has been observed. Some of the fast ion loss is calculated to be the result of outwardly directed banana orbits of the energetic beam ions created in the edge region. Other fast ion loss has been observed to be associated with bursts or oscillations in broadband, high-frequency, magnetic fluctuations. The relationship of the fast ion loss to the ELM stabilization or edge particle transport during QH-mode is not yet understood. [Copyright &y& Elsevier]

Abstract: Experiments were conducted to determine the influence of displacement damage on retention of deuterium in tungsten plasma-facing components in a tokamak. Tungsten samples, previously damaged by ion irradiation, were exposed to the outer strike point of attached H-mode plasmas in DIII-D. Nuclear reaction analysis (NRA) was used to measure the depth profile of deuterium retained in the tungsten. Displacement damage increased the concentration of retained deuterium to the maximum depth (about 2.5μm) of the damage, to concentrations up to 0.003D/W, compared to D/W<10−5 in undamaged W. Tungsten coverage on adjacent carbon surfaces of the probe was mapped by Rutherford backscattering, giving the average tungsten erosion rate and spatial variation of redeposition. [Copyright &y& Elsevier]

Abstract: Use of carbon in tokamaks leads to a serious tritium retention issue due to co-deposition. To further investigate the processes involved, a detached ELMy H-mode (6.5MW NBI) experiment was performed on DIII-D in which 13CH4 was puffed into the main vessel through the toroidaly symmetric pumping plenum at the top of lower single-null discharges. Subsequently, the 13C content of tiles taken from the vessel wall was measured. The interpretive OEDGE code was used to model the results. The 13C deposition pattern could be reproduced, in general shape and magnitude, by assuming in the code the existence of a parallel flow and a radial pinch in the scrape-off layer. Parallel flows of Mach ∼0.3 toward the inner divertor and a radial pinch ∼10–20m/s (+R-direction) were found to yield 13C deposition comparable to the experiment. [Copyright &y& Elsevier]

Abstract: High temporal and spatial resolution measurements in the boundary of the DIII-D tokamak show that edge localized modes (ELMs) are composed of fast bursts of hot, dense plasma lasting ∼25μs each that travel radially starting at the separatrix at ∼450m/s and rotate in the scrape off layer (SOL), convecting particles and energy to the SOL and walls. The temperature and density in the ELM plasma initially correspond to those at the top of the density pedestal but decay with radius in the SOL. The temperature decay length (∼1.2–1.5cm) is much shorter than the density decay length (∼3–8cm), which in turn decreases with increasing pedestal density. The local particle and energy flux at the wall during the bursts are 10–50% (∼1–2×1021 m−2 s−1) and 1–2% (∼20–30kW/m2) respectively of the LCFS average fluxes, thus particles are transported radially much more efficiently that heat. [Copyright &y& Elsevier]

Abstract: Use of carbon in tokamaks leads to a major tritium retention issue due to co-deposition. To investigate this process a low power L-mode experiment was performed on DIII-D in which 13CH4 was puffed into the main vessel through the toroidally-symmetric pumping plenum at the top of lower single-null discharges. Subsequently, the 13C content of tiles taken from the vessel wall was measured. The interpretive OEDGE code was used to model the results. It was found that the 13C deposition pattern is controlled by: (a) source strength of 13C+, (b) Δr s, radial location of the 13C+ source, (c) D ⊥, (d) M ∥, the scrape-off layer parallel Mach number. Best agreement was found for (a) ∼50% conversion efficiency 13CH4 → 13C+, (b) Δr s, ∼+3.5cm (outboard of separatrix) near 13CH4 injection location, (c) D ⊥ ∼0.3m2/s, (d) M ∥ ∼0.4 toward inside. [Copyright &y& Elsevier]

Abstract: We report on recent DIII-D experiments that integrate edge localized mode (ELM) suppression using resonant magnetic perturbations (RMPs) with divertor heat flux reduction under radiating divertor conditions. Our results illustrate the limitations in maintaining ELM suppression at gas puffing levels that were representative of good puff and pump operation in previous experiments without RMP. The electron pressure gradient in the pedestal (∇Pe ) increased steadily during gas puffing and ELMs returned once ∇Pe reached values consistent with the peeling–ballooning stability limit, as determined by edge stability analysis. Even with this return of ELMs, a radiating divertor with RMP generated higher levels of total radiated power (∼40%) than comparable standard ELMing discharges without RMP at the same density. Differences in the accumulation of the seed argon in the core plasma between RMP and non-RMP during puff and pump were less than 20%. [Copyright &y& Elsevier]

Abstract: Measurements and simulations with the UEDGE code of radiated power, and ion saturation currents and power loads to the target plates have been compared for density scans in ohmic and low confinement mode plasmas in DIII-D, ASDEX Upgrade, and JET. Simulations including cross-field drifts and assuming elevated chemical sputtering yields of 3%–4% move the numerical solutions closer to many of the measurements compared to omitting the drifts and using the published Davis–Haasz yields. Adopting these assumptions the simulations reproduce the measured currents and powers, and their functional dependence on upstream density to within a factor of 2, with the exception of the ion currents to the low field side target in ASDEX Upgrade and the high field side target in JET. The applicability of using enhanced sputtering yields is discussed by comparing measured and simulated emission from low charge state carbon in the divertor regions. [Copyright &y& Elsevier]

Abstract: Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. was injected into the secondary outer divertor to simulate sputtering there. The majority of the found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240°C, is incomplete. The principal features of the measured deposition pattern have been replicated by the OEDGE interpretive code. [Copyright &y& Elsevier]

Abstract: An experiment aimed at benchmarking the ITER scrape-off layer (SOL) power width scaling in limited L-mode discharges has been conducted on DIII-D. Scans of the main scaling parameters were performed in an inner-wall-limited (IWL) magnetic configuration. Using the near-SOL density and temperature e-folding lengths, , , determined from reciprocating Langmuir probe measurements, SOL power flux density e-folding lengths, , are derived. A few lower single null (LSN) discharges were also run for comparison. The results are generally in agreement with the ITER design assumptions, finding that and are correlated and both and are on average 2.1–2.5 times larger in IWL configurations than in LSN. In moderate elongation (κ ∼1.4) IWL discharges, is largest and agrees with the assumed ITER scaling within the estimated uncertainty (a factor of ∼2). In IWL discharges measurements are consistent with the expectations of SOL power balance. [Copyright &y& Elsevier]

Abstract: The Porous Plug Injector (PPI) has proven to be an invaluable diagnostic for in situ characterization and quantification of erosion phenomena in DIII-D. Previous work has led to derivation of three primary figures of merit for chemical erosion (CE) in attached and cold divertor conditions: relative intensity of C+ chemical and physical sources, the CE yield (Y chem), and effective photon efficiencies for chemically eroded products. Application of these figures for accounting of observed absolutely calibrated CI and CII emission intensities is demonstrated to produce a self-consistent solution at the DIII-D targets. Reinterpretation of the CI (C0) spectral lineshape profile supports the relative roles of local chemical versus physical sputtering as previously determined for CII (C+). Comparison of calculated in situ Y chem to that measured ex situ suggests a tokamak-specific lower energy threshold for CE and has potentially major implications for prediction of tritium co-deposition near the divertor targets in ITER. [Copyright &y& Elsevier]