The pumping performance of the EAST tokamak upgraded divertor has been optimized by varying the height and width of the duct that connects the plasma and plenum volumes. The optimization is enacted ...using a semi-analytic model which predicts the molecular pressure in the plenum given the plasma facing component geometry and plasma parameters along the divertor, specifically electron temperature, electron density, and the ion particle flux. The model is semi-analytic as the plasma parameters can come from experimental data (or plasma transport simulations), and uses a first-flight approximation for neutral transport. This model is computationally inexpensive, allowing for several geometries and sets of plasma parameters to be rapidly evaluated. A duct geometry is discovered that performs well for different plasma equilibria and strike point positions. The semi-analytic model results are compared to two dimensional plasma and neutral transport simulations using the SOLPS-ITER code, suggesting that charge-exchange is a significant contributor to the plenum pressure at low electron temperature. An extension to the semi-analytic model improves the comparison over a broader range of temperature. The conservative prediction of the model and the broad maximum in the pressure contours imply that the inclusion of charge-exchange is not required in the optimization procedure, but that quantitative estimates should be provided with a high-fidelity model such as SOLPS-ITER.
New evidence indicates that there is significant 3D variation in density fluctuations near the boundary of weakly 3D tokamak plasmas when resonant magnetic perturbations are applied to suppress ...transient edge instabilities. The increase in fluctuations is concomitant with an increase in the measured density gradient, suggesting that this toroidally localized gradient increase could be a mechanism for turbulence destabilization in localized flux tubes. Two-fluid magnetohydrodynamic simulations find that, although changes to the magnetic field topology are small, there is a significant 3D variation of the density gradient within the flux surfaces that is extended along field lines. This modeling agrees qualitatively with the measurements. The observed gradient and fluctuation asymmetries are proposed as a mechanism by which global profile gradients in the pedestal could be relaxed due to a local change in the 3D equilibrium. These processes may play an important role in pedestal and scrape-off layer transport in ITER and other future tokamak devices with small applied 3D fields.
The pedestal structure in NSTX is strongly affected by lithium coatings applied to the PFCs. In discharges with lithium, the density pedestal widens, and the electron temperature (Te) gradient ...increases inside a radius of ψN ∼ 0.95, but is unchanged for ψN > 0.95. The inferred effective electron thermal and particle profiles reflect the profile changes: is slightly increased in the near-separatrix region, and is reduced in the region ψN < 0.95 in the with-lithium case. The profile shows a broadening of the region with low diffusivity with lithium, while the minimum value within the steep-gradient region is comparable in the two cases. The linear microstability properties of the edge plasma without and with lithium have been analysed. At the pedestal top microtearing modes are unstable without lithium. These are stabilized by the stronger density gradient with lithium, becoming TEM-like with growth rates reduced and comparable to E × B shearing rates. In the region ψN > 0.95, both the pre- and with-lithium cases are calculated to be unstable to ETG modes, with higher growth rates with lithium. Both cases are also found to lie near the onset for kinetic ballooning modes, but in the second-stable region where growth rates decrease with increasing pressure gradient.
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.
Results and interpretation of recent experiments on DIII-D designed to evaluate divertor geometries favourable for radiative heat dispersal are presented. Two approaches examined here involved ...lengthening the parallel connection in the scrape-off layer, L|, and increasing the radius of the outer divertor separatrix strike point, ROSP, with the goal of reducing target temperature, TTAR, and increasing target density, nTAR. From one-dimensional (1D) two-point modelling based on conducted parallel heat flux, it is expected that: and , where nSEP is the midplane separatrix density. These scalings suggest that conditions conducive to a radiative divertor solution can be achieved at low nSEP by increasing either ROSP or L|. Our data are consistent with the above L| scalings. On the other hand, the observed dependence of nTAR and TTAR on ROSP displayed a more complex behaviour, under certain conditions deviating from the above scalings. Our analysis indicates that deviations from the ROSP scaling were due to the presence of convected heat flux, driven by escaping neutrals, in the more open configurations of the larger ROSP cases. A comparison of 'open' versus 'closed' divertor configurations for the H-mode plasmas in this study show that the 'closed' case provides at least 30% reduction in the peaked heat flux at common density with the 'open' case and partial divertor detachment at lower plasma density.
Calculations of the plasma response to applied non-axisymmetric fields in several DIII-D discharges show that predicted displacements depend strongly on the edge current density. This result is found ...using both a linear two-fluid-MHD model (M3D-C1) and a nonlinear ideal-MHD model (VMEC). Furthermore, it is observed that the probability of a discharge being edge localized mode (ELM)-suppressed is most closely related to the edge current density, as opposed to the pressure gradient. It is found that discharges with a stronger kink response are closer to the peeling-ballooning stability limit in ELITE simulations and eventually cross into the unstable region, causing ELMs to reappear. Thus for effective ELM suppression, the RMP has to prevent the plasma from generating a large kink response, associated with ELM instability. Experimental observations are in agreement with the finding; discharges which have a strong kink response in the MHD simulations show ELMs or ELM mitigation during the RMP phase of the experiment, while discharges with a small kink response in the MHD simulations are fully ELM suppressed in the experiment by the applied resonant magnetic perturbation. The results are cross-checked against modeled 3D ideal MHD equilibria using the VMEC code. The procedure of constructing optimal 3D equilibria for diverted H-mode discharges using VMEC is presented. Kink displacements in VMEC are found to scale with the edge current density, similar to M3D-C1, but the displacements are smaller. A direct correlation in the flux surface displacements to the bootstrap current is shown.
•A FNSF is needed to reduce the knowledge gaps to a fusion DEMO and accelerate progress toward fusion energy.•FNSF will test and qualify first-wall/blanket components and materials in a DEMO-relevant ...fusion environment.•The Advanced Tokamak approach enables reduced size and risks, and is on a direct path to an attractive target power plant.•Near term research focus on specific tasks can enable starting FNSF construction within the next ten years.
An accelerated fusion energy development program, a “fast-track” approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation, and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a=2.7m/0.77m, κ=2.3, BT=5.4T, IP=6.6 MA, βN=2.75, Pfus=127MW. The modest bootstrap fraction of ƒBS=0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q∼10 in ITER.
Developing a reactor-compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved ...H-mode confinement, H-mode power threshold reduction, and other plasma performance benefits. During the 2010 NSTX campaign, application of a relatively modest amount of Li (300 mg prior to the discharge) resulted in a ∼50% reduction in heat load on the liquid lithium divertor (LLD) attributable to enhanced divertor bolometric radiation. These promising Li results in NSTX and related modelling calculations motivated the radiative LLD concept proposed here. Li is evaporated from the liquid lithium (LL) coated divertor strike-point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating the divertor heat removal. The LL coating of divertor surfaces can also provide a 'sacrificial' protective layer to protect the substrate solid material from transient high heat flux such as the ones caused by the edge localized modes. By operating at lower temperature than the first wall, the LL covered large divertor chamber wall surfaces can serve as an effective particle pump for the entire reactor chamber, as impurities generally migrate towards lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity (e.g., ∼1 l s−1 for ∼1% level 'impurities') is envisioned for a steady-state 1 GW-electric class fusion power plant.
Abstract
In this paper, we present linear and nonlinear gyrokinetic analyses in the pedestal region of two DIII-D ELMy H-mode discharges using the CGYRO code. The otherwise matched discharges employ ...different divertor configurations to investigate the impact of varying recycling and particle source on pedestal profiles. Linear gyrokinetic simulations find electrostatic ion-scale instabilities (ion temperature gradient and trapped electron modes, ITG–TEM) are present just inside the top of the pedestal with growth rates that are enhanced significantly by parallel velocity shear. In the sharp gradient region,
E
×
B
shearing rates are comparable or larger than ion scale growth rates, suggesting the suppression of ITG–TEM modes in this region. Instead, the electron temperature profiles are found to be correlated with and just above the electron temperature gradient (ETG) instability thresholds. Using gradients varied within experimental uncertainties, nonlinear electron-scale gyrokinetic simulations predict electron heat fluxes from ETG turbulence, that when added to neoclassical (NC) ion thermal transport simulated by NEO, account for 30%–60% of the total experimental heat flux. In addition, the NC electron particle flux is found to contribute significantly to the experimental fluxes inferred from SOLPS-ITER analysis. Additional nonlinear gyrokinetic simulations are run varying input gradients to develop a threshold-based reduced model for ETG transport, finding a relatively simple dependence on
η
e
=
L
ne
/
L
Te
. Predictive transport simulations are used to validate this pedestal-specific ETG model, in conjunction with a model for NC particle transport. In both discharges, the predicted electron temperatures are always overpredicted, indicative of the insufficient stiffness in the ETG pedestal model to account for all of the experimental electron thermal transport. In the case of the closed divertor discharge with lower particle source, the predicted electron density is close to the experiment, consistent with the magnitude of NC particle transport in that discharge. However, the density profiles are overpredicted in the open divertor discharge (larger particle source), due to insufficient model transport. The implications for other mechanisms accounting for the remainder of transport in the sharp gradient region in the two discharges are discussed.
Owing to its high magnetic field, high power, and compact size, the SPARC experiment will operate with divertor conditions at or above those expected in reactor-class tokamaks. Power exhaust at this ...scale remains one of the key challenges for practical fusion energy. Based on empirical scalings, the peak unmitigated divertor parallel heat flux is projected to be greater than 10 GW m−2. This is nearly an order of magnitude higher than has been demonstrated to date. Furthermore, the divertor parallel Edge-Localized Mode (ELM) energy fluence projections (~11–34 MJ m−2) are comparable with those for ITER. However, the relatively short pulse length (~25 s pulse, with a ~10 s flat top) provides the opportunity to consider mitigation schemes unsuited to long-pulse devices including ITER and reactors. The baseline scenario for SPARC employs a ~1 Hz strike point sweep to spread the heat flux over a large divertor target surface area to keep tile surface temperatures within tolerable levels without the use of active divertor cooling systems. In addition, SPARC operation presents a unique opportunity to study divertor heat exhaust mitigation at reactor-level plasma densities and power fluxes. Not only will SPARC test the limits of current experimental scalings and serve for benchmarking theoretical models in reactor regimes, it is also being designed to enable the assessment of long-legged and X-point target advanced divertor magnetic configurations. Experimental results from SPARC will be crucial to reducing risk for a fusion pilot plant divertor design.
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