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 (ne/nGW), 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.
We report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off ...layer of the tungsten upper divertor successfully eliminated ELMs for 3-5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor Dα baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previous ELM elimination results via Li injection into the lower carbon divertor in EAST (Hu et al 2015 Phys. Rev. Lett. 114 055001). These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs (Lang et al 2017 Nucl. Fusion 57 016030), highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.
We report observation of a new high performance regime in discharges in the National Spherical Torus Experiment, where the H mode edge "pedestal" temperature doubles and the energy confinement ...increases by 50%. The spontaneous transition is triggered by a large edge-localized mode, either natural or externally triggered by 3D fields. The transport barrier grows inward from the edge, with a doubling of both the pedestal pressure width and the spatial extent of steep radial electric field shear. The dynamics suggest that 3D fields could be applied to reduce edge transport in fusion devices.
Observations of improved radio frequency (rf) heating efficiency in ITER relevant high-confinement (H-)mode plasmas on the National Spherical Tokamak Experiment are investigated by whole-device ...linear simulation. The steady-state rf electric field is calculated for various antenna spectra and the results examined for characteristics that correlate with observations of improved or reduced rf heating efficiency. We find that launching toroidal wave numbers that give fast-wave propagation in the scrape-off plasma excites large amplitude (∼kV m(-1)) coaxial standing modes between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggests that these modes are a probable cause of degraded heating efficiency.
Toroidally non-axisymmetric divertor profiles during the 3-D field application and for ELMs are studied with simultaneous observation by a new wide angle visible camera and a high speed IR camera. A ...newly implemented 3-D heat conduction code, TACO, is used to obtain divertor heat flux. The wide angle camera data confirmed the previously reported result on the validity of vacuum field line tracing on the prediction of split strike point pattern by 3-D fields as well as the phase locking of ELM heat flux to the 3-D fields. TACO calculates the 2-D heat flux distribution allowing assessment of toroidal asymmetry of peak heat flux and heat flux width. The degree of asymmetry (εDA) is defined to quantify the asymmetric heat deposition on the divertor surface and is found to have a strong positive dependence on peak heat flux.
Differences in the electron particle and thermal transport are reported between plasmas produced in a quasihelically symmetric (QHS) magnetic field and a configuration with the symmetry broken. The ...thermal diffusivity is reduced in the QHS configuration, resulting in higher electron temperatures than in the nonsymmetric configuration for a fixed power input. The density profile in QHS plasmas is centrally peaked, and in the nonsymmetric configuration the core density profile is hollow. The hollow profile is due to neoclassical thermodiffusion, which is reduced in the QHS configuration.
A fusion development facility (FDF) based on the tokamak approach with normal conducting magnetic field coils is presented. FDF is envisioned as a facility with the dual objective of carrying forward ...advanced tokamak (AT) physics and enabling the development of fusion energy applications. AT physics enables the design of a compact steady-state machine of moderate gain that can provide the neutron fluence required for FDF's nuclear science development objective. A compact device offers a uniquely viable path for research and development in closing the fusion fuel cycle because of the demand to consume only a moderate quantity of the limited supply of tritium fuel before the technology is in hand for breeding tritium.
We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor ...target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR∝RTAR2L||6/7 and TTAR∝RTAR−2L||−4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.
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