| 1. |
- Fenstermacher, M.E., et al.
(författare)
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DIII-D research advancing the physics basis for optimizing the tokamak approach to fusion energy
- 2022
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 62:4
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Tidskriftsartikel (refereegranskat)abstract
- DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter-I p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high-Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L-H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ∼8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co-I p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.
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| 2. |
- Rudakov, D. L., et al.
(författare)
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Dust measurements in tokamaks (invited)
- 2008
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Ingår i: Review of Scientific Instruments. - : AIP Publishing. - 0034-6748 .- 1089-7623. ; 79:10, s. 10F303-
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Tidskriftsartikel (refereegranskat)abstract
- Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 mu m in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C(2) dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.
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| 3. |
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| 4. |
- Bykov, I., et al.
(författare)
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Modification of adhered dust on plasma-facing surfaces due to exposure to ELMy H-mode plasma in DIII-D
- 2017
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Ingår i: NUCLEAR MATERIALS AND ENERGY. - : Elsevier BV. - 2352-1791. ; 12, s. 379-385
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Tidskriftsartikel (refereegranskat)abstract
- Transient heat load tests have been conducted in the lower divertor of DIII-D using DiMES manipulator in order to study the behavior of dust on tungsten Plasma Facing Components (PFCs) during ELMy H-mode discharges. Samples with pre- adhered, pre- characterized dust have been exposed at the outer strike point (OSP) in a series of discharges with varied intra-(inter-) ELM heat fluxes. We used C dust because of its high sublimation temperature and non-metal properties. Al dust as a surrogate for Be and W dust were employed as relevant to that in the ITER divertor. The poor initial thermal contact between the substrate and the particles led to overheating, sublimation and shrinking of the carbon dust, and wetting induced coagulation of Al dust. Little modification of the W dust was observed. An enhanced surface adhesion and improvement of the thermal contact of C and Al dust were the result of exposure. A post mortem "adhesive tape" sampling showed that 70% of Al, <5% of W and C particles could not be removed from the surface owing to the improved adhesion. Al and C but not W particles that could be lifted had W inclusions indicating damage to the substrate. This suggests that non destructive methods may be inefficient for removal of dust in ITER.
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| 5. |
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