Divertor heat and particle control experiments on the DIII-D tokamak

Mahdavi, M. A.; Allen, Steven L.; Baker, D. R.; Bastasz, B.; Brooks, N.H.; Buchenauer, D.; Campbell, R. B.; Cuthbertson, J. W.; Evans, T.E.; Fenstermacher, M.E.; Finkenthal, D. F.; Foote, J. H.; Hill, D. N.; Hillis, D.L.; Hinton, F. L.; Hogan, J.; Howald, A. M.; Hyatt, A.W.; Jackson, G.L.; Jong, R.A.; Konoshima, S.; Lasnier, C.J.; Leonard, A.W.; Lippmann, S.; Maingi, R.; Menon, M. M.; Mioduszewski, P.K.; Moyer, R. A.; Ogawa, Hironori; Pétrie, T.W.; Porter, G. D.; Rensink, M.E.; Rognlein, T.; Schaffer, M. J.; Schaubel, K.M.; Sevier, D.L.; Smith, John P.; Staebler, G. M.; Sager, G.T.; Stambaugh, R.D.; Thomas, D. M.; Wade, M. R.; Watkins, J. G.; Weschenfelder, F.; West, W.P.; Whyte, D.G.; Winter, J.; Wong, C.P.C.; Wood, R. D.

doi:10.1016/0022-3115(94)00443-9

Cited by 37 publications

(4 citation statements)

References 32 publications

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“…5.2, have been successfully applied to study SOL and divertor plasma experiments, and in many cases are reasonably successful in describing the observed phenomena. The results are in satisfactory agreement with observations of divertor detachment [144]. In these models, it is seen that carbon impurities, sputtered physically and chemically from the graphite divertor strike-plates, radiate as part of an uncontrolled but self-regulating loop.…”

Section: Power Dispersal and Removalsupporting

confidence: 82%

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

et al. 2001

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The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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Section: Power Dispersal and Removalsupporting

confidence: 82%

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

et al. 2001

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“…The core density control experiments with active cryopumping are discussed in detail elsewhere [3]; briefly, these studies showed effective control of the core electron density. Efficient pumping, as expected, was obtained when the plasma separatrix was placed close to the pump entrance; the pumping rate decreased rapidly when the separatrix was moved away from the pump.…”

Section: Summary Ofrecent Experimental Results Usedtodesign Thediii-dmentioning

confidence: 99%

“…[2], for an overview of divertor results see Ref. [3], and a detailed overview of gas puffing results is presented in Ref. [4].…”

Section: Introductionmentioning

confidence: 99%

Development of a radiative divertor for DIII-D

Allen

Brooks

Campbell

et al. 1995

Journal of Nuclear Materials

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We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized (~10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while 'I;Eremains-2 times ITER-89P scaling. However, ne increases with D2 puffing, and Zeff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (_5~0.8) is important for high 'I;E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented.

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“…Besides these favorable pressure buildup measurements, tokamaks have added active pumps during the EDA. In ELMing H-mode it has proven possible to exhaust the fuel at a sufficient rate to lower the main plasma density a factor of 2 from the normal H-mode level to the L-mode level [30]. It appears a sound physics basis exists for the necessary fuel and helium exhaust in ITER.…”

Section: Particle Controlmentioning

confidence: 99%