Effect of lithium PFC coatings on NSTX density control

Kugel, H.; Bell, M.G.; Bell, R.E.; Bush, C. E.; Gates, D.; Gray, T.; Kaita, R.; LeBlanc, B.P.; Maingi, R.; Majeski, R.; Mansfield, D. K.; Mueller, D.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Sabbagh, S.A.; Skinner, C.H.; Soukhanovskii, V.; Stevenson, T. N.; Zakharov, L.

doi:10.1016/j.jnucmat.2007.01.220

Cited by 56 publications

(21 citation statements)

References 8 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The characteristics of a typical H-mode plasma in NSTX, with an I p of 1 MA and central n e and T e of 4-6 × 10 13 cm −3 and 0.9 keV, respectively, are shown in figure 1. For these H-mode discharges, the measured B-X-O coupling efficiencies were 20-60% for f ce , 2f ce and 3f ce (18)(19)(20)(21)(22)(23)(24)(25)(26)(27)(28)(29)(30)(31)(32)(33)(34)(35)(36) emissions during the L-H transition. However, during the H-mode phase the measured T rad was less than 10 eV for all measured harmonics.…”

Section: Low H-mode Ebe Resultsmentioning

confidence: 99%

See 1 more Smart Citation

A novel high-density embedded AND-type split gate flash memory

Shen

Wang

Pan

et al. 2014

Jpn. J. Appl. Phys.

View full text Add to dashboard Cite

High-β spherical tokamak (ST) plasma conditions cut off propagation of electron cyclotron (EC) waves used for heating and current drive in conventional aspect ratio tokamaks. The electron Bernstein wave (EBW) has no density cutoff and is strongly absorbed and emitted at the EC harmonics, allowing EBWs to be used for heating and current drive in STs. However, this application requires efficient EBW coupling in the high-β, H-mode ST plasma regime. EBW emission (EBE) diagnostics and modelling have been employed on the National Spherical Torus Experiment (NSTX) to study oblique EBW to O-mode (B-X-O) coupling and propagation in H-mode plasmas. Efficient EBW coupling was measured before the L-H transition, but rapidly decayed thereafter. EBE simulations show that EBW collisional damping prior to mode conversion (MC) in the plasma scrape off reduces the coupling efficiency during the H-mode phase when the electron temperature is less than 30 eV inside the MC layer. Lithium evaporation during H-mode plasmas was successfully used to reduce this EBW collisional damping by reducing the electron density and increase the electron temperature in the plasma scrape off. Lithium conditioning increased the measured B-X-O coupling efficiency from less than 10% to 60%, consistent with EBE simulations.

show abstract

Section: Low H-mode Ebe Resultsmentioning

confidence: 99%

“…Edge conditioning using lithium (Li) coatings on the plasma facing components has been achieved with a Li evaporation oven (LITER) [20] installed on an upper vacuum vessel port in NSTX. LITER evaporates Li onto the graphite tiles of the centre stack and lower divertor before and during a discharge.…”

Section: H-mode Ebe Results With Lithium Conditioningmentioning

confidence: 99%

A novel high-density embedded AND-type split gate flash memory

Shen

Wang

Pan

et al. 2014

Jpn. J. Appl. Phys.

View full text Add to dashboard Cite

show abstract

“…To reach this aim, NSTX has been installed with an evaporative lithium system (LiThium EvaporatoR, or LiTER) to coat the graphite tiles that cover the inner walls [2]. This led to 50% reductions in L-mode density and 15% reductions in H-mode [3].…”

Section: Nstx Strike Point Control For Lld a Lld Installation At Nstxmentioning

confidence: 99%

Strike point control for the National Spherical Torus Experiment (NSTX)

et al. 2010

Self Cite

View full text Add to dashboard Cite

This paper presents the first control algorithm for the inner- and outer-strike point position for a Spherical Torus (ST) fusion experiment and the performance analysis of the controller. A liquid lithium divertor (LLD) will be installed on NSTX which is believed to provide better pumping than lithium coatings on carbon PFCs. The shape of the plasma dictates the pumping rate of the lithium by channelling the plasma to LLD, where the strike point location is the most important shape parameter. Simulations show that the density reduction depends on the proximity of the strike point to LLD. Experiments were performed to study the dynamics of the strike point, design a new controller to change the location of the strike point to the desired location and stabilize it. The most effective poloidal field (PF) coils in changing inner- and outer-strike points were identified using equilibrium code. The PF coil inputs were changed in a step fashion between various set points and the step response of the strike point position was obtained. From the analysis of the step responses, proportional–integral–derivative controllers for the strike points were obtained and the controller was tuned experimentally for better performance. The strike controller was extended to include the outer-strike point on the inner plate to accommodate the desired low outer-strike points for the experiment with the aim of achieving ‘snowflake’ divertor configuration in NSTX.

show abstract

“…This feature results in near zero recycling of the plasma on the lithium surface when practically no hydrogen comes back from the surface into the plasma (assuming lithium is not saturated with hydrogen). 12 Such conditions can result in a crucial increase of plasma temperature near the wall, which in turn can improve the core confinement by reducing the plasma turbulence and a) Electronic address: rajiv@ipr.res.in associated transport caused by temperature gradient. The experiments on TFTR, T-11M, and CDX-U tokamaks with partial coverage of the first wall with lithium have shown reduction of plasma recycling.…”

Section: Introductionmentioning

confidence: 99%

Influence of plasma surface interactions on tokamak startup

Goswami¹

2013

Physics of Plasmas

View full text Add to dashboard Cite

The startup phase of a tokamak is a complex phenomenon involving burnthrough of the low-Z impurities and rampup of I p , the plasma current. The design considerations of a tokamak are closely connected with the startup modeling. Plasma evolution is analysed using a zero-dimensional model. The particle and energy balance is considered of two subclasses of plasmas which are penetrable by neutral gas, together with another component, neutrals trapped in the wall. The first subclass includes plasmas being penetrated by slow neutrals of ($few eV) temperature. The second includes plasmas being penetrated only by fast neutrals having a temperature comparable to that of the ions. The impact of impurities on energy balance is considered through their generation by ion induced desorption of adsorbed oxygen on the first wall and physical and chemical sputtering of carbon. The paper demonstrates self-consistently that the evolution of initial phase of the discharge is intimately linked to the condition of the plasma facing components (PFCs) and the resultant plasma surface interactions. V C 2013 AIP Publishing LLC. [http://dx.

show abstract

Effect of lithium PFC coatings on NSTX density control

Cited by 56 publications

References 8 publications

A novel high-density embedded AND-type split gate flash memory

A novel high-density embedded AND-type split gate flash memory

Strike point control for the National Spherical Torus Experiment (NSTX)

Influence of plasma surface interactions on tokamak startup

Contact Info

Product

Resources

About