Impurity behavior is a growing topic, as proved by the increased experimental, theoretical and modelin g efforts. Radiation, transport, plasma-wall interact ion and first wall conditioning are indeed key aspe cts - where impurities play important roles - which need to be controlled and understood. The transition to tungsten plasma facing components seems to be an es sential step on the path to reactor scale fusion de vices and has been a major focus of work in tokamaks and laboratory based experiments in recent years. The drawback is the important impact of tungsten accumu lation mainly due to neoclassical inward pinch whic h can affect plasma performances, as in JET beam heat ed discharges with the ITER like wall [T. Puetteric h et al. "Tungsten Screening and Impurity Control in JET " 2012 IAEA -CN-197/EX/P3-15]. The headline of this contribution is the evidence t hat impurity core penetration is prevented in Reversed Field Pinch. This provides an interesting case for comparison and for improving the present understanding on impurity transport on different ma gnetic configurations and in different regions of t he fusion parameter space. In tokamaks an outward pinc h can occur when RF power is coupled to electrons, avoiding the impurity accumulation often affecting internal transport barrier regimes [Angioni]. In th e case of impurity hole observed in the LHD stellarator [K .Ida et al. Physiscs of Plasmas 16, 056111 ( 2009) ] , impurities are expelled from the core, after the fo rmation of an ion temperature internal transport ba rrier, due to an outward velocity driven by turbulence. Hi gh energy and low impurity confinement times have been obtained in high density H-mode (HDH) plasma i n the Wendelstein 7-AS stellarator [R. Burhenn et al. 29th EPS Conference on Plasma Phys. and Contr. Fusion Montreux, 17-21 June 2002 ECA Vol. 26B, P- 4.043 (2002)] The experimental evidence of impurity outward conve ction in RFX-mod is based on nickel Laser Blow Off injection and on neon gas puffing experime nts. It has been documented in various experimental conditions [S.Menmuir et al. Plasma Phys. Control. Fusion 52 (2010) 095001], including the improved confinement self-organized helical regimes occurrin g at high plasma current (I>1.2 MA), and confirmed by intrinsic C and O impurity behavior. The helical st ate features strong internal transport barriers for electron energy, in a region of null magnetic shear. The imp urity flux convective term is positive (outward) ov er the whole plasma radius. The outward pinch is higher in a localized radial region, where a poloidal flow s hear is also observed, forming a barrier which is wider and stronger in the helical regime. Such a barrier opposes the impurity penetration preventing core plasma con tamination. The main gas instead does not experienc e a significant outward pinch and shows just a reductio n of the particle diffusion inside the helical stru cture [D.Terranova et al., Nuclear Fusion 50 (2010) 03050 06]. With the present knowledge of the ion temperature profile, the outward pinch velocity of the impurities in RFX-mod cannot be ascribed to a classical effect. Gyrokinetic calculations (GS2) of turbulent transport to evaluate the effect of elec trostatic and electromagnetic turbulences on the impurity flu xes are compared with the experimental impurity convection. Impurity expulsion, due to classical temperature sc reening effect, has instead been found in MST RFP i n the enhanced confinement Pulsed Poloidal Current Dr ive scenario [S T A Kumar et al, Plasma Phys. Control. Fusion 53, 032001 (2011)]. In RFX- mod, the wall is entirely covered by graphi te tiles, so that the density behaviour is dominate d by recycling. In order to control the electron density and to overcome the H retention problem, modificat ion of C wall into a W one is foreseen. The impurity scree ning described above represents therefore an appeal ing feature in the perspective of the change towards a W wall. To check if W confirms the C, O, Ne and Ni outward convections, W Laser Blow Off injections wi ll be carried out.
Impurity screening in RFX-mod RFP plasmas
Lorella Carraro;
2013
Abstract
Impurity behavior is a growing topic, as proved by the increased experimental, theoretical and modelin g efforts. Radiation, transport, plasma-wall interact ion and first wall conditioning are indeed key aspe cts - where impurities play important roles - which need to be controlled and understood. The transition to tungsten plasma facing components seems to be an es sential step on the path to reactor scale fusion de vices and has been a major focus of work in tokamaks and laboratory based experiments in recent years. The drawback is the important impact of tungsten accumu lation mainly due to neoclassical inward pinch whic h can affect plasma performances, as in JET beam heat ed discharges with the ITER like wall [T. Puetteric h et al. "Tungsten Screening and Impurity Control in JET " 2012 IAEA -CN-197/EX/P3-15]. The headline of this contribution is the evidence t hat impurity core penetration is prevented in Reversed Field Pinch. This provides an interesting case for comparison and for improving the present understanding on impurity transport on different ma gnetic configurations and in different regions of t he fusion parameter space. In tokamaks an outward pinc h can occur when RF power is coupled to electrons, avoiding the impurity accumulation often affecting internal transport barrier regimes [Angioni]. In th e case of impurity hole observed in the LHD stellarator [K .Ida et al. Physiscs of Plasmas 16, 056111 ( 2009) ] , impurities are expelled from the core, after the fo rmation of an ion temperature internal transport ba rrier, due to an outward velocity driven by turbulence. Hi gh energy and low impurity confinement times have been obtained in high density H-mode (HDH) plasma i n the Wendelstein 7-AS stellarator [R. Burhenn et al. 29th EPS Conference on Plasma Phys. and Contr. Fusion Montreux, 17-21 June 2002 ECA Vol. 26B, P- 4.043 (2002)] The experimental evidence of impurity outward conve ction in RFX-mod is based on nickel Laser Blow Off injection and on neon gas puffing experime nts. It has been documented in various experimental conditions [S.Menmuir et al. Plasma Phys. Control. Fusion 52 (2010) 095001], including the improved confinement self-organized helical regimes occurrin g at high plasma current (I>1.2 MA), and confirmed by intrinsic C and O impurity behavior. The helical st ate features strong internal transport barriers for electron energy, in a region of null magnetic shear. The imp urity flux convective term is positive (outward) ov er the whole plasma radius. The outward pinch is higher in a localized radial region, where a poloidal flow s hear is also observed, forming a barrier which is wider and stronger in the helical regime. Such a barrier opposes the impurity penetration preventing core plasma con tamination. The main gas instead does not experienc e a significant outward pinch and shows just a reductio n of the particle diffusion inside the helical stru cture [D.Terranova et al., Nuclear Fusion 50 (2010) 03050 06]. With the present knowledge of the ion temperature profile, the outward pinch velocity of the impurities in RFX-mod cannot be ascribed to a classical effect. Gyrokinetic calculations (GS2) of turbulent transport to evaluate the effect of elec trostatic and electromagnetic turbulences on the impurity flu xes are compared with the experimental impurity convection. Impurity expulsion, due to classical temperature sc reening effect, has instead been found in MST RFP i n the enhanced confinement Pulsed Poloidal Current Dr ive scenario [S T A Kumar et al, Plasma Phys. Control. Fusion 53, 032001 (2011)]. In RFX- mod, the wall is entirely covered by graphi te tiles, so that the density behaviour is dominate d by recycling. In order to control the electron density and to overcome the H retention problem, modificat ion of C wall into a W one is foreseen. The impurity scree ning described above represents therefore an appeal ing feature in the perspective of the change towards a W wall. To check if W confirms the C, O, Ne and Ni outward convections, W Laser Blow Off injections wi ll be carried out.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
