Simulation of JET ITER-Like Wall pulses at high neon seeding rate

Telesca, G.; Ivanova‐Stanik, I.; Brezinsek, S.; Czarnecka, A.; Drewelow, P.; Giroud, C.; Huber, A.; Wiesen, S.; Wischmeier, M.; Zagórski, R.; Contributors, Jet

doi:10.1088/1741-4326/aa8381

Cited by 10 publications

(10 citation statements)

References 23 publications

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“…Numerical modeling of seeded discharges [14,15,16,17] leads to a similar conclusion: nitrogen seeding allows for operation with an acceptable level of core confinement and high radiative fraction in the divertor, while for noble gases to find such an operational window is more difficult. These result might indicate that noble gases leak from the divertor more efficiently than the nitrogen.…”

Section: Introductionmentioning

confidence: 80%

On mechanisms of impurity leakage and retention in the tokamak divertor

Senichenkov

Kaveeva

Sytova

et al. 2019

Plasma Phys. Control. Fusion

119

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Impurity seeding into a tokamak divertor for radiative cooling is considered as a tool for achieving detached/semi-detached regimes required to meet the condition of acceptable heat loads on divertor plates. Experiments aimed at searching of the operational window with a significant reduction of poloidal heat fluxes due to the impurity radiation and without the decreasing of confinement are performed on many tokamaks. Critical issue in these experiments is which fraction of impurities is retained in the divertor region and which is extracted upstream to the scrape-off layer (SOL). In the present paper a physical mechanism of impurity transport from a divertor towards upstream and back to the divertor is analyzed. It is demonstrated that the widespread concept that the impurity leaks if the parallel thermal force exceeds the friction due to main ions and retains otherwise-is not correct. In contrast, the impurity leaks if it crosses the stagnation point of impurity ion poloidal velocity profile before the ionization, and retains if it ionizes closer to the target than the location of that stagnation point. Thus the leakage efficiency depends on the relative spatial positions of the impurity atom ionization source and the stagnation point of the impurity ion poloidal velocity profile. The impurity ion poloidal velocity is a sum of poloidal projection of its parallel velocity and the E × B drift velocity, where the former should be defined from the parallel impurity force balance equation. It is demonstrated that the solution of this equation may be approximated by the balance of friction and thermal forces in all regimes, while other terms are smaller. This allows for expressing the impurity parallel velocity through the main ion one and makes the distribution of the parallel (poloidal) fluxes of the main ions, including Pfirsch-Schlüter (PS) fluxes and E × B drift fluxes, to be an important element of the impurity transport. It is shown that impurity distribution in the edge plasma is rather sensitive to the value of the impurity ion ionization potential. This analysis is supported by the simulation results obtained for the ASDEX Upgrade tokamak with various seeding rates of N and Ne with the SOLPS-ITER code. The importance of inclusion of self-consistent drift flows is demonstrated by the comparison to result of corresponding simulations with drifts turned off.

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Section: Introductionmentioning

confidence: 80%

On mechanisms of impurity leakage and retention in the tokamak divertor

Senichenkov

Kaveeva

Sytova

et al. 2019

Plasma Phys. Control. Fusion

119

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“…Simulations were performed using COREDIV code which is based on an integrated approach coupling the radial transport in the core and the 2D multifluid model of the SOL. As this work is a follow-up on our previous calculations the detailed description and parameters used in COREDIV can be found in references [12][13][14][15] and only the main points of the model are reported here.…”

Section: The Corediv Modelmentioning

confidence: 99%

Influence of the impurities in the hybrid discharges with high power in JET ILW

Ivanova‐Stanik¹,

Challis²,

Chomiczewska³

et al. 2022

Nucl. Fusion

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The aim of this paper is to numerically study the influences of the impurities on the high power hybrid discharges in the JET ITER-like wall (ILW) configuration in the DD and deuterium–tritium (DT) scenarios. Numerical simulations with the COREDIV code of hybrid discharges with 32 MW auxiliary heating, 2.2 MA plasma current and 2.8 T toroidal magnetic field in the ILW corner configuration are presented. In the simulations five impurity species are used: intrinsic: beryllium (Be) and nickel (Ni) from the side walls, helium (He) from DT reaction, tungsten (W) from divertor and extrinsic neon (Ne) or argon (Ar) by gas puff. The extrapolation of the DD discharges to DT plasmas at the original input power of 32 MW and taking into account only the thermal component of the alpha-power, does not show any significant difference regarding the power to the target with respect to the DD case. Simulations show that sputtering due to D and T is negligible. In contrast, the simulations at auxiliary heating 39 MW show that the power to the target is possibly too high to be sustained for about 5 s by strike-point sweeping alone without any control by Ne seeding. The tungsten is produced mainly by Ni, Be and seeded impurities.

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“…[1,2] Experiments on impurity seeding using different gases (mainly N, Ne, and Ar, more rare Kr) are performed on many machines for several years, [3][4][5][6][7][8][9][10][11][12] and they are supported by analytical models and numerical modelling. [13][14][15][16][17] It is widely accepted that as the impurty seeding rate increases, target operation regime moves from the attachment (strong heat flux density peak near the strike point) through the partial detachment (notable reduction of the heat flux density in the strike point vicinity), the pronounced detachment (notable decrease of maximal heat flux density and maximal electron temperature in the far scrape-off layer [SOL]) to the full detachment (flattening of heat flux density and electron temperature along whole target with typical values below 1 MW∕m 2 and 2 eV correspondingly). [5] The disadvantage of intensive impurity seeding is the possible impurity leakage from the divertor, which leads to the impurity penetration into the confined region and to the fuel dilution.…”

Section: Introductionmentioning

confidence: 99%

Detached regime with highly radiating X‐point: Physics and modelling

Senichenkov

Rozhansky

Shtyrkhunov

et al. 2022

Contributions to Plasma Physics

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A possibility to operate with fully detached targets and highly radiating X-point was recently demonstrated on ASDEX Upgrade, JET, and DIII-D with intensive impurity seeding (N, Ne, Ar) and feedback control. Notable feature of such a regime is the radiated power fraction of up to 90% of discharge power inside the separatrix without a confinement degradation, or even with confinement improvement (in terms of H-factor), which leads to a full detachment of both targets. Therefore, this regime might be attractive for the next generation reactor scale tokamaks like DEMO or CFETR, where most of the power should be radiated. In the present report, the recent achievements in the understanding of such regime are reviewed. The effectiveness of divertor target shielding by different impurity gases is discussed, paying attention to its first ionization potential. SOLPS-ITER modelling results illustrating the impurity flow pattern are presented, and the importance of drift flows is demonstrated. The effect of the machine size on the effectiveness of the divertor shielding by impurity radiation is discussed. It is demonstrated that in bigger machines impurity radiation is better kept in the divertor and divertor asymmetry is less pronounced. As the seeding rate increases, both targets fully detach, and neutrals start to penetrate inside the separatrix above the X-point, leading to the plasma cooling there and to the formation of an observable localized radiation spot in the X-point vicinity.The cold zone above the X-point thus behaves as an energy sink like a divertor in conventional regime, so that the perpendicular decay length of perpendicular turbulent energy flow appears to be of the order of 𝜆 q of conventional divertor.During the formation of the radiating X-point a potential hill/well (depending on the direction of B × ∇B drift) is formed on perturbed closed flux surfaces, and the corresponding E × B drift vortex fluxes appear to give major contribution to the particle balance. SOLPS-ITER simulations show that the radial electric field significantly deviates from neoclassical formula and may even change sign (become positive). Thus, the zone of maximal poloidal rotation shear may shift inwards, which might be responsible for the inward shift of the transport barrier and for the confinement improvement. This effect is different to the discussed extension of peeling-ballooning stability boundary by intensive impurity injection, which

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Simulation of JET ITER-Like Wall pulses at high neon seeding rate

Cited by 10 publications

References 23 publications

On mechanisms of impurity leakage and retention in the tokamak divertor

On mechanisms of impurity leakage and retention in the tokamak divertor

Influence of the impurities in the hybrid discharges with high power in JET ILW

Detached regime with highly radiating X‐point: Physics and modelling

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