Effect of neutral transport on ITER divertor performance

Kukushkin, A. S.; Pacher, H.D.; Kotov, V.; Reiter, D.; Coster, D.; Pacher, G.W.

doi:10.1088/0029-5515/45/7/008

Cited by 84 publications

(112 citation statements)

References 23 publications

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“…For ITER, SOLPS predictions for high power/high density scenarios required for fusion energy production, indicate that neutral penetration into the core plasma is negligible, breaking the link between separatrix density (controlled by gas puffing for divertor load control) and pedestal density (n ped , controlled by pellet fuelling). 45 These effects become more important at lower main plasma density (hn e i/n GW $ 0.5) for the separatrix values expected from SOLPS. EPED1 (Ref.…”

Section: -12mentioning

confidence: 98%

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

Sips

Giruzzi²,

Ide

et al. 2014

Physics of Plasmas

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The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ∼10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ∼ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q95 ∼ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at ne/nGW ∼ 0.85. Most experiments, however, obtain stable discharges at H98(y,2) ∼ 1.0 only for βN = 2.0–2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (li(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13–15 MA and densities down to ne/nGW ∼ 0.5, long pulse operation (>1000 s) in ITER is possible at Q ∼ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (>50 MW). H-mode operation in helium may be possible at input powers above 35 MW at a toroidal field of 2.65 T, for studying H-modes and ELM mitigation. In hydrogen, H-mode operation is expected to be marginal, even at 2.65 T with 60 MW of input power. Simulation code benchmark studies using hybrid and steady state scenario parameters have proved to be a very challenging and lengthy task of testing suites of codes, consisting of tens of sophisticated modules. Nevertheless, the general basis of the modelling appears sound, with substantial consistency among codes developed by different groups. For a hybrid scenario at 12 MA, the code simulations give a range for Q = 6.5–8.3, using 30 MW neutral beam injection and 20 MW ICRH. For non-inductive operation at 7–9 MA, the simulation results show more variation. At high edge pedestal pressure (Tped ∼ 7 keV), the codes predict Q = 3.3–3.8 using 33 MW NB, 20 MW EC, and 20 MW ion cyclotron to demonstrate the feasibility of steady-state operation with the day-1 heating systems in ITER. Simulations using a lower edge pedestal temperature (∼3 keV) but improved core confinement obtain Q = 5–6.5, when ECCD is concentrated at mid-radius and ∼20 MW off-axis current drive (ECCD or LHCD) is added. Several issues remain to be studied, including plasmas with dominant electron heating, mitigation of transient heat loads integrated in scenario demonstrations and (burn) control simulations in ITER scenarios.

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Section: -12mentioning

confidence: 98%

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

Sips

Giruzzi²,

Ide

et al. 2014

Physics of Plasmas

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“…Therefore, there are just few of such examples. In Kukushkin et al (2005b), Kotov et al (2006) the impact of radiation transport was considered for some ITER cases. It was found that even though the runs with and without radiation trapping effects show significant difference in the distribution of the plasma parameters, the results such as the maximum divertor heat load, crucial from the engineering point of view, do not change much.…”

Section: Radiation Transportmentioning

confidence: 99%

Physics of ultimate detachment of a tokamak divertor plasma

Krasheninnikov¹,

2017

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The basic physics of the processes playing the most important role in divertor plasma detachment is reviewed. The models used in the two-dimensional edge plasma transport codes that are widely used to address different issues of the edge plasma physics and to simulate the experimental data, as well as the numerical schemes and convergence issues, are described. The processes leading to ultimate divertor plasma detachment, the transition to and the stability of the detached regime, as well as the impact of the magnetic configuration and divertor geometry on detachment, are considered. A consistent, integral physical picture of ultimate detachment of a tokamak divertor plasma is developed.

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Section: Introductionmentioning

confidence: 99%

“…We now use routinely the B2-Eirene code package version solps4.2 [1][2][3][4], which relies on a nonlinear model of neutral particle transport, including neutral-neutral and molecule-ion collisions. The model assumptions are those of [1]. We fix the plasma power entering the scrape-off layer (SOL) at 100 MW, fix the D ion flow from the core across the core-edge interface (CEI) at 17 € Pa -m 3 s −1 , and vary the density WITH the flux injected by gas puffing.…”

Section: Introductionmentioning

confidence: 99%

Impurity seeding and scaling of edge parameters in ITER

Pacher

Kukushkin

Pacher

et al. 2009

Journal of Nuclear Materials

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ITER divertor and edge modelling with the ITER B2-EIRENE code including neutral-neutral and molecule-ion collisions has led to updated scaling of the helium density and flux and the DT flux with the pumping speed and the divertor neutral pressure. The replacement of carbon by the addition of seeded impurities (neon) strongly modifies the upstream DT density and DT neutral influx at constant detachment state, i.e. divertor neutral pressure and has a smaller influence on the peak divertor power load. The core plasma performance is modelled with the integrated Core-Pedestal-SOL (ICPS) model implemented in ASTRA; the operating window with seeded impurities is smaller than with DT injection alone.

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Effect of neutral transport on ITER divertor performance

Cited by 84 publications

References 23 publications

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

Physics of ultimate detachment of a tokamak divertor plasma

Impurity seeding and scaling of edge parameters in ITER

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