Analysis and Optimization of the Impact of Ferromagnetic Inserts on the Toroidal Field Ripple in ITER

Lamzin, E.; Amoskov, Victor M.; Gapionok, E.; Gribov, Y.; Maximenkova, N.; Sytchevsky, S.

doi:10.1109/tasc.2011.2179402

Cited by 6 publications

(6 citation statements)

References 9 publications

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“…(1) Ferromagnetic structures of the six tritium breeding test blanket modules (TBMs), inserted in three dedicated equatorial vacuum vessel ports, directly facing the plasma [29] and (2) Ferromagnetic inserts (FIs)-ferromagnetic plates of the vacuum vessel shielding blocks located between inner and outer walls of the vacuum vessel located under the toroidal field coils at the outboard side of the vacuum vessel for reduction of the toroidal field ripple [30,31].…”

Section: Effect Of Ferromagnetic Structures On Plasma Vertical Instab...mentioning

confidence: 99%

Plasma vertical stabilisation in ITER

et al. 2015

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This paper describes the progress in analysis of the ITER plasma vertical stabilisation (VS) system since its design review in [2007][2008]. Two indices characterising plasma VS were studied. These are (1) the maximum value of plasma vertical displacement due to free drift that can be stopped by the VS system and (2) the maximum root mean square value of low frequency noise in the dZ/dt measurement signal used in the VS feedback loop. The first VS index was calculated using the PET code for 15 MA plasmas with the nominal position and shape. The second VS index was studied with the DINA code in the most demanding simulations for plasma magnetic control of 15 MA scenarios with the fastest plasma current ramp-up and early X-point formation, the fastest plasma current ramp-down in a divertor configuration, and an H to L mode transition at the current flattop. The studies performed demonstrate that the VS in-vessel coils, adopted recently in the baseline design, significantly increase the range of plasma controllability in comparison with the stabilising systems VS1 and VS2, providing operating margins sufficient to achieve ITER's goals specified in the project requirements.Additionally two sets of the DINA code simulations were performed with the goal of assessment of the capability of the PF system with the VS in-vessel coils: (i) to control the position of runaway electrons generated during disruptions in 15 MA scenarios and (ii) to trigger ELMs in H-mode plasmas of 7.5 MA /2.65 T scenarios planned for the early phase of ITER operation. It was also shown that ferromagnetic structures of the vacuum vessel (ferromagnetic inserts) and test blanket modules insignificantly affect the plasma VS.

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Section: Effect Of Ferromagnetic Structures On Plasma Vertical Instab...mentioning

confidence: 99%

Plasma vertical stabilisation in ITER

et al. 2015

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“…These equilibria will be used for a sensitivity study of transport with ripple. Note that ITER toroidal field ripple at the outer separatrix midplane for 5.3 T is expected to be 0.3% thanks to the use of ferromagnetic inserts [4], while without them it would be 1.0%. The same scan for the region around the X point.…”

Section: Mhd Equilibriummentioning

confidence: 99%

“…Despite this, tokamaks are not axisymmetric in practice. For instance, as many other tokamaks, ITER contains some elements in its design that break magnetic field symmetry such as: i) the finite number of toroidal field coils (this is common to all tokamaks) and ferromagnetic inserts to reduce the non-axisymmetry of the toroidal field [4], ii) the 3D fields created by the in-vessel ELM control coils, designed to suppress edge localized unstable modes (ELM) [5], iii) the Test Blanket Modules (TBMs) [6], which contain ferromagnetic steel, and will be used for demonstration of tritium production and iv) the magnetic field reduction systems [7] for the massive neutral beam injectors (NBI) to ensure a sufficiently low field in their interior. Although these three-dimensional magnetic fields should not affect the ITER plasma core, they can significantly influence the stability of the pedestal, plasma-wall interaction localization and the impurity transport dynamics [2].…”

Section: Introductionmentioning

confidence: 99%

Effect of non-axisymmetric perturbations on the ambipolar E _r and neoclassical particle flux inside the ITER pedestal region

Reynolds-Barredo¹,

Tribaldos²,

Loarte³

et al. 2020

Nucl. Fusion

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The transport dynamics of impurities in the pedestal region of ITER plasmas is of crucial interest since this regulates the penetration of impurities from the edge into the core plasma, where an excessive accumulation of impurities can degrade their fusion performance. In the pedestal region of H-mode tokamak plasmas, anomalous transport is highly reduced and impurity transport is found to be well described by the neoclassical theory. Under these conditions, perturbations to the axisymmetric tokamak geometry can strongly affect both the radial electric field and particle transport. In this work, we describe the results of numerical studies performed to quantify the effects on the pedestal ambipolar electric field and radial particle fluxes of the non-axisymmetric fields, associated with both the intrinsic toroidal field ripple and extrinsic fields applied for ELM control, for ITER Q = 10 plasma conditions with emphasis on high Z impurity transport. It is found that the effect of the ITER toroidal field ripple on high Z impurity transport is negligible. On the contrary, extrinsic three-dimensional fields applied for ELM control cause a strong modification of the pedestal ambipolar electric (to less negative values) and the appearance of multi-valued solutions for the pedestal electric field, analogue to core stellarator transport significantly increasing the outward character of neoclassical pedestal transport for both the main plasma ions (D and T in ITER) and high Z impurities, suggesting a strong modification of the background plasma profiles. Finally, it is found that for the Z impurity, its quantitative evaluation has uncertainties (with important implications for the radial flow direction) associated with the high poloidal Mach number ~ 1, due to the high pedestal electric field and large ion mass, which renders the first order neoclassical theory questionable. Detailed ITER MHD 3D equilibria and a benchmark of SFINCS against NEO and NCLASS codes has also been obtained.

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“…In this paper, the NB fast ion losses were calculated by considering their first orbits, however without considering the toroidal field (TF) coil ripple. A study on the NB fast ion losses induced by TF ripple using the ASCOT code [27] has shown a good confinement of the fast ions with the TF ripple of 0.34% [28] at 5.3 T.…”

Section: Heating and Current Drivementioning

confidence: 99%

Development of ITER non-activation phase operation scenarios

et al. 2017

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Non-activation phase operations in ITER in hydrogen (H) and helium (He) will be important for commissioning of tokamak systems, such as diagnostics, heating and current drive (HCD) systems, coils and plasma control systems, and for validation of techniques necessary for establishing operations in DT. The assessment of feasible HCD schemes at various toroidal fields (2.65-5.3 T) has revealed that the previously applied assumptions need to be refined for the ITER non-activation phase H/He operations. A study of the ranges of plasma density and profile shape using the JINTRAC suite of codes has indicated that the hydrogen pellet fuelling into He plasmas should be utilized taking the optimization of IC power absorption, neutral beam shine-through density limit and H-mode access into account. The EPED1 estimation of the edge pedestal parameters has been extended to various H operation conditions, and the combined EPED1 and SOLPS estimation has provided guidance for modelling the edge pedestal in H/He operations. The availability of ITER HCD schemes, ranges of achievable plasma density and profile shape, and estimation of the edge pedestal parameters for H/He plasmas have been integrated into various time-dependent tokamak discharge simulations. In this work, various H/He scenarios at a wide range of plasma current (7.5-15 MA) and field (2.65-5.3 T) have been developed for the ITER non-activation phase operation, and the sensitivity of the developed scenarios to the used assumptions has been investigated to provide guidance for further development.

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Analysis and Optimization of the Impact of Ferromagnetic Inserts on the Toroidal Field Ripple in ITER

Cited by 6 publications

References 9 publications

Plasma vertical stabilisation in ITER

Plasma vertical stabilisation in ITER

Effect of non-axisymmetric perturbations on the ambipolar E _r and neoclassical particle flux inside the ITER pedestal region

Development of ITER non-activation phase operation scenarios

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Analysis and Optimization of the Impact of Ferromagnetic Inserts on the Toroidal Field Ripple in ITER

Cited by 6 publications

References 9 publications

Plasma vertical stabilisation in ITER

Plasma vertical stabilisation in ITER

Effect of non-axisymmetric perturbations on the ambipolar E r and neoclassical particle flux inside the ITER pedestal region

Development of ITER non-activation phase operation scenarios

Contact Info

Product

Resources

About

Effect of non-axisymmetric perturbations on the ambipolar E _r and neoclassical particle flux inside the ITER pedestal region