| 1. |
- Murari, A., et al.
(författare)
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A control oriented strategy of disruption prediction to avoid the configuration collapse of tokamak reactors
- 2024
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Ingår i: Nature Communications. - 2041-1723 .- 2041-1723. ; 15:1
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Tidskriftsartikel (refereegranskat)abstract
- The objective of thermonuclear fusion consists of producing electricity from the coalescence of light nuclei in high temperature plasmas. The most promising route to fusion envisages the confinement of such plasmas with magnetic fields, whose most studied configuration is the tokamak. Disruptions are catastrophic collapses affecting all tokamak devices and one of the main potential showstoppers on the route to a commercial reactor. In this work we report how, deploying innovative analysis methods on thousands of JET experiments covering the isotopic compositions from hydrogen to full tritium and including the major D-T campaign, the nature of the various forms of collapse is investigated in all phases of the discharges. An original approach to proximity detection has been developed, which allows determining both the probability of and the time interval remaining before an incoming disruption, with adaptive, from scratch, real time compatible techniques. The results indicate that physics based prediction and control tools can be developed, to deploy realistic strategies of disruption avoidance and prevention, meeting the requirements of the next generation of devices.
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| 2. |
- Labit, B., et al.
(författare)
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Dependence on plasma shape and plasma fueling for small edge-localized mode regimes in TCV and ASDEX Upgrade
- 2019
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:8
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Tidskriftsartikel (refereegranskat)abstract
- © 2019 Institute of Physics Publishing. All rights reserved. Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (ne,sep/nG ∼ 0.3), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened.
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| 3. |
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| 4. |
- Kurki-Suonio, Taina, 1959, et al.
(författare)
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Effect of the European design of TBMs on ITER wall loads due to fast ions in the baseline (15 MA), hybrid (12.5 MA), steady-state (9 MA) and half-field (7.5 MA) scenarios
- 2016
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 56:11
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Tidskriftsartikel (refereegranskat)abstract
- We assess the effect of the European design of the pebble-bed helium-cooled test blanket modules (TBM) on fast ion power loads on ITER material surfaces. For this purpose, the effect of not only the TBMs but also the ferritic inserts (FI), used for mitigating the toroidal field ripple, were included in unprecedented detail in the reconstruction of the 3-dimensional magnetic field. This is important because, due to their low collisionality, fast ions follow the magnetic geometry much more faithfully than the thermal plasma. The Monte Carlo orbit-following code ASCOT was used to simulate all the foreseen operating scenarios of ITER: the baseline 15 MA standard H-mode operation, the 12.5 MA hybrid scenario, the 9 MA advanced scenario, and the half-field scenario with helium plasma that will be ITER's initial operating scenario. The effect of TBMs was assessed by carrying out the simulations in pairs: one including only the effect of ferritic inserts, and the other including also the perturbation due to TBMs. Both thermonuclear fusion alphas and NBI ions from ITER heating beams were addressed. The TBMs are found to increase the power loads, but the absolute values remain small. Neither do they produce any additional hot spots.
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| 5. |
- Kurki-Suonio, Taina, 1959, et al.
(författare)
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Protecting ITER walls: fast ion power loads in 3D magnetic field
- 2017
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Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 1361-6587 .- 0741-3335. ; 59:1
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Tidskriftsartikel (refereegranskat)abstract
- The fusion alpha and beam ion with steady-state power loads in all four main operating scenarios of ITER have been evaluated by the ASCOT code. For this purpose, high-fidelity magnetic backgrounds were reconstructed, taking into account even the internal structure of the ferritic inserts and tritium breeding modules (TBM). The beam ions were found to be almost perfectly confined in all scenarios, and only the so-called hybrid scenario featured alpha loads reaching 0.5 MW due to its more triangular plasma. The TBMs were not found to jeopardize the alpha confinement, nor cause any hot spots. Including plasma response did not bring dramatic changes to the load. The ELM control coils (ECC) were simulated in the baseline scenario and found to seriously deteriorate even the beam confinement. However, the edge perturbation in this case is so large that the sources have to be re-evaluated with plasma profiles that take into account the ECC perturbation.
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| 6. |
- Lier, A., et al.
(författare)
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The impact of fusion-born alpha particles on runaway electron dynamics in ITER disruptions
- 2023
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Ingår i: Nuclear Fusion. - 1741-4326 .- 0029-5515. ; 63:5
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Tidskriftsartikel (refereegranskat)abstract
- In the event of a tokamak disruption in a D-T plasma, fusion-born alpha particles take several milliseconds longer to thermalise than the background. As the damping rates drop drastically following the several orders of magnitudes drop of temperature, Toroidal Alfvén Eigenmodes (TAEs) can be driven by alpha particles in the collapsing plasma before the onset of the current quench. We employ kinetic simulations of the alpha particle distribution and show that the TAEs can reach sufficiently strong saturation amplitudes to cause significant core runaway electron (RE) transport in unmitigated ITER disruptions. As the eigenmodes do not extend to the plasma edge, this effect leads to an increase of the RE plateau current. Mitigation via massive material injection however changes the Alfvén frequency and can lead to mode suppression. A combination of the TAE-caused core RE transport with other perturbation sources could lead to a drop of runaway current in unmitigated disruptions.
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| 7. |
- Liu, Yueqiang, 1971, et al.
(författare)
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Modelling of 3D fields due to ferritic inserts and test blanket modules in toroidal geometry at ITER
- 2016
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 56:6, s. Art. no. 066001-
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Tidskriftsartikel (refereegranskat)abstract
- Computations in toroidal geometry are systematically performed for the plasma response to 3D magnetic perturbations produced by ferritic inserts (FIs) and test blanket modules (TBMs) for four ITER plasma scenarios: the 15 MA baseline, the 12.5 MA hybrid, the 9 MA steady state, and the 7.5 MA half-field helium plasma. Due to the broad toroidal spectrum of the FI and TBM fields, the plasma response for all the n = 1-6 field components are computed and compared. The plasma response is found to be weak for the high-n (n > 4) components. The response is not globally sensitive to the toroidal plasma flow speed, as long as the latter is not reduced by an order of magnitude. This is essentially due to the strong screening effect occurring at a finite flow, as predicted for ITER plasmas. The ITER error field correction coils (EFCC) are used to compensate the n = 1 field errors produced by FIs and TBMs for the baseline scenario for the purpose of avoiding mode locking. It is found that the middle row of the EFCC, with a suitable toroidal phase for the coil current, can provide the best correction of these field errors, according to various optimisation criteria. On the other hand, even without correction, it is predicted that these n = 1 field errors will not cause substantial flow damping for the 15 MA baseline scenario.
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| 8. |
- Sarkimaki, K., et al.
(författare)
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An advection-diffusion model for cross-field runaway electron transport in perturbed magnetic fields
- 2016
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Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 1361-6587 .- 0741-3335. ; 58:12
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Tidskriftsartikel (refereegranskat)abstract
- Disruption-generated runaway electrons (RE) present an outstanding issue for ITER. The predictive computational studies of RE generation rely on orbit-averaged computations and, as such, they lack the effects from the magnetic field stochasticity. Since stochasticity is naturally present in post-disruption plasma, and externally induced stochastization offers a prominent mechanism to mitigate RE avalanche, we present an advection-diffusion model that can be used to couple an orbit-following code to an orbit-averaged tool in order to capture the cross-field transport and to overcome the latter's limitation. The transport coefficients are evaluated via a Monte Carlo method. We show that the diffusion coefficient differs significantly from the well-known Rechester-Rosenbluth result. We also demonstrate the importance of including the advection: it has a two-fold role both in modelling transport barriers created by magnetic islands and in amplifying losses in regions where the islands are not present.
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| 9. |
- Tinguely, R. A., et al.
(författare)
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Modeling the complete prevention of disruption-generated runaway electron beam formation with a passive 3D coil in SPARC
- 2021
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 61:12
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Tidskriftsartikel (refereegranskat)abstract
- The potential formation of multi-mega-ampere beams of relativistic 'runaway' electrons (REs) during sudden terminations of tokamak plasmas poses a significant challenge to the tokamak's development as a fusion energy source. Here, we use state-of-the-art modeling of disruption magnetohydrodynamics coupled with a self-consistent evolution of RE generation and transport to show that a non-axisymmetric in-vessel coil will passively prevent RE beam formation during disruptions in the SPARC tokamak, a compact, high-field, high-current device capable of achieving a fusion gain Q > 2 in deuterium-tritium plasmas.
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| 10. |
- Tinguely, R. A., et al.
(författare)
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On the minimum transport required to passively suppress runaway electrons in SPARC disruptions
- 2023
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Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 1361-6587 .- 0741-3335. ; 65:3
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Tidskriftsartikel (refereegranskat)abstract
- In Izzo et al (2022 Nucl. Fusion 62 096029), state-of-the-art modeling of thermal and current quench (CQ) magnetohydrodynamics (MHD) coupled with a self-consistent evolution of runaway electron (RE) generation and transport showed that a non-axisymmetric (n = 1) in-vessel coil could passively prevent RE beam formation during disruptions in SPARC, a compact high-field tokamak projected to achieve a fusion gain Q > 2 in DT plasmas. However, such suppression requires finite transport of REs within magnetic islands and re-healed flux surfaces; conservatively assuming zero transport in these regions leads to an upper bound of RE current ∼ 1 M A compared to ∼ 8.7 M A of pre-disruption plasma current. Further investigation finds that core-localized electrons, within r / a < 0.3 and with kinetic energies ∼ 0.2 - 15 M e V , contribute most to the RE plateau formation. Yet only a relatively small amount of transport, i.e. a diffusion coefficient ∼ 18 m 2 s − 1 , is needed in the core to fully mitigate these REs. Properly accounting for (a) the CQ electric field’s effect on RE transport in islands and (b) the contribution of significant RE currents to disruption MHD may help achieve this.
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| 11. |
- Varje, J., et al.
(författare)
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Effect of plasma response on the fast ion losses due to ELM control coils in ITER
- 2016
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 56:4
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Tidskriftsartikel (refereegranskat)abstract
- Mitigating edge localized modes (ELMs) with resonant magnetic perturbations (RMPs) can increase energetic particle losses and resulting wall loads, which have previously been studied in the vacuum approximation. This paper presents recent results of fusion alpha and NBI ion losses in the ITER baseline scenario modelled with the Monte Carlo orbit following code ASCOT in a realistic magnetic field including the effect of the plasma response. The response was found to reduce alpha particle losses but increase NBI losses, with up to 4.2% of the injected power being lost. Additionally, some of the load in the divertor was found to be shifted away from the target plates toward the divertor dome.
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| 12. |
- Weckmann, Armin, et al.
(författare)
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Review on global migration, fuel retention and modelling after TEXTOR decommission
- 2018
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Ingår i: NUCLEAR MATERIALS AND ENERGY. - : ELSEVIER SCIENCE BV. - 2352-1791. ; 17, s. 83-112
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Forskningsöversikt (refereegranskat)abstract
- Before decommissioning of the TEXTOR tokamak in 2013, the machine was conditioned with a comprehensive migration experiment where MoF6 and N-15(2) were injected on the very last operation day. Thereafter, all plasmafacing components (PFCs) were available for extensive studies of both local and global migration of impurities - Mo, W, Inconel alloy constituents, 15 N, F - and fuel retention studies. Measurements were performed on 140 limiter tiles out of 864 throughout the whole machine to map global transport. One fifth of the introduced molybdenum could be found. Wherever possible, the findings are compared to results obtained previously in other machines. This review incorporates both published and unpublished results from this TEXTOR study and combines findings with analytical methods as well as modelling results from two codes, ERO and ASCOT. The main findings are: Both local and global molybdenum transport can be explained by toroidal plasma flow and (sic) x (sic) drift. The suggested transport scheme for molybdenum holds also for other analysed species, namely tungsten from previous experiments and medium-Z metals (Cr-Cu) introduced on various occasions. Analytical interpretation of several deposition profile features is possible with basic geometrical and plasma physics considerations. These are deposition profiles on the collector probe, the lower part of the inner bumper limiter, the poloidal cross-section of the inner bumper limiter, and the poloidal limiter. Any deposition pattern found in this TEXTOR study, including fuel retention, has neither poloidal nor toroidal symmetry, which is often assumed when determining deposition profiles on global scale. Fuel retention is highly inhomogeneous due to local variation of plasma parameters - by auxiliary heating systems and impurity injection - and PFC temperature. Local modelling with ERO yields good qualitative agreement but too high local deposition efficiency. Global modelling with ASCOT shows that the radial electric field and source form have a high impact on global deposition patterns, while toroidal flow has little influence. Some of the experimental findings could be reproduced. Still, qualitative differences between simulated and experimental global deposition patterns remain. The review closes with lessons learnt during this extensive TEXTOR study which might be helpful for future scientific exploitation of other tokamaks to be decommissioned.
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