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| 2. |
- Coelho, R., et al.
(författare)
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Synthetic Diagnostics In The European Union Integrated Tokamak Modelling Simulation Platform
- 2013
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Ingår i: Fusion science and technology. - 1536-1055 .- 1943-7641. ; 63:1, s. 1-8
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Tidskriftsartikel (refereegranskat)abstract
- The European Union Integrated Tokamak Modelling Task Force (ITM-TF) has developed a standardized platform and an integrated modeling suite of codes for the simulation and prediction of a complete plasma discharge in any tokamak. The framework developed by ITM-TF allows for the development of sophisticated integrated simulations (workflows) for physics application, e.g., free-boundary equilibrium with feedback control, magnetohydrodynamic stability analysis, core/edge plasma transport, and heating and current drive. A significant effort is also under way to integrate synthetic diagnostic modules in the ITM-TF environment, namely, focusing on three-dimensional reflectometry, motional Stark effect, and neutron and neutral particle analyzer diagnostics. This paper gives an overview of the conceptual design of ITM-TF and preliminary results of the aforementioned synthetic diagnostic modules.
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| 3. |
- 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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| 4. |
- 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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| 5. |
- 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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| 6. |
- 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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