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| 2. |
- Coburn, J., et al.
(författare)
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Energy deposition and melt deformation on the ITER first wall due to disruptions and vertical displacement events
- 2022
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 62:1
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Tidskriftsartikel (refereegranskat)abstract
- An analysis workflow has been developed to assess energy deposition and material damage for ITER vertical displacement events (VDEs) and major disruptions (MD). This paper describes the use of this workflow to assess the melt damage to be expected during unmitigated current quench (CQ) phases of VDEs and MDs at different points in the ITER research plan. The plasma scenarios are modeled using the DINA code with variations in plasma current I (p), disruption direction (upwards or downwards), Be impurity density n (Be), and diffusion coefficient chi. Magnetic field line tracing using SMITER calculates time-dependent, 3D maps of surface power density q (perpendicular to) on the Be-armored first wall panels (FWPs) throughout the CQ. MEMOS-U determines the temperature response, macroscopic melt motion, and final surface topology of each FWP. Effects of Be vapor shielding are included. Scenarios at the baseline combination of I (p) and toroidal field (15 MA/5.3 T) show the most extreme melt damage, with the assumed n (Be) having a strong impact on the disruption duration, peak q (perpendicular to) and total energy deposition to the first wall. The worst-cases are upward 15 MA VDEs and MDs at lower values of n (Be), with q (perpendicular to,max) = 307 MW m(-2) and maximum erosion losses of similar to 2 mm after timespans of similar to 400-500 ms. All scenarios at 5 MA avoided melt damage, and only one 7.5 MA scenario yields a notable erosion depth of 0.25 mm. These results imply that disruptions during 5 MA, and some 7.5 MA, operating scenarios will be acceptable during the pre-fusion power operation phases of ITER. Preliminary analysis shows that localized melt damage for the worst-case disruption should have a limited impact on subsequent stationary power handling capability.
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| 3. |
- Coburn, J., et al.
(författare)
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First wall energy deposition during vertical displacement events on ITER
- 2020
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Ingår i: Physica Scripta. - : IOP Publishing. - 0031-8949 .- 1402-4896. ; T171:1
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Tidskriftsartikel (refereegranskat)abstract
- The beryllium (Be) first wall energy deposition and melt damage profiles resulting from the current quench phase of an unmitigated, 5 MA/1.8 T upward vertical displacement event for ITER are investigated. Time dependent 2D magnetic flux profiles are calculated with the DINA code and used as input for the SMITER 3D field line tracing software. 3D maps of the wetted area and perpendicular heat flux q(perpendicular to) show that the majority of the energy deposition occurs on the upper first wall panels #8 and #9 SMITER simulations predict q(perpendicular to,peak) approximate to 190 MW m(-2) on the surfaces of upper FWPs #8 and #9 at the end of the similar to 450 ms current quench. The surface heat flux maps generated by SMITER are used as input in the MEMOS-U code, which models Be melt formation and dynamics. Simulations reveal peak surface temperatures of similar to 2200 K, inward surface damage of similar to 0.5 mm in depth, and average melt velocities of similar to 2 m s(-1). Although VDEs are in principle the easiest disruptive instability to avoid, the analysis demonstrates that any non-mitigated events or intentional VDEs taking place during low I-p, early operational phases of ITER for the purposes of estimating disruption forces, must be kept to a low number.
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| 4. |
- Coburn, J., et al.
(författare)
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Reassessing energy deposition for the ITER 5 MA vertical displacement event with an improved DINA model
- 2021
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Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 28
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Tidskriftsartikel (refereegranskat)abstract
- The beryllium (Be) main chamber wall interaction during a 5 MA/1.8 T upward, unmitigated VDE scenario, first analysed in [J. Coburn et al., Phys. Scr. T171 (2020) 014076] for ITER, has been re-evaluated using the latest energy deposition analysis software. Updates to the DINA disruption model are summarized, including an improved numerical convergence for the OD power balance, limitations on the safety factor within the plasma core, and the choice to maintain a constant plasma + halo poloidal cross-section. Such updates result in a broad halo region and higher radiated power fractions compared to previous models. The new scenario lasts for similar to 75 ms and deposits similar to 29 MJ of energy, with the radial distribution of parallel heat flux q parallel to(r) resembling an exponential falloff with an effective lambda(q) = 75 -198 mm. A maximum halo width w(h) of 0.52 m at the outboard midplane is observed. SMITER field line tracing and energy deposition simulations calculate a q(perpendicular to,max) of similar to 83 MW/m(2) on the upper first wall panels (FWP). Heat transfer calculations with the MEMOS-U code show that the FWP surface temperature reaches similar to 1000 K, well below the Be melt threshold. Variations of this 5 MA scenario with Be impurity densities from 0 to 3.10(19) m(-3) also remain below the melt threshold despite differences in energy deposition and duration. These results are in contrast to the early study which predicted melt damage to the first wall [J. Coburn et al., Phys. Scr. T171 (2020) 014076], and emphasize the importance of accurate models for the halo width w(h) and the heat flux distribution q parallel to(r) within that halo width. The 2020 halo model in DINA has been compared with halo current experiments on COMPASS, JET, and Alcator C-Mod, and the preliminary results build confidence in the broad halo width predictions. Results for the 5 MA VDE are compared with those for a 15 MA equivalent, generated using the new DINA model. At the higher current, significant melting of the upper FWP is to be expected.
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| 5. |
- Corre, Y., et al.
(författare)
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Sustained W-melting experiments on actively cooled ITER-like plasma facing unit in WEST
- 2021
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Ingår i: Physica Scripta. - : IOP Publishing. - 0031-8949 .- 1402-4896. ; 96:12
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Tidskriftsartikel (refereegranskat)abstract
- The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the first time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m(-2) for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230 mu m followed by a re-solidified tungsten bump of 200 mu m in the JxB direction.
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| 7. |
- 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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| 8. |
- Meyer, H.F., et al.
(författare)
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Overview of physics studies on ASDEX Upgrade
- 2019
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:11
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Forskningsöversikt (refereegranskat)abstract
- The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q 95 = 5.5, ) at low density. Higher installed electron cyclotron resonance heating power 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG. Stable high-density H-modes with MW m-1 with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently . This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated. Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of and E r allow for inter ELM transport analysis confirming that E r is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of 'natural' no ELM regimes have been extended. Stable I-modes up to have been characterised using -feedback. Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle - measured for the first time - or the cross-phase angle of and fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvénic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO.
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| 9. |
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| 10. |
- Ratynskaia, Svetlana V., et al.
(författare)
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Resolidification-controlled melt dynamics under fast transient tokamak plasma loads
- 2020
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Ingår i: Nuclear Fusion. - : Institute of Physics (IOP). - 0029-5515 .- 1741-4326. ; 60:10
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Tidskriftsartikel (refereegranskat)abstract
- Studies of macroscopic melt motion induced by fast transient power loads and the ensuing damage to plasma-facing components are critical for ITER. Based on state-of-the-art experiments, heat transfer is argued to be strongly entangled with fluid dynamics. The physics model of the MEMOS-U code is introduced. Simulations are reported of recent tokamak experiments concerning deliberate transient melting of beryllium main chamber tiles (JET) and tungsten divertor components (ASDEX Upgrade, JET). MEMOS-U is demonstrated to capture the main physics responsible for melt dynamics and to reproduce the observed surface deformation. The intricate role of resolidification is elucidated.
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