| 1. |
-
- 2018
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:1
-
Forskningsöversikt (refereegranskat)
|
|
| 2. |
- Bombarda, F., et al.
(författare)
-
Runaway electron beam control
- 2019
-
Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 1361-6587 .- 0741-3335. ; 61:1
-
Tidskriftsartikel (refereegranskat)
|
|
| 3. |
- Meyer, H., et al.
(författare)
-
Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution
- 2017
-
Ingår i: Nuclear Fusion. - : Institute of Physics Publishing (IOPP). - 0029-5515 .- 1741-4326. ; 57:10
-
Tidskriftsartikel (refereegranskat)abstract
- Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n = 2 RMP maintaining good confinement H-H(98,H-y2) approximate to 0.95. Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes.
|
|
| 4. |
- Meyer, H., et al.
(författare)
-
Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution
- 2017
-
Ingår i: Nuclear Fusion. - : Institute of Physics Publishing (IOPP). - 0029-5515 .- 1741-4326. ; 57:10
-
Tidskriftsartikel (refereegranskat)abstract
- Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n = 2 RMP maintaining good confinement H-H(98,H-y2) approximate to 0.95. Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes.
|
|
| 5. |
- Reimerdes, H., et al.
(författare)
-
Overview of the TCV tokamak experimental programme
- 2022
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 62:4
-
Tidskriftsartikel (refereegranskat)abstract
- The tokamak a configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019-20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T (e)/T (i) similar to 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with 'small' (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019-20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.
|
|
| 6. |
- Coda, S., et al.
(författare)
-
Physics research on the TCV tokamak facility: From conventional to alternative scenarios and beyond
- 2019
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:11
-
Tidskriftsartikel (refereegranskat)abstract
- The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device's unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly non-inductive H-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power 'starvation' reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached L-mode phase, increasing the outer connection length reduces the in-out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and possibly divertor pumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECRH and 1 MW neutral beam injection heating will be added.
|
|
| 7. |
- Coda, S., et al.
(författare)
-
Overview of the TCV tokamak program : Scientific progress and facility upgrades
- 2017
-
Ingår i: Nuclear Fusion. - : Institute of Physics Publishing. - 0029-5515 .- 1741-4326. ; 57:10
-
Tidskriftsartikel (refereegranskat)abstract
- The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from progress in individual controller design and have evolved steadily towards controller integration, mostly within an environment supervised by a tokamak profile control simulator. TCV has demonstrated effective wall conditioning with ECRH in He in support of the preparations for JT-60SA operation.
|
|
| 8. |
- Labit, B., et al.
(författare)
-
Dependence on plasma shape and plasma fueling for small edge-localized mode regimes in TCV and ASDEX Upgrade
- 2019
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:8
-
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.
|
|
| 9. |
- Marco, Aitor, et al.
(författare)
-
A Variable Structure Control Scheme Proposal for the Tokamak a Configuration Variable
- 2019
-
Ingår i: Complexity. - : Hindawi Publishing Corporation. - 1076-2787 .- 1099-0526.
-
Tidskriftsartikel (refereegranskat)abstract
- Fusion power is the most significant prospects in the long-term future of energy in the sense that it composes a potentially clean, cheap, and unlimited power source that would substitute the widespread traditional nonrenewable energies, reducing the geographical dependence on their sources as well as avoiding collateral environmental impacts. Although the nuclear fusion research started in the earlier part of 20th century and the fusion reactors have been developed since the 1950s, the fusion reaction processes achieved have not yet obtained net power, since the generated plasma requires more energy to achieve and remain in necessary particular pressure and temperature conditions than the produced profitable energy. For this purpose, the plasma has to be confined inside a vacuum vessel, as it is the case of the Tokamak reactor, which consists of a device that generates magnetic fields within a toroidal chamber, being one of the most promising solutions nowadays. However, the Tokamak reactors still have several issues such as the presence of plasma instabilities that provokes a decay of the fusion reaction and, consequently, a reduction in the pulse duration. In this sense, since long pulse reactions are the key to produce net power, the use of robust and fast controllers arises as a useful tool to deal with the unpredictability and the small time constant of the plasma behavior. In this context, this article focuses on the application of robust control laws to improve the controllability of the plasma current, a crucial parameter during the plasma heating and confinement processes. In particular, a variable structure control scheme based on sliding surfaces, namely, a sliding mode controller (SMC) is presented and applied to the plasma current control problem. In order to test the validity and goodness of the proposed controller, its behavior is compared to that of the traditional PID schemes applied in these systems, using the RZIp model for the Tokamak a Configuration Variable (TCV) reactor. The obtained results are very promising, leading to consider this controller as a strong candidate to enhance the performance of the PID-based controllers usually employed in this kind of systems.
|
|
| 10. |
- Buttery, R. J., et al.
(författare)
-
The Impact of 3-D Fields on Tearing Mode Stability of H-modes
- 2011
-
Ingår i: Nuclear Fusion. - 1741-4326 .- 0029-5515. ; 51, s. 073016-
-
Tidskriftsartikel (refereegranskat)abstract
- New processes have been discovered in the interaction of 3D fields with tearing mode stability at low torque and modest β on DIII-D and NSTX. These are thought to arise from the plasma response at the tearing resonant surface, which theoretically is expected to depend strongly on plasma rotation and underlying intrinsic tearing stability. This leads to sensitivities additional to those previously identified at low density where the plasma rotation is more readily stopped, or at high βN where ideal MHD responses amplify the fields (where βN is the plasma β divided by the ratio of plasma current to minor radius multiplied by toroidal field). It is found that the threshold size for 3D fields to induce modes tends to zero as the natural tearing βN limit is approached. 3D field sensitivity is further enhanced at low rotation, with magnetic probing detecting an increased response to applied fields in such regimes. Modelling with the MARS-F code confirms the interpretation with the usual plasma screening response breaking down in low rotation plasmas and a tearing response developing, opening the door to additional sensitivities to β and the current profile. Typical field thresholds to induce modes in torque-free βN ~ 1.5 H-modes are well below those in ohmic plasmas or plasmas near the ideal βN limit. The strong interaction with the tearing mode βN limit is identified through rotation shear, which is decreased by the 3D field, leading to decreased tearing stability. Thus both locked and rotating mode field thresholds can be considered in terms of a torque balance, with sufficient braking leading to destabilization of a mode. On this basis new measurements of the principal parameter scalings for error field threshold have been obtained in torque-free H-modes leading to new predictions for error field sensitivity in ITER. The scalings have similar exponents to ohmic plasmas, but with seven times lower threshold at the ITER baseline βN value of 1.8, and a linear dependence on proximity to the tearing mode βN limit (~2.2 at zero torque). This reinforces the need to optimize error field correction strategies in ITER, and implement sources to drive plasma rotation.
|
|
| 11. |
|
|
| 12. |
- Lanctot, M. J., et al.
(författare)
-
Validation of the linear ideal magnetohydrodynamic model of three-dimensional tokamak equilibria
- 2010
-
Ingår i: Physics of Plasmas. - : AIP Publishing. - 1089-7674 .- 1070-664X. ; 17:3, s. 030701-
-
Tidskriftsartikel (refereegranskat)abstract
- The first quantitative comparison of linear ideal magnetohydrodynamic (MHD) theory with external magnetic measurements of the nonaxisymmetric plasma perturbation driven by external long-wavelength magnetic fields in high-temperature tokamak plasmas is presented. The comparison yields good (within 20%) agreement for plasma pressures up to ∼ 75% of the ideal stability limit calculated without a conducting wall. For higher plasma pressures, the ideal MHD model tends to overestimate the perturbed field indicating the increasing importance of stabilizing nonideal effects.
|
|
| 13. |
- Okabayashi, M., et al.
(författare)
-
Control of the resistive wall mode with internal coils in the DIII-D tokamak
- 2005
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 45:12, s. 1715-1731
-
Tidskriftsartikel (refereegranskat)abstract
- Internal coils, 'I-Coils', were installed inside the vacuum vessel of the DIII-D device to generate non-axisymmetric magnetic fields to act directly on the plasma. These fields are predicted to stabilize the resistive wall mode (RWM) branch of the long-wavelength external kink mode with plasma beta close to the ideal wall limit. Feedback using these I-Coils was found to be more effective as compared to using external coils located outside the vacuum vessel. Locating the coils inside the vessel allows for a faster response and the coil geometry also allows for better coupling to the helical mode structure. Initial results were reported previously (Strait E.J. et al 2004 Phys. Plasmas 11 2505). This paper reports on results from extended feedback stabilization operations, achieving plasma parameters up to the regime of Cβ ≈ 1.0 and open loop growth rates of γopenτw ≳ 25 where the RWM was predicted to be unstable with only the 'rotational viscous stabilization mechanism'. Here Cβ ≈ (β - βno-wall.limit)/(βideal.wall.limit - βno-wall.limit) is a measure of the beta relative to the stability limits without a wall and with a perfectly conducting wall, and τw is the resistive flux penetration time of the wall. These feedback experimental results clarified the processes of dynamic error field correction and direct RWM stabilization, both of which took place simultaneously during RWM feedback stabilization operation. MARS-F modelling provides a critical rotation velocity in reasonable agreement with the experiment and predicts that the growth rate increases rapidly as rotation decreases below the critical. The MARS-F code also predicted that for successful RWM magnetic feedback, the characteristic time of the power supply should be limited to a fraction of the growth time of the targeted RWM. The possibility of further improvements in the presently achievable range of operation of feedback gain values is also discussed.
|
|
| 14. |
|
|
| 15. |
|
|
| 16. |
|
|
| 17. |
- Chu, M.S., et al.
(författare)
-
Response of a resistive and rotating tokamak to external magnetic perturbations below the Alfven frequency
- 2011
-
Ingår i: Nuclear Fusion. - 1741-4326 .- 0029-5515. ; 51, s. 073036-
-
Tidskriftsartikel (refereegranskat)abstract
- Motivated by the recent experimental observation that plasma stability can be improved by external magnetic perturbations, the general problem of plasma response to external magnetic perturbations is investigated. Different (vacuum, ideal and resistive) plasma response models are considered and compared. Plasma response, in experiments where stabilization was achieved, is obtained through computation using the MARS-F code, with a plasma model that includes both plasma resistivity and rotation. The resultant magnetic field line stochasticity is much reduced from that obtained formerly using the vacuum plasma model. This reduced stochasticity is more consistent with the favourable experimental observation of enhanced stability. Examples are given for the response of an ITER plasma to perturbations generated by the correction coils; and the response of a plasma to external coils (antenna) up to the Alfvén frequency.
|
|
| 18. |
- In, Y., et al.
(författare)
-
Requirements for active resistive wall mode (RWM) feedback control
- 2010
-
Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 1361-6587 .- 0741-3335. ; 52:10, s. 104004-
-
Tidskriftsartikel (refereegranskat)abstract
- The requirements for active resistive wall mode (RWM) feedback control have been systematically investigated and established using highly reproducible current-driven RWMs in ohmic discharges in DIII-D. The unambiguous evaluation of active RWM feedback control was not possible in previous RWM studies primarily due to the variability of the onset of the pressure-driven RWMs; the stability of the pressure-driven RWM is thought to be sensitive to various passive stabilization mechanisms. Both feedback control specifications and physics requirements for RWM stabilization have been clarified using the current-driven RWMs in ohmic discharges, when little or no passive stabilization effects are present. The use of derivative gain on top of proportional gain is found to be advantageous. An effective feedback control system should be equipped with a power supply with bandwidth greater than the RWM growth rate. It is beneficial to apply a feedback field that is toroidally phase-shifted from the measured RWM phase in the same direction as the plasma current. The efficacy of the RWM feedback control will ultimately be determined by the plasma fluctuations on internal diagnostics, as well as on external magnetics. The proximity of the feedback coils to the plasma appears to be an important factor in determining the effectiveness of the RWM feedback coils. It is desirable that an RWM feedback control system simultaneously handles error field correction at a low frequency, along with direct RWM feedback at a high frequency. There is an indication of the influence of a second least stable RWM, which had been theoretically predicted but never identified in experiments. A preliminary investigation based on active MHD spectroscopic measurement showed a strong plasma response around 400 Hz where the typical plasma response associated with the first least stable RWM was expected to be negligible. Present active feedback control requirements are based on a single mode assumption, so the investigation of the second least stable RWM is of high interest.
|
|
| 19. |
- La Haye, R.J., et al.
(författare)
-
Scaling of the Plasma Rotation Needed for Stabilizing the n = 1 Resistive Wall Mode (Ideal Kink) in the DIII D Tokamak
- 2004
-
Ingår i: Nuclear Fusion. - 1741-4326 .- 0029-5515. ; 44, s. 1197-
-
Tidskriftsartikel (refereegranskat)abstract
- Experiments in the DIII-D tokamak show that the n = 1 ideal kink can be stabilized by a resistive wall if the plasma is rotating fast enough. A database of the onset of the n = 1 resistive wall mode as a function of the equilibrium toroidal magnetic field, the plasma density and the toroidal rotation has been assembled for plasmas with beta between the theoretically predicted no wall and ideal wall stability limits. The critical rotation frequency is found to scale as the inverse of the Alfvén time with ? ?A 0.02 (evaluated at the q = 2 surface at ? 0.6) or ? ?S 0.7, where ?S is the sound time. The dependence of ? ? A or ? ?S on ?N/?N,no wall from 1?2 is weak and suggests the plasmas are in the 'intermediate dissipation' regime.
|
|
| 20. |
- Lanctot, M.J., et al.
(författare)
-
Measurement and modeling of three-dimensional equilibriua in DIII-D
- 2011
-
Ingår i: Physics of Plasmas. - 1089-7674 .- 1070-664X. ; 18, s. 056121-
-
Tidskriftsartikel (refereegranskat)abstract
- A detailed experiment-theory comparison reveals that linear ideal MHD theory is in quantitative agreement with external magnetic and internal soft x-ray measurements of the plasma response to externally applied non-axisymmetric fields over a broad range of beta and rotation. This result represents a significant step toward the goal of advancing the understanding of three-dimensional tokamak equilibria. Both the magnetic and soft x-ray measurements show the driven plasma perturbation increases linearly with the applied perturbation, suggesting the relevance of linear plasma response models. The magnetic and soft x-ray measurements are made at multiple toroidal and poloidal locations, allowing well resolved measurements of the global structure. The comparison also highlights the need to include kinetic effects in the MHD model once beta exceeds 80% of the kink mode limit without a conducting wall. Two distinct types of response fields are identified by the linear ideal MHD model: one that consists of localized currents at the rational surfaces that cancel the applied resonant field and another that is excited by the components of the external field that couple to the kink mode. Numerical simulations show these two fields have similar amplitudes in ITER-shaped DIII-D discharges where n = 3 fields are used to suppress edge localized modes.
|
|
| 21. |
- Liu, Yueqiang, 1971, et al.
(författare)
-
Modeling of resistive wall mode and its control in experiments and ITER
- 2006
-
Ingår i: Physics of Plasmas. - : AIP Publishing. - 1089-7674 .- 1070-664X. ; 13:5
-
Tidskriftsartikel (refereegranskat)abstract
- Active control of the resistive wall mode (RWM) for DIII-D [Luxon and Davis, Fusion Technol. 8, 441 (1985)] plasmas is studied using the MARS-F code [Y. Q. Liu, Phys. Plasmas 7, 3681 (2000)]. Control optimization shows that the mode can be stabilized up to the ideal wall beta limit, using the internal control coils (I-coils) and poloidal sensors located at the outboard midplane, in combination with an ideal amplifier. With the present DIII-D power supply model, the stabilization is achieved up to 70% of the range between no-wall and ideal-wall limits. Reasonably good quantitative agreement is achieved between MARS-F simulations and experiments on DIII-D and JET (Joint European Torus) [P. H. Rebut, Nucl. Fusion 25, 1011 (1985)] on critical rotation for the mode stabilization. Dynamics of rotationally stabilized plasmas is well described by a single mode approximation; whilst a strongly unstable plasma requires a multiple mode description. For ITER [R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Controlled Fusion 44, 519 (2002)], the MARS-F simulations show the plasma rotation may not provide a robust mechanism for the RWM stabilization in the advanced scenario. With the assumption of ideal amplifiers, and using optimally tuned controllers and sensor signals, the present feedback coil design in ITER allows stabilization of the n=1 RWM for plasma pressures up to 80% of the range between the no-wall and ideal-wall limits.
|
|
| 22. |
- Liu, Yueqiang, 1971, et al.
(författare)
-
Resistive wall mode control code maturity: progress and specific examples
- 2010
-
Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 1361-6587 .- 0741-3335. ; 52:10
-
Tidskriftsartikel (refereegranskat)abstract
- Two issues of the resistive wall mode (RWM) control code maturity are addressed: the inclusion of advanced mode damping physics beyond the ideal MHD description, and the possibility of taking into account the influence of 3D features of the conducting structures on the mode stability and control. Examples of formulations and computational results are given, using the MARS-F/K codes and the CarMa code. The MARS-K calculations for a DIII-D plasma shows that the fast ion contributions, which can give additional drift kinetic stabilization in the perturbative approach, also drive an extra unstable branch of mode in the self-consistent kinetic modelling. The CarMa modelling for the ITER steady state advanced plasmas shows about 20% reduction in the RWM growth rate by the volumetric blanket modules. The multi-mode analysis predicts a weak interaction between the n = 0 and the n = 1 RWMs, due to the 3D ITER walls. The CarMa code is also successfully applied to model the realistic feedback experiments in RFX.
|
|
| 23. |
- Okabayashi, M., et al.
(författare)
-
Off-axis Fishbone-like Instability and Excitation of Resistive Wall Mode in JT-60U and DIII-D
- 2011
-
Ingår i: Physics of Plasmas. - 1089-7674 .- 1070-664X. ; 18, s. 056121-
-
Tidskriftsartikel (refereegranskat)abstract
- An energetic-particle (EP)-driven “off-axis-fishbone-like mode (OFM)” often triggers a resistive wall mode (RWM) in JT-60U and DIII-D devices, preventing long-duration high-βN discharges. In these experiments, the EPs are energetic ions (70–85 keV) injected by neutral beams to produce high-pressure plasmas. EP-driven bursting events reduce the EP density and the plasma rotation simultaneously. These changes are significant in high-βN low-rotation plasmas, where the RWM stability is predicted to be strongly influenced by the EP precession drift resonance and by the plasma rotation near the q = 2 surface (kinetic effects). Analysis of these effects on stability with a self-consistent perturbation to the mode structure using the MARS-K code showed that the impact of EP losses and rotation drop is sufficient to destabilize the RWM in low-rotation plasmas, when the plasma rotation normalized by Alfvén frequency is only a few tenths of a percent near the q = 2 surface. The OFM characteristics are very similar in JT-60U and DIII-D, including nonlinear mode evolution. The modes grow initially like a classical fishbone, and then the mode structure becomes strongly distorted. The dynamic response of the OFM to an applied n = 1 external field indicates that the mode retains its external kink character. These comparative studies suggest that an energetic particle-driven “off-axis-fishbone-like mode” is a new EP-driven branch of the external kink mode in wall-stabilized plasmas, analogous to the relationship of the classical fishbone branch to the internal kink mode.
|
|
| 24. |
|
|
| 25. |
- Schutte, Katrin, et al.
(författare)
-
Application of the BRAFO tiered approach for benefit-risk assessment to case studies on heat processing contaminants
- 2012
-
Ingår i: Food and Chemical Toxicology. - : Elsevier BV. - 0278-6915. ; 50, s. 724-735
-
Tidskriftsartikel (refereegranskat)abstract
- The aim of the European Funded Project BRAFO (benefit-risk analysis of foods) project was to develop a framework that allows quantitative comparison of human health risks and benefits of foods based on a common scale of measurement. This publication describes the application of the BRAFO methodology to three different case studies: the formation of acrylamide in potato and cereal based products, the formation of benzo(a)pyrene through smoking and grilling of meat and fish and the heat-treatment of milk. Reference, alternative scenario and target population represented the basic structure to test the tiers of the framework. Various intervention methods intended to reduce acrylamide in potato and cereal products were evaluated against the historical production methods. In conclusion the benefits of the acrylamide-reducing measures were considered prevailing. For benzo(a)pyrene, three illustrated alternative scenarios were evaluated against the most common smoking practice. The alternative scenarios were assessed as delivering benefits, introducing only minimal potential risks. Similar considerations were made for heat treatment of milk where the comparison of the microbiological effects of heat treatment, physico-chemical changes of milk constituents with positive and negative health effects was assessed. In general, based on data available, benefits of the heat treatment were outweighing any risks. (C) 2012 ILSI Europe. Published by Elsevier Ltd. All rights reserved.
|
|
| 26. |
- Sheikh, U. A., et al.
(författare)
-
Impact of the new TCV baffled divertor upgrade on pedestal structure and performance
- 2021
-
Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 26
-
Tidskriftsartikel (refereegranskat)abstract
- A new set of carbon tiles, neutral beam heating optics and gas baffles were installed on TCV during the baffled divertor upgrade in early 2019. The installation of the baffles allows a deconvolution of the roles of main chamber and divertor neutral pressure on the H-mode pedestal structure. This physical barrier allows relatively high neutral pressures to be constrained to the divertor, thus preventing neutrals from entering the main chamber and potentially degrading core confinement. This study presents the experimentally measured and modelled pedestal heights and structure for a series of H-mode discharges prior to and after this upgrade. Increased pedestal performance at high divertor neutral pressure was observed after the baffled divertor upgrade. This was consistent across all triangularities and outer target locations investigated and is attributed to higher pedestal top temperatures being maintained at high gas injection rates. ASTRA simulations indicated beam heating power coupled to the plasma did not significantly vary after the baffled divertor upgrade or as a function of divertor neutral gas pressure. Analysis of the pedestal structure exposed a strong correlation between pedestal performance and the density pedestal position prior to and after the baffled divertor upgrade. The baffled divertor upgrade limited the outward shift of the density pedestal, thus maintaining higher pedestal performance at high divertor neutral pressures. Stability analysis indicated the majority of discharges studied were within 25% of the stability boundary. No correlation was found between the distance from the stability boundary and pedestal performance or structure. Comparison with the EPED1 model indicated that TCV discharges do not have a fixed dependence between pedestal.... and pedestal width. A large variation in the EPED1 relating parameter was observed and found to vary with the density pedestal position.
|
|
