A self-consistent analysis of the relevant physics regarding the formation and termination of runaway beams during mitigated disruptions by Ar and Ne injection is presented for selected ITER ...scenarios with the aim of improving our understanding of the physics underlying the runaway heat loads onto the plasma facing components (PFCs) and identifying open issues for developing and accessing disruption mitigation schemes for ITER. This is carried out by means of simplified models, but still retaining sufficient details of the key physical processes, including: (a) the expected dominant runaway generation mechanisms (avalanche and primary runaway seeds: Dreicer and hot tail runaway generation, tritium decay and Compton scattering of γ rays emitted by the activated wall), (b) effects associated with the plasma and runaway current density profile shape, and (c) corrections to the runaway dynamics to account for the collisions of the runaways with the partially stripped impurity ions, which are found to have strong effects leading to low runaway current generation and low energy conversion during current termination for mitigated disruptions by noble gas injection (particularly for Ne injection) for the shortest current quench times compatible with acceptable forces on the ITER vessel and in-vessel components (τres∼22 ms). For the case of long current quench times (τres∼66 ms), runaway beams up to 10 MA can be generated during the disruption current quench and, if the termination of the runaway current is slow enough, the generation of runaways by the avalanche mechanism can play an important role, increasing substantially the energy deposited by the runaways onto the PFCs up to a few hundreds of MJs. Mixed impurity (Ar or Ne) plus deuterium injection proves to be effective in controlling the formation of the runaway current during the current quench, even for the longest current quench times, as well as in decreasing the energy deposited on the runaway electrons during current termination.
Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active ...low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of ∼2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix. Therefore, specifications for the rotation of the 3D perturbation applied for ELM control in order to avoid excessive localized erosion of the ITER divertor target are derived. It is shown that a rotation frequency in excess of 1 Hz for the whole toroidally asymmetric divertor power flux pattern is required (corresponding to n Hz frequency in the variation of currents in the coils, where n is the toroidal symmetry of the perturbation applied) in order to avoid unacceptable thermal cycling of the divertor target for the highest power fluxes and worst toroidal power flux asymmetries expected. The possible use of the in-vessel vertical stability coils for ELM control as a back-up to the main ELM control systems in ITER is described and the feasibility of its application to control ELMs in low plasma current H-modes, foreseen for initial ITER operation, is evaluated and found to be viable for plasma currents up to 5-10 MA depending on modelling assumptions.
High-performance and long-pulse operation is a crucial goal of current magnetic fusion research. Here, we demonstrate a high-connement plasma regime known as an H-mode with a record pulse length of ...over 30 s in the Experimental Advanced Superconducting Tokamak sustained by lower hybrid wave current drive (LHCD) with advanced lithium wall conditioning. We nd that LHCD provides a exible boundary control for a ubiquitous edge instability in H-mode plasmas known as an edge-localized mode, which leads to a marked reduction in the heat load on the vessel wall compared with standard edge-localized modes. LHCD also induces edge plasma ergodization that broadens the heat deposition footprint. The heat transport caused by this ergodization can be actively controlled by regulating the edge plasma conditions. This potentially offers a new means for heat-ux control, which is a key issue for next-step fusion development. PUBLICATION ABSTRACT
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IJS, IZUM, KILJ, NUK, PILJ, PNG, SAZU, UL, UM, UPUK
In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible ...with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today’s devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
The XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance ...is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates. The simulation results compare well against the empirical scaling λq ∝ 1/BPγ obtained from present tokamak devices, where λq is the divertor heat-flux width mapped to the outboard midplane, γ = 1.19 as found by Eich et al (2013 Nucl. Fusion 53 093031), and BP is the magnitude of the poloidal magnetic field at the outboard midplane separatrix surface. This empirical scaling predicts λq 1 mm when extrapolated to ITER, which would require operation with very high separatrix densities (nsep/nGreenwald > 0.6) (Kukushkin et al 2013 J. Nucl. Mater. 438 S203) in the Q = 10 scenario to achieve semi-detached plasma operation and high radiative fractions for acceptable divertor power fluxes. Using the same simulation code and technique, however, the projected λq for ITER's model plasma is 5.9 mm, which could be suggesting that operation in the ITER Q = 10 scenario with acceptable divertor power loads may be obtained over a wider range of plasma separatrix densities and radiative fractions. The physics reason behind this difference is, according to the XGC1 results, that while the ion magnetic drift contribution to the divertor heat-flux width is wider in the present tokamaks, the turbulent electron contribution is wider in ITER. Study will continue to verify further this important projection. A high current C-Mod discharge is found to be in a mixed regime: While the heat-flux width by the ion neoclassical magnetic drift is still wider than the turbulent electron heat-flux width, the heat-flux magnitude is dominated by the narrower electron heat-flux.
Plasma wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper, different aspects of the PWI are assessed ...in their importance for the initial wall materials choice: CFC for the strike point tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor/Be first wall and all-W or all-C. One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a
Q
=
10 ITER discharge G. Federici et al., J. Nucl. Mater. 290–293 (2001) 260 which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4
±
3 higher, this margin has been adopted as uncertainty of the scaling. With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated: It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5
MJ. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated. For low-
Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms. For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account. Finally, the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
The neoclassical transport of tungsten in the H-mode edge transport barrier (ETB) of ITER is investigated and scaling relations for the transport coefficients are evaluated. The major finding is that ...the radial convection velocity of tungsten is outward directed for a large proportion of the tested pedestal profiles. This is due to a combination of the high pedestal temperatures and the high separatrix densities making the outward directed temperature screening term the predominant contribution of the collisional convection. The high densities at the separatrix are needed to control the power exhaust and the tungsten sputtering in the divertor and the high pedestal temperatures are expected to be achieved in ITER to meet its fusion performance objectives. The density and temperature profiles, which deliver simultaneously optimum fusion performance, power exhaust and tungsten sputtering control in ITER, also ensure optimum collisional radial transport of tungsten leading to hollow tungsten density profiles in the ETB. In helium H-mode plasmas, which are considered for the initial ITER non-active operation, this very favourable neoclassical transport effect is reduced and only about half of the considered edge plasma profiles yield an outward directed drift.
A review of the main concepts, proposed solutions and results of R&D as well as outstanding issues on power exhaust in tokamaks is presented with specific emphasis on the expected issues that need to ...be resolved for ITER and future fusion reactors such as DEMO for both conventional and advanced divertor concepts.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK, ZRSKP
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