Yu. Gribov

Disruption Scenarios, their Mitigation and Operation Window in ITER

The impacts of plasma disruptions on ITER have been investigated in detail to confirm the robustness of the design of the machine to the potential consequential loads. The loads include both electro-magnetic (EM) and heat loads on the in-vessel components and the vacuum vessel. Several representative disruption scenarios are specified based on newly derived physics guidelines for the shortest current quench time as well as the maximum product of halo current fraction and toroidal peaking factor arising from disruptions in ITER. Disruption simulations with the DINA code and EM load analyses with a 3D finite element method code are performed for these scenarios. Some margins are confirmed in the EM load on in-vessel components due to induced eddy and halo currents for these representative scenarios. However, the margins are not very large. The heat load on various parts of the first wall due to the vertical movement and the thermal quench (TQ) is calculated with a 2D heat conduction code based on the database of heat deposition during disruptions and simulation results with the DINA code. For vertical displacement event, it is found that the beryllium (Be) wall does not melt during the vertical movement, prior to the TQ. Significant melting is anticipated for the upper Be wall and the tungsten divertor baffle due to TQ after the vertical movement. However, its impact could be substantially mitigated by implementing a reliable detection system of the vertical movement and a mitigation system, e.g. massive noble gas injection. Some melting of the upper Be wall is anticipated at major disruptions. At least several tens of unmitigated disruptions must be considered even if an advanced prediction/mitigation system is implemented. With these unmitigated disruptions, the loss of the Be layer is expected to be within ?30?100??m/event out of a 10?mm thick Be first wall.

Fourier analysis of 3D error fields in tokamaks

An analysis has been carried out to study error fields anticipated in the ITER machine. The error fields were simulated with the use of a computer code, PRORCODE, developed at the Efremov Institute. The Fourier decomposition of the error fields was performed on equilibrium rational magnetic surfaces inside the plasma filament to evaluate lower modes (m,n) = (1,1), (2,1), (3,1). Major sources of error field associated with misalignments of the magnet system in tokamaks are discussed.

Statistical analysis of expected error fields in tokamaks and their correction

Error fields in tokamak magnet systems are primarily associated with displacements of toroidal field and poloidal field coils, arising during their manufacture and assembly. The Fourier analysis of error fields on a rational magnetic flux surface, anticipated in the ITER machine, exhibited the lower-order modes ((m, n)=(1,1), (2,1), (3,1)) of error fields as the most crucial ones. A statistical study using the Monte Carlo simulations has demonstrated that an averaged three-mode error field from the superposition of coil displacements can be two to three times higher than the acceptable level. The estimates obtained have been used to assess the capability of the correction coil system.

Structural load specification for ITER tokamak components

The substantial mechanical loads which can develop in multiple components are a major technical challenge associated with the design of the ITER tokamak. The various loads acting on ITER can be grouped into several types: inertial loads, associated with gravity and seismic events; pressure loads, particularly significant for the ITER pressure equipment; electromagnetic loads, which affect all conducting structures as a consequence of transient events inducing rapid magnetic field changes and which possibly involve currents flowing between the plasma and in-vessel components; thermal loads, which are extremely severe in the plasma facing components; assembly loads, typically due to preloads imposed during assembly.

Computation technology based on KOMPOT and KLONDIKE codes for magnetostatic simulations in tokamaks

A computation technology has been developed as a tool to provide accurate and detailed magnetostatic simulations for tokamaks. Calculation algorithms are described which allow modelling of complex magnet systems with regard to electromagnetic compatibility of their components and subsystems. The efficiency of the technology has been proved in computations of a typical test blanket module for the International Thermonuclear Experimental Reactor.

Analysis and optimization of the impact of ferromagnetic inserts on the toroidal field ripple in the International Thermonuclear Experimental Reactor

Perturbations of the axial symmetry of a toroidal magnetic field, or the toroidal field (TF) ripple, in tokamaks can potentially lead to losses of high-energy particles from the plasma. These losses can be significant and are one of the major concerns with respect to both the energy blance of the deuterium–tritium plasma and the heat load limits in the International Thermonuclear Experimental Reactor (ITER). In the ITER design, ferromagnetic inserts are used to reduce the TF ripple. In this work the impact of the ferromagnetic inserts on the TF ripple in the ITER is analysed to optimize their filling factor within the design constraints. The analysis of the high-mode spectrum of the amplitudes of ripple high modes is also carried out.

Assessment of current drive efficiency and the synergetic effect for ECCD and LHCD and the possibility of long pulse operation in ITER

Steady state operation is preferable for fusion reactors. The possibility of extending the pulse length in ITER is considered taking into account the capabilities of the planned electron-cyclotron current drive (ECCD) and low-hybrid current drive (LHCD). The ECCD efficiency for current drive at different locations is assessed. The possibility of extending the pulse length by the increase in the current drive efficiency due to the synergetic effect for combined ECCD and LHCD at the same location is assessed. The calculated synergetic effect of ECCD and LHCD on the current drive efficiency is less than 10% for ITER parameters. Long pulse operation with the energy multiplication factor Pfus/Paux = Q > 5 and duration t > 3000 s will be possible in the case of enhanced confinement with respect to the ELMy H-mode scaling HH98y,2 ~ 1.3–1.4.

Assessment of error field from solitary ferromagnetic elements located outside of ITER tokamak

The error fields produced by localized steel objects in the ITER tokamak building have been studied in this paper. The critical mass of the objects, producing in the plasma region the ‘3-mode’ error field of 0.1 Unit (10−6 of the toroidal magnetic field), is estimated for a set of object positions. It is shown that the critical mass is scaled with distance as R 7, if R≥30 m.

Stray magnetic field produced by ITER tokamak complex

The stray magnetic field produced by the ITER tokamak complex, including the effect of ferromagnetic materials in building structures, has been studied. The results obtained show that the magnetic fields produced by the tokamak can be significantly modified by the ferromagnetic structures of the building in areas distant from the tokamak. It is shown that stray fields produced by the ITER tokamak complex can exceed 100 Gs in areas where service staff are possibly located. Such a level of stray fields should be in agreement with the medical and safety engineering limits as well as to ensure the normal operations of equipment sensitive to magnetic fields. The results are presented in the form of a set of field maps, which can be widely used for practical applications.

Dynamic response of the ITER vacuum vessel to electromagnetic loads during VDEs

During vertical displacement events (VDEs) plasma halo currents can flow partly through the passive structure. Additionally induced currents occur in the passive structure. Due to these electrical currents, major electromagnetic forces act on the passive structures and hence on the vacuum vessel (VV). As these forces change in time the vessel response is dynamic. This response determines important design drivers such as the reaction forces at the vessel supports, the vessel displacements and stress levels in the vessel structure, and it affects all components attached to the vessel. It is expected that the most severe dynamic response of the vessel occurs during asymmetric VDEs with slow current quench. Experiments on existing tokamak machines have shown that asymmetric loads can rotate around the vertical machine axis. This possible rotation is considered here. Using the finite element (FE) method the dynamic response of the vessel was analyzed in full transient dynamic analyses for the worst case VDEs according to the ITER VV load specification [2]. A 360° FE model of the VV is used since the loads are partly asymmetric. One major difficulty in this assessment was to predict how the sideways load is shared between three simultaneously acting support types. Attention was therefore given to the modeling of the VV supports including the coupling effect with the toroidal magnetic field.