Yu. Igitkhanov

The ITER divertor concept

Abstract The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions Te in the divertor can be reduced to

The Helias reactor HSR4/18

The Helias reactor is an upgraded version of the Wendelstein 7-X experiment. A straightforward extrapolation of Wendelstein 7-X leads to HSR5/22, which has 5 field periods and a major radius of 22 m. HSR4/18 is a more compact Helias reactor with 4 field periods and an 18 m major radius. Stability limit and energy confinement times are nearly the same as in HSR5/22, thus the same fusion power (3000 MW) is expected in both configurations. Neoclassical transport in HSR4/18 is very low, and the effective helical ripple is below 1%. The article describes the power balance of the Helias reactor, and the blanket and maintenance concepts. The coil system of HSR4/18 comprises 40 modular coils with NbTi superconducting cables. The reduction from 5 to 4 field periods and the concomitant reduction in size will also reduce the cost of the Helias reactor.

Lifetime of Be-, CFC- and W-Armoured ITER Divertor Plates

Abstract The thermal performance of divertor plates armoured with beryllium (thickness 2–30 mm), carbon fibre composite (CFC, 2–60 mm) or tungsten (W3Re, 2–60 mm) was modelled for transients of 10 s at 20 MW/m 2 — with and without vapour shielding. Due to the high resulting temperatures special attention was paid to the material properties up to and above melting or sublimation. Pure copper was chosen for the flat heat sink of Be and W3Re, the coolant tube material was DS-Cu in all cases. The melt layer depth, the surface temperature and peak power densities to the coolant were obtained in full 2D geometry using finite element modelling. The code was modified to include melting. Evaporation and vapour shielding at elevated temperatures improve the cooling efficiency and reduce the incident power (converting it to radiation, some of which is radiated elsewhere) and thus reduce the depth of melting. For Be armour 10 mm thick, slow plasma transients at 20 MW/m 2 after 10 s will induce melting to a depth of ∼ 2 mm without vapour shielding, and ∼ 0.15 mm with vapour shielding. For up to 60 mm of CFC the peak temperatures attained stay below sublimation; for up to 20 mm of W3Re the surface temperature stay below melting. The lifetime of the divertor plates is determined from the reduction of thickness due to these transients (10% of shots, 10 s), as well as from disruptions (10% of shots) and sputtering. When the maximum or initial armour thickness is given either by the maximum permitted surface temperature during steady state operation at 5 MW/m 2 (which is 1050 K for Be, permitting 11 mm armour thickness, 1780 K for carbon, permitting 40 mm of CFC after neutron irradiation), or in the case of W3Re by the requirement to stay below melting after 10 s at 20 MW/m 2 (permitting 20 mm), and assuming a minimum remaining thickness of 2 mm for all three armour materials, the predicted lifetimes without vapour shielding during transients are as follows: 120–225 shots for Be, 5840–8170 shots for CFC with moderate chemical sputtering, and 2370–7740 shots for W3Re.

Erosion lifetime of ITER divertor plates

Abstract For ITER divertor operation in the detached regime, erosion of divertor plates made of beryllium, carbon fibre composites and tungsten-rhenium alloys is evaluated. Erosion due to (a) physical and, for CFC, chemical sputtering (including surface temperature variation over the lifetime because of neutron irradiation), (b) disruption thermal quench using published data and (c) evaporation and melt layer loss (metals) for infrequent high-power transients (20 MW/m 2 , up to 10 s) is evaluated both with and without the reduction of net incident power by radiation from the evaporated impurities (‘low density vapour shield’). The composite lifetime is calculated for normal operation (detached, 10% transients, 10% disruptions); the effect of an increase in transient frequency and in incident power (semi-attached operation) is determined. The resulting erosion lifetime excludes beryllium, whereas both CFC and W-3%Re yield acceptable lifetimes.

Understanding of H mode pedestal characteristics using the multimachine pedestal database

With use of a multimachine pedestal database, essential issues for each regime of ELM types are investigated. They include (i) understanding and prediction of pedestal pressure during type I ELMs, a reference operation mode of a future tokamak reactor; (ii) identification of the operation regime of type II ELMs, which have small ELM amplitude with good confinement characteristics; (iii) identification of the upper stability boundary of type III ELMs for access to the higher confinement regimes with type I or II ELMs; (iv) understanding the relation between core confinement and pedestal temperature in conjunction with the confinement degradation in high density discharges. Both scaling and model based approaches for expressing pedestal pressure are shown to roughly scale the experimental data well and could be used to make initial predictions for a future reactor. q and delta are identified as important parameters for obtaining the type II ELM regime. A theoretical model of type III ELMs is shown to reproduce the upper stability boundary reasonably well. It is shown that there exists some critical pedestal temperature below which the core confinement starts to degrade. It is also shown that it is possible to obtain improved pedestal conditions for good confinement in high density discharges by increasing the plasma triangularity.

Requirements for pellet injection in ITER scenarios with enhanced particle confinement

The requirements for pellet injection parameters for plasma fuelling are assessed for ITER scenarios with enhanced particle confinement. The assessment is based on the integrated transport simulations including models of pedestal transport, reduction of helium transport and boundary conditions compatible with SOL/divertor simulations. The requirements for pellet injection for the inductive H-mode scenario (HH98y,2 = 1) are reconsidered taking account of a possible reduction of the particle loss obtained in some experiments at low collisionalities. The assessment of fuelling requirements is carried out for the hybrid and steady state (SS) scenarios with enhanced confinement with HH98y,2 > 1. A robustness of plasma performance to the variation of particle transport is demonstrated. A new type of SS scenario is considered with neutral beam current drive and electron cyclotron current drive instead of lower hybrid current drive (LHCD) to extend the range of stable operation and to avoid the reduction of the edge LHCD efficiency caused by pellet injection.

Transition dynamics and confinement scaling in COMPASS-D H mode plasmas

An experimental analysis is discussed of different phases of ECRH H?mode discharges on COMPASS-D, from L-H transition triggering to steady state. Comprehensive high resolution measurements in the transport barrier region have enabled significant progress to be made in assessing H?mode transition models, together with additional experiments aimed at resolving controlling trigger mechanisms. The application of four possible transition models to the data indicates that all demonstrate limiting or critical parameter values at the transition. One of them, based on Alfv?n drift wave turbulence suppression, exhibits precursor behaviour in the stability regime diagram and predicts the density and magnetic field dependences of the H?mode power threshold observed on COMPASS-D. Analysis of the evolution of the local radial electric field and its shear indicates that significant shear only develops after the L-H transition. Stationary ELMy H?modes are achieved at high ?* with good confinement, and dimensionless scaling over a range of ?* has been carried out, providing valuable confinement data in a regime where heat is deposited primarily to the electrons.

Analysis of the Magnetic Field Perturbations During the Assembly of Wendelstein 7-X

Abstract The magnetic configurations of the Wendelstein 7-X (W7-X) stellarator are sensitive to perturbations of the magnetic field resonant with ι/2π = 1. Such perturbations can be caused by deviations of the current filament positions of the real coil system from the design due to the accuracy achievable during the manufacture of the coils and assembly of the magnet system. The sensitivity of the magnetic field to the different types of error has been investigated by introducing randomly distributed errors to the coil shapes and positions within the given tolerances. A statistical analysis of these error distributions was performed. This procedure will be used to assess the magnetic configuration of W7-X before the completion of each assembly step.

Concept of a Helias ignition experiment

The Helias ignition experiment is an upgraded version of the Wendelstein 7-X experiment. The magnetic configuration is a four-period Helias configuration (major radius 18 m, plasma radius 2.0 m, B = 4.5 T), which presents a more compact option than the five-period configuration. Much effort has been focused on two versions of the four-period configuration. One option is the power reactor HSR4/18 providing at least 3 GW of fusion power and the second is the ignition experiment HSR 4/18i aiming at a minimum of fusion power and the demonstration of self-sustaining burn. The design criteria of the ignition experiment HSR 4/18i are the following: The experiment should demonstrate a safe and reliable route to ignition; self-sustained burn without external heating; steady-state operation during several hundred seconds; reliability of the technical components and tritium breeding in a test blanket. The paper discusses the technical issues of the coil system and describes the vacuum vessel and the shielding blanket. The power balance will be modelled with a transport code and the ignition conditions will be investigated using current scaling laws of energy confinement in stellarators. The plasma parameters of the ignition experiment are: peak density 2–3×1020 m−3, peak temperature 11–15 keV, average beta 3.6% and fusion power 1500–1700 MW.

A 1-D Predictive Model for Energy and Particle Transport in H-Mode

In the last few years, significant progress has been made in the understanding of H-mode plasmas (e.g. ion temperature profile stiffness, pedestal physics, etc). Based on this improved understanding, a set of rules (models) comprising a physics picture of the H-mode has been implemented in the ASTRA code in order to improve the understanding of experimental observations and ultimately to provide a predictive capability for ITER complementary to the scaling relations. The model has been verified for consistency with experimental observations in ASDEX-UG and JET plasmas. Numerical coefficients for the transport, required because of simplifications or missing quantitative information, are determined for one plasma (e.g. from JET) and then held constant for all others (JET, ASDEX-UP, ITER). After benchmarking the model to experimental results, it was also applied to ITER. It predicts that Q = 10 can be achieved in ITER but only with at least a 50% deep fuelling contribution (inside the H-mode pedestal). However, in existing machines as well as in our model runs for existing machines, gas puffing is sufficient to achieve the observed density pedestal and line average densities. A second important result from the predictive runs for ITER is that electron energy transport in the plasma core, the neoclassical transport in the pedestal and the CX losses at the plasma edge are important constraints for a better performance. Thus future theoretical and experimental work should concentrate on these areas in order to improve our predictions.