Divertor Power Load Research Articles

Overview of recent experimental results on MAST

The low aspect ratio of the mega amp spherical tokamak (MAST) allows differentiation between different forms of the H-mode threshold scaling. With optimized fuelling using inboard puffing, and a connected double null divertor (DND) magnetic configuration, the H-mode power threshold data lie about 1.7 times higher than recent scaling laws. Slight magnetic configuration changes, of the order of the ion Larmor radius, around a connected DND significantly influence H-mode access. H-mode confinement in discharges with low frequency edge localized modes (ELMs) is generally consistent with international scaling laws, e.g. IPB98(y,2). Strong indications of both particle and energy internal transport barriers have been seen. Normalized beta values βN > 5 have been obtained, approaching the ideal n = 1 no wall external kink stability limit. Sawtooth triggered neo-classical tearing modes have been observed; numerical modelling of the island evolution reproduces mode behaviour well and confirms the significance of stabilizing field curvature effects. Divertor power loading studies, including transient effects due to ELMs, show a strong bias of power efflux to the outboard targets, where it is more easily handled. ELM energy losses, Δ WELM, are less than 4% of the stored energy in all regimes explored so far, but ELM effluxes extending 30 cm outside the outboard separatrix have been measured. Toroidally asymmetric divertor biasing resulted in significant broadening of the Dα profile on the biased components and a reduction in the total power to the unbiased components. Halo current magnitudes and asymmetries are generally small compared with conventional tokamaks; recent measurements show that the plasma behaves more as a voltage source than a current source. Initial neutral beam current drive experiments indicate non-inductively driven current values (INBI ∼ 0.3Ip) comparable with code predictions.

Next-step-targeted experiments on the Mega-Amp Spherical Tokamak

Since its first physics campaign, the principal parameters on MAST (Mega-Amp Spherical Tokamak) [A. Sykes et al., Nuclear Fusion 41, 1423 (2001)] have been brought up towards their design values. Considerable advances have been made in a range of physics areas of direct relevance to the International Thermonuclear Experimental Reactor (ITER) [ITER Physics Basis, Nuclear Fusion 39, 2175 (1999)]. In this paper, results on H-mode access, global confinement and pedestal studies are presented and compared with conventional aspect ratio tokamak scalings. Physics and engineering requirements relevant to next step spherical tokamak devices are discussed, in particular the plasma formation, current ramp-up and sustainment, and plasma exhaust. Results of first experiments directly targeting these issues are presented: Plasma current up to 0.5 MA has been produced without use of the central solenoid flux, and current ramp-up and sustainment without use of the central solenoid flux has been demonstrated. Experiments on neutral beam heating and current drive (CD) demonstrate up to 50% bootstrap current fraction and good CD efficiency, and divertor power loading has been found to be tolerable and have a favorable outboard asymmetry.

Core plasma operation consistent with SOL parameters in ITER

The constraints of divertor operation have been applied to ITER core plasma simulations by imposing boundary conditions on the calculation, implemented as a set of scaling relations derived from B2-Eirene modelling which describe the effect of the divertor. The core plasma simulations use the integrated core pedestal sol (ICPS) model, based on ITG transport for ions and RLW-like transport for electrons, which includes an increase of transport when the ballooning limit is attained in order to simulate the effect of ELMs in a time-averaged fashion. At the nominal average core density for ITER with the ICPS transport model stationary operation with a fusion power multiplier Q of 12–16 is obtained and a reasonable operating range exists for realistic pumping speeds and particle throughput at a peak divertor power load controlled to remain below 10 MW m−2. Fuelling of the plasma must be predominantly direct core fuelling, and the resulting required core-fuelling rates are realistic. Increased pumping speed and particle throughput are beneficial for maximizing Q. The relaxation to these stationary conditions is very slow, and transient values of fusion multiplier are appreciably higher.

Scaling of ITER divertor parameters – interpolation from 2D modelling and extrapolation

Detailed modelling studies of the divertor plasma for ITER have been carried out. Using these results, scaling relationships are developed linking SOL power, density, throughput, pumping speed, peak divertor power load, and helium density for ITER conditions in order to systematise the results and to extrapolate them beyond the range presently covered by the simulations. The key parameter for the scalings is the neutral pressure in the divertor. Both peak power load and helium density vary as the square of the power at constant pressure. The inclusion of helium elastic collisions reduces the helium density and leads to a steeper reduction with increasing pressure. Variants of the input conditions, i.e. different geometry, no helium elastic collisions, carbon walls, are also discussed, the consistency of the edge modelling with conditions required in the core is treated, and extrapolation to higher power operation is carried out.

First physics results from the MAST Mega-Amp Spherical Tokamak

First physics results are presented from MAST (Mega-Amp Spherical Tokamak), one of the new generation of purpose built spherical tokamaks (STs) now commencing operation. Some of these results demonstrate, for the first time, the novel effects of low aspect ratio, for example, the enhancement of resistivity due to neo-classical effects. H-mode is achieved and the transition to H-mode is accompanied by a tenfold steepening of the edge density gradient which may enable the successful application of electron Bernstein wave heating in STs. Studies of halo currents show that these less than expected from conventional tokamak results, and measurements of divertor power loading confirm that most of the power flows to the outer strike points, easing the power handling on the inner points (a critical issue for STs).

Asymmetries in the divertor power loading in START

Langmuir probe measurements from the divertor strike points for a series of very similar, highly up-down magnetically symmetric double-null divertor discharges are presented. These show a strong in–out electron density ( n e) asymmetry, as well as up-down asymmetries in the electron temperature ( T e), the power density scale length ( λ q ) and n e, on the inboard side. These up-down asymmetries seem to be due to particle drift effects and not due to an imbalance in the magnetic X-points. It is estimated that only about 35% of the power that enters the scrape-off layer (SOL) reaches the divertor tiles. There is a high neutral density in the START SOL and it is conjectured that the bulk of this missing power is lost through charge exchange, or elastic collisions, of the ions with these neutrals as they flow towards the divertor tiles. The highest power density ( q), of about 4–8 MW/m 2, is found to be at the upper outboard strike point, not the inboard, because the major radii of the inboard and outboard strike points is similar and n e is much higher on the outboard side.

Studies in JET divertors of varied geometry. I: Non-seeded plasma operation

Results of experiments investigating the performance of the JET Mark IIA divertor are reported and compared with the performance of its Mark I predecessor. The principal effect of reducing the divertor width (increasing closure) was to increase pumping for both deuterium and impurities while reducing upstream neutral pressure. Neither the orientation of the divertor target relative to the divertor plasma nor the width of the divertor had a major influence on core plasma performance in ELMy H modes. Changing the core triangularity and thus the edge magnetic shear modifies the ELM frequency in ELMy H mode plasmas, thereby changing the peak divertor power loading. The integrated performance of the core and divertor plasmas is reviewed with a view to extrapolation to the requirements of ITER. The confinement of JET ELMy H modes with hot, medium density edges is good (H97 ≈ 1) and follows a gyro-Bohm scaling. The impurity content of these discharges is low and within the ITER requirements. When an attempt is made to raise the density with deuterium gas fuelling, the ELM frequency increases and the confinement, especially in the edge, decreases. Good confinement can be achieved in JET either by producing a large edge pedestal, typically in discharges with NB heating or by centrally peaked heating with ICRH schemes. Large amplitude type I ELMs, which are present in all discharges with a large edge pedestal, would result in unacceptable divertor plate erosion when scaled to ITER. Since the power deposition profile due to α heating in ITER is calculated to be intermediate between the JET NB and RF heating profiles, it is likely that operation in ITER with small ELMs in order to reduce first wall loading will result in degraded confinement compared with present day scaling laws.

Studies in JET divertors of varied geometry. II: Impurity seeded plasmas

In current large tokamaks, non-intrinsic seeded impurities have been used to produce divertor power loads which would be considered acceptable when extrapolated to ITER. Many devices have achieved the goals of high fractional radiated powers, small frequent ELMs and detachment which are characteristic of radiative H mode regimes. The influence of divertor geometry on these characteristics is described. It has been a matter of concern that the Zeff associated with the seeded impurities may exceed that allowable in ITER and also that the degradation in energy confinement may be unacceptable. Confidence can only be built in the prediction of these parameters in ITER if reliable scalings are available for impurity content and energy confinement which have a sound physics basis. Work is described at JET in this area whilst using multimachine data to characterize the size scaling and provide a context for the JET data. Predicted levels for the impurity content of seeded ITER plasmas appear to be of marginal acceptability. Discharges run in the JET Mark I, Mark IIA and Mark IIAP divertors are compared and indicate that increased divertor closure has brought relatively minor benefits in highly radiative discharges. The acceptability of the energy confinement of radiation for ITER remains unclear. Dimensionless parameter scaling experiments have been conducted in which β, q25, fractional radiated power and Zeff are held constant for a range of ρ*. The price paid for high edge radiation and small ELMs appears to be a 25% loss in total stored energy as a result of edge pedestal degradation. However, the underlying energy confinement scaling may still be consistent with gyro-Bohm scaling, which would give an adequate margin for ITER. This conclusion is, however, sensitive to the scaling of confinement with collisionality, which is difficult to determine due to the coupling between ρ* and ν* which is a consequence of radiation dominated regimes.

Compatibility of improved confinement regimes with reactor requirements

The reactor physics requirements are outlined in terms of characteristic performance measures. Four reactor designs of different scope are taken as examples. ITER is based on the ELMy H-mode as an operating regime demonstrated over many energy confinement times. Designs for power producing reactors require higher power densities and bootstrap current fractions, and therefore higher values of normalized . The capability of different enhanced confinement regimes, combining internal transport barriers with the L- or H-mode edge to satisfy these demands, together with the requirements on helium pumping and divertor power load, is discussed.

A D-3He fusion reactor with an edge radiating layer

The possibility of operations with advanced fuels in high temperature, high-β plasmas is investigated. At high β values, diamagnetic effects reduce the toroidal field strength in the plasma core and tend to localize the synchrotron radiation source in the outer part of the plasma column. Thus, the core energy balance is not affected by synchrotron radiation losses, but a large amount of the fusion power is radiated by the edge and the divertor power load is decreased. A small fraction of the radiated power can be recirculated for non-inductive current drive, to compensate for the large amount of bootstrap current, naturally present in such configurations. In this scheme, synchrotron radiation is also used to recover the fusion power, by direct conversion to electric power.

Scaling radiative plasmas to ITER

In the radiative regime non-intrinsic impurities have been used to produce divertor power loads which would be considered acceptable when extrapolated to ITER. However, it has been a matter of concern that the maximum Z eff currently deemed acceptable for ITER has been exceeded by a large margin in radiative plasma experiments in large divertor machines such as JET, JT60-U, AUG and DIII-D. This paper points out that without a suitable scaling law, comparisons of Z eff between current machines and ITER are meaningless. Results from a multi-machine database are presented which show that there appears to be a remarkably simple and robust scaling which relates Z eff to line averaged density, total radiated power and main plasma surface area. A similar scaling has been found in code simulations with EDGE2D and DIVIMP. The consequences for ITER are discussed.

Gas target and detached plasma regimes for the ITER divertor

A number of solutions to the problem of the ITER divertor power loading have been proposed. The most promising ones are those characterized by an ionization front which is detached from the divertor plate. The space between such a front and the plate is filled with a cold, T e < 1 eV, and (relatively) low density plasma, n e < 10 14cm −3. A regular high recycling regime and a detached (gas target) regime have been simulated with the PLANET code at a high scrape-off power of 220 MW. It is found that, in the detached case, over 90% of this power can be converted into radiation and that the gas target is well confined, but the midplane density is rather high, n e ≅ 1 × 10 14cm −3.

Power radiated from ITER by impurities

The MIST code has been used to model impurity radiation from the edge and core plasmas in ITER. A broad range of parameters have been varied, including Z eff, impurity species, impurity transport coefficients, and plasma temperature and density profiles. For a set of these parameters representative of the baseline ITER ignition scenario, it is seen that impurity radiation can make a major reduction in divertor power load without compromising core energy confinement. Scalings of impurity radiation with atomic number and machine size are also discussed.

Two-dimensional fluid simulations of the ITER SOL plasma

Simulations of scrape-off layer plasmas in ITER have been performed using a 2-d fluid model with a set of perpendicular energy and particle transport coefficients matching those determined in a limited number of experiments on presently operating tokamaks. These simulations include single- and double-null magnetic configurations, and ignited and driven operation. Predictions are made of divertor power loading, density scrape-off length, effective core/edge recycling coefficient, and divertor sheath electron temperature.