G. F. Matthews

Chapter 4: Power and particle control

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137–2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma–materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral–neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma–materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

Plasma detachment in JET Mark I divertor experiments

The experimental characteristics of divertor detachment in the JET tokamak with the Mark?I pumped divertor are presented for ohmic, L?mode and ELMy H?mode experiments with the main emphasis on discharges with deuterium fuelling only. The range over which divertor detachment is observed for the various regimes, as well as the influence of divertor configuration, direction of the toroidal field, divertor target material and active pumping on detachment, will be described. The observed detachment characteristics, such as the existence of a considerable electron pressure drop along the field lines in the scrape-off layer (SOL), and the compatibility of the decrease in plasma flux to the divertor plate with the observed increase of neutral pressure and D? emission from the divertor region, will be examined in the light of existing results from analytical and numerical models for plasma detachment. Finally, a method to evaluate the degree of detachment and the window of detachment is proposed, and all the observations of the JET Mark?I divertor experiments are summarized in the light of this new quantitative definition of divertor detachment.

Overview of the ITER-like wall project

Work is in progress to completely replace, in 2008/9, the existing JET CFC tiles with a configuration of plasma facing materials consistent with the ITER design. The ITER-like wall (ILW) will be cr ...

Overview of material re-deposition and fuel retention studies at JET with the Gas Box divertor

in the period 1998-2001 the JET tokamak was operated with the MkII Gas Box divertor. On two occasions during that period a number of limiter and divertor tiles were retrieved from the torus and the ...

Erosion/deposition in JET during the period 1999–2001

Coated divertor and wall tiles exposed in JET for the 1999-2001 operations have been used to assess erosion/deposition. Deposited films of up to 90 mum thickness at the inner wall of the divertor t ...

A comparison of experimental measurements and code results to determine flows in the JET SOL

Two reciprocating probe systems, at the same poloidal position at the top of the JET torus but toroidally separated by 180degrees, have been used to measure parallel flow in the scrape-off layer (SOL) of lower single-null, diverted plasmas. One system uses the entrance slit plates of a retarding field analyser to record upstream and downstream flux densities, whilst the second employs two pins of a nine-pin turbulent transport probe. Measurements have been made for both forward and reversed toroidal field directions. The results from both probe systems are similar. In the forward field direction, that is with the ion B x del(B) over right arrow drift direction downwards towards the divertor, a strong parallel flow is measured at the top of the machine in the direction from the outer to the inner divertor. The flow generally has a low value, Mach number M similar to 0.2, close to the separatrix, but rises in the region of high magnetic shear close to the separatrix to a maximum of M similar to 0.5 some 20 mm outside the separatrix. In contrast, for a reversed field, the measured flow is small (close to zero) throughout much of the SOL but rises near the separatrix to a value equal in both magnitude and direction to that observed in the forward field. There is thus some symmetry in the flow with respect to field reversal but with a symmetry axis given by a positive offset of around M similar to 0.2. This paper presents simulations using the EDGE2D/Nimbus code, which predicts very low values of parallel flow Mach number near the probe position. The possibility of impurities released from the probe surfaces increasing the flow velocity is explored using the code.

Retarding field energy analyzer for the JET plasma boundary

Retarding field analyzers (RFAs) are essentially the only practical tool with which to measure the distribution of ion energies in the boundary plasma of magnetic fusion devices. The technique has long been attempted in such facilities with varying degrees of success, but, by virtue of the delicate nature of these material probes and the hostile environment in which they are inserted, its use has been limited to smaller plasma machines with easier access and more tolerable plasma conditions. This article describes in detail a new RFA probe head recently constructed and tested on the large JET tokamak facility and now being used for plasma physics studies. Emphasis is placed on the details of design and mechanical construction as they relate to the particularly harsh conditions imposed both by the JET boundary plasma and the requirement that the probe be used to sample this plasma by insertion via a multiple plunge, fast reciprocating drive system. Some preliminary physics results are also included demonstrating both successful operation of the new device and discussing its limitations.

ELM energy and particle losses and their extrapolation to burning plasma experiments

Abstract Analysis of Type I ELMs from present experiments shows that ELM energy losses decrease with increasing pedestal plasma collisionality ( ν ∗ ped ) and/or increasing τ Front ∥ , where ( τ ∥ Front =2π Rq 95 / c s ,ped ) is the typical ion transport time from the pedestal to the divertor target. ν ∗ ped and τ Front ∥ are not the only parameters that affect the ELMs, also the edge magnetic shear influences the plasma volume affected by the ELMs. ELM particle losses are influenced by this ELM affected volume and are weakly dependent on other pedestal plasma parameters. ‘Minimum’ Type I ELMs, with energy losses acceptable for ITER, where there is no change in the plasma temperature profile during the ELM, are observed for some conditions in JET and DIII-D. The duration of the divertor ELM power pulse is well correlated with τ Front ∥ and not with the duration of the ELM-associated MHD activity. Similarly, the time scale of ELM particle fluxes is also determined by τ Front ∥ . The extrapolation of present experimental results to ITER is summarised.

Investigation of the fluxes to a surface at grazing angles of incidence in the tokamak boundary

The paper describes an experimental investigation of the effects of grazing angles of incidence on the fluxes of power and particles to a surface inserted into the boundary of the DITE tokamak. Related thermographic data from the TFTR tokamak are also presented. A summary is given of the important issues relating to angle of incidence which are relevant to large tokamaks such as JET, NET and ITER and which require experimental confirmation. An array of Langmuir probes has been constructed to carry out experiments as a function of angle on DITE. Results from this tilting probe array (TPA) are described, which show how the ion saturation current, electron saturation current, floating potential and the fitted electron temperature vary with surface angle. It is shown that when the magnetic field is at grazing angles of incidence to the surface a conventional interpretation of the probe characteristics fails completely. Power flux distributions from infra-red thermography of the TFTR moveable limiter are presented, which show a substantially higher heat flux at the tangency point than is calculated conventionally.

Impact and mitigation of disruptions with the ITER-like wall in JET

Disruptions are a critical issue for ITER because of the high thermal and magnetic energies that are released on short timescales, which results in extreme forces and heat loads. The choice of material of the plasma-facing components (PFCs) can have significant impact on the loads that arise during a disruption. With the ITER-like wall (ILW) in JET made of beryllium in the main chamber and tungsten in the divertor, the main finding is a low fraction of radiation. This has dropped significantly with the ILW from 50?100% of the total energy being dissipated during disruptions in CFC wall plasmas, to less than 50% on average and down to just 10% for vertical displacement events (VDEs). All other changes in disruption properties and loads are consequences of this low radiation: long current quenches (CQs), high vessel forces caused by halo currents and toroidal current asymmetries as well as severe heat loads. Temperatures close to the melting limit have been locally observed on upper first wall structures during deliberate VDE and even at plasma currents as low as 1.5 MA and thermal energy of about 1.5?MJ only. A high radiation fraction can be regained by massive injection of a mixture of 10% Ar with 90% D2. This accelerates the CQ thus reducing the halo current and sideways impulse. The temperature of PFCs stays below 400??C. MGI is now a mandatory tool to mitigate disruptions in closed-loop operation for currents at and above 2.5?MA in JET.