D.G. Whyte

Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in todays tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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.

Chapter 3: MHD stability, operational limits and disruptions

Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

Transport by intermittent convection in the boundary of the DIII-D tokamak

Intermittent plasma objects (IPOs) featuring higher pressure than the surrounding plasma, and responsible for ∼50% of the E×BT radial transport, are observed in the scrape off layer (SOL) and edge of the DIII-D tokamak [J. Watkins et al., Rev. Sci. Instrum. 63, 4728 (1992)]. Conditional averaging reveals that the IPOs, produced at a rate of ∼3×103 s−1, are positively charged and also polarized, featuring poloidal electric fields of up to 4000 V/m. The IPOs move poloidally at speeds of up to 5000 m/s and radially with E×BT/B2 velocities of ∼2600 m/s near the last closed flux surface (LCFS), and ∼330 m/s near the wall. The IPOs slow down as they shrink in radial size from 4 cm at the LCFS to 0.5 cm near the wall. The IPOs appear in the SOL of both L and H mode discharges and are responsible for nearly 50% of the SOL radial E×B transport at all radii; however, they are highly reduced in absolute amplitude in H-mode conditions.

Transport by intermittency in the boundary of the DIII-D tokamak

A271 TRANSPORT BY INTERMITTENCY IN THE BOUNDARY OF THE DIII-D TOKAMAK. Intermittent plasma objectives (IPOs) featuring higher pressure than the surrounding plasma, are responsible for {approx} 50% of the E x B{sub T} radial transport in the scrape off layer (SOL) of the DIII-D tokamak in L- and H-mode discharges. Conditional averaging reveals that the IPOs are positively charged and feature internal poloidal electric fields of up to 4000 V/m. The IPOs move radially with E x B{sub T}/B{sup 2} velocities of {approx} 2600 m/s near the last closed flux surface (LCFS), and {approx} 330 m/s near the wall. The IPOs slow down as they shrink in radial size from 4 cm at the LCFS to 0.5 cm near the wall. The skewness (i.e. asymmetry of fluctuations from the average) of probe and beam emission spectroscopy (BES) data indicate IPO formation at or near the LCFS and the existence of positive and negative IPOs which move in opposite directions. The particle content of the IPOs at the LCFS is linearly dependent on the local density and decays over {approx} 3 cm into the SOL while their temperature decays much faster ({approx} 1 cm).

Sputtering yield measurements during low energy xenon plasma bombardment

The sputtering yields of molybdenum, titanium, beryllium, and carbon have been measured during xenon ion bombardment from a plasma in the energy range between 10 and 200 eV. The erosion rates of Mo, Be, and C are measured both spectroscopically in the plasma and using the standard weight loss technique. Spectroscopic measurements of Ti sputtering yields, where no atomic physics data is available, are normalized to the weight loss measurements. The erosion rates of the metals decrease with the reduced mass of the metal–xenon combination and decrease with the increasing metal’s binding energy, as expected. The erosion results for bombardment of graphite indicate that the sputtering rate of carbon, as atoms, from the surface is insufficient to explain the total carbon weight loss measured. The multiple mechanisms for carbon erosion during plasma bombardment are discussed and the sputter rates of carbon atoms and carbon dimers are presented.

I-mode: an H-mode energy confinement regime with L-mode particle transport in Alcator C-Mod

An improved energy confinement regime, I-mode, is studied in Alcator C-Mod, a compact high-field divertor tokamak using ion cyclotron range of frequencies (ICRFs) auxiliary heating. I-mode features an edge energy transport barrier without an accompanying particle barrier, leading to several performance benefits. H-mode energy confinement is obtained without core impurity accumulation, resulting in reduced impurity radiation with a high-Z metal wall and ICRF heating. I-mode has a stationary temperature pedestal with edge localized modes typically absent, while plasma density is controlled using divertor cryopumping. I-mode is a confinement regime that appears distinct from both L-mode and H-mode, combining the most favourable elements of both. The I-mode regime is investigated predominately with ion ∇B drift away from the active X-point. The transition from L-mode to I-mode is primarily identified by the formation of a high temperature edge pedestal, while the edge density profile remains nearly identical to L-mode. Laser blowoff injection shows that I-mode core impurity confinement times are nearly identical with those in L-mode, despite the enhanced energy confinement. In addition, a weakly coherent edge MHD mode is apparent at high frequency ~100–300 kHz which appears to increase particle transport in the edge. The I-mode regime has been obtained over a wide parameter space (BT = 3–6 T, Ip = 0.7–1.3 MA, q95 = 2.5–5). In general, the I-mode exhibits the strongest edge temperature pedestal (Tped) and normalized energy confinement (H98 > 1) at low q95 ( 4 MW). I-mode significantly expands the operational space of edge localized mode (ELM)-free, stationary pedestals in C-Mod to Tped ~ 1 keV and low collisionality , as compared with EDA H-mode with Tped . The I-mode global energy confinement has a relatively weak degradation with heating power; leading to increasing H98 with heating power.

Fluctuation-driven transport in the DIII-D boundary

Cross-field fluctuation-driven transport is studied in edge and scrape-off layer (SOL) plasmas in the DIII-D tokamak using a fast reciprocating Langmuir probe array allowing local measurements of the fluctuation-driven particle and heat fluxes. Two different non-diffusive mechanisms that can contribute strongly to the cross-field transport in the SOL of high-density discharges are identified and compared. The first of these involves intermittent transport events that are observed at the plasma separatrix and in the SOL. Intermittence has qualitatively similar character in L-mode and ELM-free H-mode. Low-amplitude ELMs observed in high-density H-mode produce in the SOL periods with cross-field transport enhanced to L-mode levels and featuring intermittent events similar to those in L-mode. The intermittent transport events are compatible with the concept of plasma filaments propagating across the SOL due to E×B drifts. The intermittent character of the transport in the SOL is also in agreement with predictions of the non-linear numerical simulations performed with an imposed driving flux. Another type of non-diffusive transport is often seen in high-density H-modes with prolonged ELM-free periods, where the transport near the separatrix is dominated by quasi-coherent modes driving particle and/or heat fluxes exceeding L-mode levels. These modes may play an important role by providing particle and/or heat exhaust between ELMs.

Far SOL transport and main wall plasma interaction in DIII-D

Far Scrape-Off Layer (SOL) and near-wall plasma parameters in DIII-D depend strongly on the discharge parameters and confinement regime. In L-mode discharges cross-field transport increases with the average discharge density and flattens far SOL profiles, thus increasing plasma contact with the low field side (LFS) main chamber wall. In H-mode between edge localized modes (ELMs) the plasma?wall contact is weaker than in L-mode. During ELM fluxes of particles and heat to the LFS wall increase transiently above the L-mode values. Depending on the discharge conditions, ELMs are responsible for 30?90% of the net ion flux to the outboard chamber wall. ELMs in high density discharges feature intermittent transport events similar to those observed in L-mode and attributed to blobs of dense hot plasma formed inside the separatrix and propagating radially outwards. Though the blobs decay with radius, some of them survive long enough to reach the outer wall and possibly cause sputtering. In lower density H-modes, ELMs can feature blobs of pedestal density propagating all the way to the outer wall.

Operation of Alcator C-Mod with high-Z plasma facing components and implications

Studies of potential plasma facing component (PFC) materials for a magnetic fusion reactor generally conclude that tungsten is the best choice due to its low tritium (T) retention, capability to handle high heat fluxes with low erosion, and robustness to nuclear damage and activation. ITER [F. Perkins et al., Nucl. Fusion 39, 2137 (1999)] may operate with all tungsten PFCs to provide the necessary operational experience for a reactor. Alcator C-Mod [I. Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] operates with molybdenum (Mo) high-Z PFCs, which have very similar properties to tungsten. The experiments described herein have provided a unique comparison of operation with or without in situ boron coatings applied to the molybdenum PFCs; the latter are likely most relevant to ITER and beyond. ICRF-heated H-modes were readily achieved without boron coatings although the resultant enhancement in energy confinement was typically small (HITER,89∼1). Molybdenum concentrations, nMo∕ne, rise rapidly after the H-...