One of the scientific success stories of fusion research over the past decade is the development of the ExB shear stabilization model to explain the formation of transport barriers in magnetic confinement devices. This model was originally developed to explain the transport barrier formed at the plasma edge in tokamaks after the L (low) to H (high) transition. This concept has the universality needed to explain the edge transport barriers seen in limiter and divertor tokamaks, stellarators, and mirror machines. More recently, this model has been applied to explain the further confinement improvement from H (high)-mode to VH (very high)-mode seen in some tokamaks, where the edge transport barrier becomes wider. Most recently, this paradigm has been applied to the core transport barriers formed in plasmas with negative or low magnetic shear in the plasma core. These examples of confinement improvement are of considerable physical interest; it is not often that a system self-organizes to a higher energy state with reduced turbulence and transport when an additional source of free energy is applied to it. The transport decrease that is associated with ExB velocity shear effects also has significant practical consequences for fusion research. The fundamental physics involved in transport reduction is the effect of ExB shear on the growth, radial extent and phase correlation of turbulent eddies in the plasma. The same fundamental transport reduction process can be operational in various portions of the plasma because there are a number ways to change the radial electric field Er. An important theme in this area is the synergistic effect of ExB velocity shear and magnetic shear. Although the ExB velocity shear appears to have an effect on broader classes of microturbulence, magnetic shear can mitigate some potentially harmful effects of ExB velocity shear and facilitate turbulence stabilization.

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Effects of E×B velocity shear and magnetic shear on turbulence and transport in magnetic confinement devices

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 today's 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.

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

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.

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Chapter 4: Power and particle control

The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

https://iopscience.iop.org/article/10.1088/0029-5515/47/6/S02/pdf

Chapter 2: Plasma confinement and transport

Infrared thermography for condition monitoring – A review

Radiation-induced thermoelectric sensitivity in the mineral-insulated cable of magnetic diagnostic coils for ITER

Shine-through for hydrogen and helium beams in a plasma was measured. Shine-through for an atomic hydrogen beam with f1:f2:f3=7%:32%:61% (where f1, f2 and f3 represent fractions of the full, half and third energy, respectively) and a full energy of 140 keV, was smaller than the value predicted by usual cross-sections without inclusion of excitation processes at an electron density (ne) above 3*1013 cm-3. A fractional beam stopping increment ( delta ), which was defined by delta =( sigma s- sigma s(0)) sigma s(0) (here sigma s was an experimental stopping cross-section and sigma s(0) was the cross-section without including multistep processes via excited states), was about 0.2 at ne=(3.1-5.0)*1013 cm-3 and Zeff=1.6-2.5. The value of delta was consistent with the calculation by Janev et al. (1989, Nucl. Fusion 29, 2125). For a 160-190 keV helium beam, the enhancement of the stopping cross-section was not observed, as expected.

http://iopscience.iop.org/article/10.1088/0741-3335/32/6/002/pdf

Measurements of neutral beam stopping for hydrogen and helium in the JT-60 plasma

Spatial distributions of absolute line intensities of D I and C II-IV in the divertor region have been measured simultaneously with a 38-channel optical fibre array in JT-60U. The radiative losses in the divertor region have been derived from the intensities. In L-mode discharges without the MARFE, the fractions of the radiative losses due to deuterium and carbon are 36% and 64%, respectively. The carbon radiation is related to the carbon influx by a simple impurity model based on the confinement time of the carbon ion on the assumption of the physical sputtering. The power radiated due to carbon accounts for a large increase in the radiative losses by the MARFE.

Spectroscopic study of radiative losses in the JT-60U divertor plasma

Simulation of VDE under intervention of vertical stability control and vertical electromagnetic force on the ITER vacuum vessel

High-efficiency ICRF heating in DIVA is studied. At least 80% of the RF power coupled to the plasma is deposited with the bulk ions in the narrow central region, and about 70% of the deposited power is absorbed by the majority deuterium. The ion temperature increases up to three times the base temperature and thus exceeds the electron temperature.

https://iopscience.iop.org/article/10.1088/0029-5515/20/10/008/pdf

Tatsuo Sugie

Papers

Radiation-induced thermoelectric sensitivity in the mineral-insulated cable of magnetic diagnostic coils for ITER

Measurements of neutral beam stopping for hydrogen and helium in the JT-60 plasma

Spectroscopic study of radiative losses in the JT-60U divertor plasma

Simulation of VDE under intervention of vertical stability control and vertical electromagnetic force on the ITER vacuum vessel

High-efficiency ICRF heating in DIVA