Tatsuo Sugie

Bio: Tatsuo Sugie is an academic researcher from Japan Atomic Energy Agency. The author has contributed to research in topics: Divertor & Tokamak. The author has an hindex of 17, co-authored 63 publications receiving 1264 citations.

Chapter 7: Diagnostics

Abstract: In order to support the operation of ITER and the planned experimental programme an extensive set of plasma and first wall measurements will be required. The number and type of required measurements will be similar to those made on the present-day large tokamaks while the specification of the measurements—time and spatial resolutions, etc—will in some cases be more stringent. Many of the measurements will be used in the real time control of the plasma driving a requirement for very high reliability in the systems (diagnostics) that provide the measurements. The implementation of diagnostic systems on ITER is a substantial challenge. Because of the harsh environment (high levels of neutron and gamma fluxes, neutron heating, particle bombardment) diagnostic system selection and design has to cope with a range of phenomena not previously encountered in diagnostic design. Extensive design and R&D is needed to prepare the systems. In some cases the environmental difficulties are so severe that new diagnostic techniques are required. a Author to whom any correspondence should be addressed.

Technological challenges of ITER diagnostics

Abstract: The technological challenges associated with preparing and installing diagnostic systems on ITER are summarized and the solutions that have been developed to overcome them are described. Areas where further developments are needed are highlighted.

Progress in research and development of mirrors for ITER diagnostics

Abstract: Metallic mirrors will be used as plasma-viewing elements in all optical and laser diagnostic systems in ITER. In the harsh environment of ITER, the performance of mirrors will decrease mainly because of the erosion of their surfaces and deposition of impurities. The deterioration of the optical properties of diagnostic mirrors will directly affect the entire performance of the respective ITER diagnostics, possibly leading to their shutdown. Therefore, R&D on mirrors is of crucial importance for ITER diagnostics. There is a coordinated worldwide R&D programme supervised by the Specialists Working Group on first mirrors of the International Tokamak Physics Activity, Topical Group on Diagnostics. This paper provides an overview of new results in the field of first mirrors, covering the manufacturing of ITER mirror prototypes, investigations of mitigation of deposition and mirror cleaning and the predictive modelling of the mirror performance in ITER. The current status of research on beryllium deposition—a new critical area of mirror research—is given along with an outlook for future activities.

The spectral profile of the line emitted from the divertor region of JT-60U

Abstract: In order to understand the recycling and emission processes of deuterium atoms, spectral profiles of the line emitted from the divertor region of JT-60U have been observed with a high-resolution spectrometer and analysed by simulation with a three-dimensional neutral particle transport code. The profile has been explained as composed of narrow and broad components; the narrow component is attributed to dissociative excitation and electron collisional excitation of the atoms produced by dissociation, and the broad component is attributed to electron collisional excitation of the atoms produced by reflection and charge exchange. In low-density plasmas, the simulated line profile agrees reasonably well with that observed, although the component attributed to the atoms reflected at the divertor tiles is overestimated by a factor of about two. Dissociative excitation from deuterium molecules and molecular ions plays an important role for the line intensity. The ratio of the line intensity to the deuterium atom flux for high-energy deuterium atoms, which are produced by the reflection and charge exchange, is reduced, because the fast atoms readily escape from the divertor plasma. The width of the narrow component in a low-density case corresponds to a temperature of deuterium atoms of 1.3 eV, and that in a high-density case corresponds to a temperature of 2.2 eV.

Observation of the limiter H mode in the JT-60 tokamak with lower-hybrid current drive

Abstract: The {ital H} mode was achieved in limiter discharges of a tokamak with lower-hybrid current drive for the first time. Simultaneous application of rf powers at two different frequencies such as 1.74 and 2.23 GHz or 1.74 and 2.0 GHz appeared to be effective in the attainment of the {ital H} mode. The threshold lower-hybrid power was as low as Ohmic-heating power with hydrogen plasmas. A nearly steady-state edge-localized-mode-free {ital H} mode with a duration up to 3.3 sec was established without significant impurity accumulation.

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

Abstract: 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.

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

Abstract: 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.

Chapter 4: Power and particle control

Abstract: 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 2: Plasma confinement and transport

Abstract: 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.