L. C. Ingesson

Bio: L. C. Ingesson is an academic researcher from European Atomic Energy Community. The author has contributed to research in topics: Divertor & Jet (fluid). The author has an hindex of 12, co-authored 17 publications receiving 1043 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.

Plasma detachment in JET Mark I divertor experiments

Abstract: 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.

Disruption heat loads on the JET MkIIGB divertor

Abstract: Combined analysis of divertor thermocouple and IR camera measurements during JET disruptions can provide valuable information on the distribution of the energy loads, even if the stored energy of the JET plasmas is small compared to that foreseen for the next-generation tokamaks. Typically the energy collected at the divertor represents a small fraction of the pre-disruption plasma energy; this is consistent with the high level of radiation observed and with part of the magnetic energy being transferred to the plasma-coupled conductors. The data for this paper are taken from the whole set of disruptive plasmas of JET operation in the years 2000 and 2001. In most of the MkIIGB disruptions, the plasma displaces upwards (away from the divertor); therefore, only a small number of downward events are available for analysis. However, divertor heat loads seem to be more strongly correlated to the delay of the loss of the X-point with respect to the thermal quench than the direction of the plasma displacement. When the plasma thermal energy is lost with the plasma still in X-point configuration, the septum and the tiles wetted by the strike-points, often more than one tile per strike-point, experience a sharp increase in temperature, equivalent to up to 1 MJ m-2. When the thermal quench occurs at the same time as, or after, the loss of plasma vertical control, no significant divertor tile temperature obreak increase can be observed for both upwards and downwards events. Most of the disruptions purposely made to produce runaway electrons went towards the divertor and, although not systematically, lead to local (mostly at the septum) temperature increase equivalent to a load up to 2 MJ m-2, often toroidally asymmetric.

Impurity-seeded plasma experiments on JET

Abstract: Scaling to larger tokamaks of high confinement plasmas with radiating edges, induced by impurities, is being studied through internationally collaborative experiments on JET. In campaigns till the end of 2000, three different regimes have been explored. A small number of limiter L-mode discharges seeded with neon have most closely repeated the approach used on TEXTOR-94, but different collisionality and particle transport in JET impede central peaking of the density associated with improved confinement. Divertor L-modes at intermediate density, again with neon injection, have pursued transiently enhanced states found on DIII-D. Confinement up to H-mode quality, together with radiation fractions of ?40%, have briefly been obtained, though central Zeff quickly increases. Most effectively, neon and argon seeding of higher density ELMy H-modes formed mainly at low triangularity on the septum of the MkIIGB divertor, resembling a pumped-limiter arrangement, have been examined. Good confinement has been sustained at densities close to the Greenwald level in `afterpuff' (AP) phases following the end of main gas fuelling, for little change of central Zeff but up to ?60% radiation. Outstanding normalized properties up to H97 = 0.99 at fGwd = 0.94 have thus been achieved, above the conventional H-mode density limit for diverted plasmas. Stationarity of states has also been extended to many energy confinement times by including low, extra gas inputs in the `AP', suggestive of an optimized fuelling scheme. Further development in 2001 is reported separately in [1]. Accompanying ELMs are generally reduced in frequency though not evidently in size, electron pedestal pressure being almost unchanged from unseeded behaviour. There are indications of the most favourable impurity species scaling with plasma parameters, performance, radiation and its concentration within a mantle all increasing with argon compared to neon in JET. These benefits in terms of integrated properties are just as required for long burning pulses in ITER, supporting its use of a radiating mantle to assist not only power exhaust but performance too. Impurity-seeded H-modes can therefore contribute directly to next-step scenario development.

Towards the realization on JET of an integrated H-mode scenario for ITER

Abstract: ELMy H-mode experiments at JET in 2000/mid-2002 have focused on discharges with normalized parameters for plasma density, energy confinement and beta similar to those of the ITER Q(DT) = 10 referen ...

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 1: Overview and summary

Abstract: The ITER Physics Basis presents and evaluates the physics rules and methodologies for plasma performance projections, which provide the basis for the design of a tokamak burning plasma device whose goal is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. This Chapter summarizes the physics basis for burning plasma projections, which is developed in detail by the ITER Physics Expert Groups in subsequent chapters. To set context, the design guidelines and requirements established in the report of ITER Special Working Group 1 are presented, as are the specifics of the tokamak design developed in the Final Design Report of the ITER Engineering Design Activities, which exemplifies burning tokamak plasma experiments. The behaviour of a tokamak plasma is determined by the interaction of many diverse physics processes, all of which bear on projections for both a burning plasma experiment and an eventual tokamak reactor. Key processes summarized here are energy and particle confinement and the H-mode power threshold; MHD stability, including pressure and density limits, neoclassical islands, error fields, disruptions, sawteeth, and ELMs; power and particle exhaust, involving divertor power dispersal, helium exhaust, fuelling and density control, H-mode edge transition region, erosion of plasma facing components, tritium retention; energetic particle physics; auxiliary power physics; and the physics of plasma diagnostics. Summaries of projection methodologies, together with estimates of their attendant uncertainties, are presented in each of these areas. Since each physics element has its own scaling properties, an integrated experimental demonstration of the balance between the combined processes which obtains in a reactor plasma is inaccessible to contemporary experimental facilities: it requires a reactor scale device. It is argued, moreover, that a burning plasma experiment can be sufficiently flexible to permit operation in a steady state mode, with non-inductive plasma current drive, as well as in a pulsed mode where current is inductively driven. Overall, the ITER Physics Basis can support a range of candidate designs for a tokamak burning plasma facility. For each design, there will remain a significant uncertainty in the projected performance, but the projection methodologies outlined here do suffice to specify the major parameters of such a facility and form the basis for assuring that its phased operation will return sufficient information to design a prototype commercial fusion power reactor, thus fulfilling the goal of the ITER project.

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.

Characteristics of type I ELM energy and particle losses in existing devices and their extrapolation to ITER

Abstract: Analysis of Type I ELMs from ongoing experiments shows that ELM energy losses are correlated with the density and temperature of the pedestal plasma before the ELM crash. The Type I ELM plasma energy loss normalized to the pedestal energy is found to correlate across experiments with the collisionality of the pedestal plasma (ν*ped), decreasing with increasing ν*ped. Other parameters affect the ELM size, such as the edge magnetic shear, etc, which influence 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. In JET and DIII-D, under some conditions, ELMs can be observed (`minimum' Type I ELMs with energy losses acceptable for ITER), that do not affect the plasma temperature. The duration of the divertor ELM power pulse is correlated with the typical ion transport time from the pedestal to the divertor target (τ||Front = 2πRq95/cs,ped) and not with the duration of the ELM-associated MHD activity. Similarly, the timescale of ELM particle fluxes is also determined by τ||Front. The extrapolation of the present experimental results to ITER is summarized.