Modelling of non-linear edge harmonic oscillations and the effect of non-axisymmetric magnetic coils

Quiescent H-mode (QH-mode) operation is highly desirable relative to the well known H- mode operation because it allows for high energy density in the core as well as potentially im- proved confinement of energy and particles. A feature of QH-mode is that the Edge Localized Mode (ELM) instabilities are replaced with continuous Edge Harmonic Oscillations (EHO) [1]. Control and reproduction of robust QH-mode regimes could be crucial for future fusion opera- tion. In some cases it may be useful to modify the main characteristics of QH-mode discharges while keeping constant the key elements needed to achieve QH-mode (such as applied torque, edge current density, pedestal pressure, etc.). With that in mind, the effect of global toroidal mode seeding in the vacuum magnetic field on the amplitude of EHO is investigated. Previ- ous numerical simulations have shown that free boundary equilibrium calculations are able to recover non-linear saturated 3D equilibria with edge corrugations associated with EHO in QH- mode [2, 3]. Here we use the VMEC free boundary code to do a proof of principle investigation of the coupling between EHO and global toroidal mode seeding. This is done by first producing a QH-mode plasma using the vacuum magnetic field modeled by JET-like toroidal and poloidal field coils. Then, a global toroidal mode n=1,2 is seeded numerically by perturbing the vacuum field using a set of non-axisymmetric external coils associated with the Error Field Correction Coils (EFCC’s) in JET and the plasma response is studied. Due to the EFCC’s geometry, it is crucial to account for up-down asymmetry in the VMEC code. Spectral decomposition of the 3D plasma displacement with respect to the equivalent 2D axisymmetric equilibrium is per- formed and compared with linear numerical simulations using the KINX code. A study that effectively isolates the coupling of global n=1,2 toroidal modes with current driven and pres- sure driven EHO is presented, and the impact of external coils and associated plasma response clearly determined.

Chapter 4: Power and particle control

Richard Pitts

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.

2007

SOLPS5 simulations of Type I ELMing H-mode at JET

Barbora Gulejová, Richard Pitts

This paper aims to contribute both to the ongoing process of scrape-off layer code-experiment and code-code benchmarking. Results are presented from SOLPS5 simulations of two high power JET H-modes with similar magnetic configuration, concentrating in the first case on the ELM-free phase of high 1p, similar to 8 MJ stored energy plasmas with ELMs approaching 1 MJ, modeled for the first time with this code package. A second pulse, with lower stored energy and smaller ELMs, originally considered in detail by Kallenbach with the EDGE2D-NIMBUS code package [Kallenbach et al., Plasma Phys. Control. Fus. 46 (2004) 431], has been modeled as a benchmarking exercise featuring a high level of complexity including carbon impurities and the full ELM cycle. Good agreement is found between the code results. The SOLPS5 results are used to analyse the energy balance during the ELM cycle. In both H-mode discharges, a strong inward particle pinch in the pedestal region is found to be necessary to match measured upstream profiles. (C) 2009 Elsevier B.V. All rights reserved.

Elsevier2009

Extremely Shaped Plasmas to Improve the Tokamak Concept

Francesco Piras

Energy is essential for human existence and our future depends on plentiful and accessible sources of energy. The world population is fast growing and the average energy used per capita increases. One of the greatest challenges for human beings is that of meeting the growing demand for energy in a responsible, equitable and sustainable way. The possibility to obtain energy by "fusing" light atoms addresses these needs. Nuclear fusion reactions are clean, safe and the amount of fuel present on Earth (hydrogen isotopes) is practically inexhaustible and well distributed. Nuclear fusion is a natural process that occurs in all active stars like our Sun. Since the first demonstration of a deuterium fusion reaction (Rutherford 1933), researchers worldwide have tried to replicate this process on Earth by building a thermonuclear fusion reactor. Nevertheless, the challenge posed by the construction of a nuclear fusion reactor is greater than the one presented earlier by the development of a fission reactor. During the IAEA Conference in Geneva in the early 1958, L.A.Artsimovich declared: "Plasma physics is very difficult. Worldwide collaboration is needed for progress" and E.Teller, at the same conference: "Fusion technology is very complex. It is almost impossible to build a fusion reactor in this century". They were right. The extremely high temperature and density necessary to fuse hydrogen isotopes makes it difficult indeed to create a successful fusion reactor. Even though the physics of the fusion reaction appears clear, we are still facing problems on the road towards bulding the "box" that can efficiently confine the hot gas in the state of plasma. The best results so far have been obtained confining a plasma with strong magnetic fields in a toroidal configuration ("tokamak"). The Centre de Recherches en Physique des Plasmas in Switzerland actively studies this promising configuration towards the development of a nuclear fusion reactor. The experimental activity of the Tokamak à Configuration Variable (TCV) mainly focuses on the research of optimized plasma shapes capable of improving the global performance and solve the technological challenges of a tokamak reactor. Several theoretical and experimental results show the importance of the plasma shape in tokamaks. The maximum value of β (an indicator of the confinement efficiency) is for example related to the ratio between the height and the width of the plasma. The plasma shape can also affect the power necessary to access improved confinement regimes, as well as the plasma stability. This thesis reports on a contribution towards the optimization of the tokamak plasma shape. In particular, it describes the theoretical and experimental studies carried out in the TCV tokamak on two innovative plasma shapes: the doublet shaped plasma and the snowflake divertor. Doublet shaped plasmas have been studied in the past by the General Atomics group. Since then, the development of new plasma diagnostics and the discovery of new confinement regimes have given new reasons for interest in this unusual configuration. TCV is the only tokamak worldwide theoretically able to establish and control this configuration. This thesis illustrates new motivations for creating doublet plasmas. The vertical stability of the configuration is studied using a rigid model and the results are compared with those obtained with the KINX MHD stability code. The best strategy for controlling a doublet on TCV is also investigated, and a possible setup of the TCV control system is suggested for the doublet configuration. Analyzing the possible scenarios for doublet creation, the most promising scenario consists of the creation of two independent plasmas, which are subsequently merged to establish a doublet. For this reason, particular attention needs to be devoted to the problem of the plasma start-up. In this thesis, a general analysis of the TCV ohmic and assisted with ECH plasma start-up is presented, and recent attempts to create a doublet plasma are reported. Since the magnetic field reconstruction at the breakdown time is important to better diagnose these plasmas, the entire magnetic system of TCV has been calibrated with an original technique, also described in the manuscript. The last part of this thesis is devoted to the snowflake divertor configuration. This innovative plasma shape has been proposed and theoretically studied by Dr. D.D.Ryutov from the Lawrence Livermore National Laboratory. In Ryutov's articles, this configuration was proposed to alleviate the problems of the plasma-wall interaction and possibly affect the plasma edge stability. The TCV tokamak was the first to report the creation and control of a snowflake configuration, and the candidate was the principal investigator of this work. These results are accordingly discussed in this thesis. Details are provided in particular on the strategy used to establish the configuration. An edge-localized mode (ELM) H-mode regime, supported by electron cyclotron heating, has been successfully established in a snowflake. This regime exhibits 2 to 3 times lower ELM frequency but only a 20%-30% increase in normalized ELM energy (ΔWELM/WP ) compared to an identically-shaped, conventional, single-null, diverted H-mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. Finally, the capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally and is also reported in this thesis.