Challenges on the road towards fusion electricity

The ultimate aim of fusion research is to generate electricity by fusing light atoms into heavier ones, thereby converting mass into energy. The most efficient fusion reaction is based on merging the hydrogenic isotopes: Deuterium (2D) and Tritium (3T) into Helium (4He) and a neutron, which releases 17.6 MeV in the form of kinetic energy of the reaction products

JET experiments with tritium and deuterium–tritium mixtures

Extensive preparations are now underway for an experiment in the Joint European Torus (JET) using tri-tium and deuterium–tritium mixtures. The goals of this experiment are described as well as the progressthat has been made in developing plasma operational scenarios and physics reference pulses for usein deuterium–tritium and full tritium plasmas. At present, the high performance plasmas to be testedwith tritium are based on either a conventional ELMy H-mode at high plasma current and magnetic field(operation at up to 4 MA and 4 T is being prepared) or the so-called improved H-mode or hybrid regimeof operation in which high normalised plasma pressure at somewhat reduced plasma current results inenhanced energy confinement. Both of these regimes are being re-developed in conjunction with JET’sITER-like Wall (ILW) of beryllium and tungsten. The influence of the ILW on plasma operation and perfor-mance has been substantial. Considerable progress has been made on optimising performance with theall-metal wall. Indeed, operation at the (normalised) ITER reference confinement and pressure has beenre-established in JET albeit not yet at high current. In parallel with the physics development, extensivetechnical preparations are being made to operate JET with tritium. The state and scope of these prepara-tions is reviewed, including the work being done on the safety case for DT operation and on upgradingmachine infrastructure and diagnostics. A specific example of the latter is the planned calibration at14 MeV of JET neutron diagnostics

Thomson scattering with laser intra-cavity multi-pass system to study fast changing structures in fusion plasma

\u3cp\u3eAdvanced Thomson Scattering (TS) diagnostics with a Laser Intra-cavity Multi-pass Probing (LIMP) system were developed and implemented in several tokamaks in recent decades. The LIMP system provides a significant gain of the laser probing energy along with generation of many pulses at a high repetition rate within a single pumping pulse. Probing lasers with a LIMP system have considerably extended the capability of the TS diagnostic and have opened new application fields in plasma physics.The paper presents the principles of the LIMP system which underlie its efficiency as well as some of its applications in physical studies performed in the TEXTOR tokamak. The TS diagnostic with LIMP has enabled detailed measurements of fluctuations of electron temperature and density along the whole plasma diameter. The structure of thesefluctuations, which are typically1-2% in amplitude, has been resolved. The diagnostic made it possible to measure the fine structure of rotating magnetic islands in the TEXTOR plasma. The observed structuressignificantly differ from the classical shape of helical magnetic perturbations. LIMP isapplied for measurementsofnot only electrons butalso of other plasma particles. Laser-heated dust micro-particles were detected in plasma via their thermal radiation. The diagnostic provides the location of particlesalong with its velocity, temperature and dimension. .\u3c/p\u3

Diagnostics for plasma control on DEMO : challenges of implementation

As a test fusion power plant, DEMO will have to demonstrate reliability and very long pulse/steady-state operation, which calls for unprecedented robustness and reliability of all diagnostic systems (also requiring adequate redundancy). But DEMO will have higher levels of neutron and gamma fluxes, and fluences, nuclear heating, and fluxes of particles than ITER, and probably reduced physical access. In particular, the neutron fluence will be about 15–50 times higher than that in ITER. As a consequence, some diagnostics that will work in ITER are likely to be unfeasible in DEMO. It is important, therefore, to develop a new way of thinking with respect to that employed to date in which diagnostics are added after the machine has been basically designed: if certain diagnostics are deemed essential for the control of DEMO, they will have to be taken into account during the entire design phase

Overview of the JET results

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor

Two-dimensional visualization of growth and burst of the edge-localized filaments in KSTAR H-mode plasmas

The filamentary nature and dynamics of edge-localized modes (ELMs) in the KSTAR high-confinement mode plasmas have been visualized in 2D via electron cyclotron emission imaging. The ELM filaments rotating with a net poloidal velocity are observed to evolve in three distinctive stages: initial linear growth, interim quasisteady state, and final crash. The crash is initiated by a narrow fingerlike perturbation growing radially from a poloidally elongated filament. The filament bursts through this finger, leading to fast and collective heat convection from the edge region into the scrape-off layer, i.e., ELM crash

Sawtooth precursor oscillations on DIII-D

The sawtooth oscillation, observed in tokamak plasmas with a central safety factor of less than unity, is a periodic disruptive instability characterized by a slow ramping of central plasma density and temperature, followed by a fast relaxation resulting in flattening of both profiles. Elongated neutral-beamheated discharges on the DIII-D tokamak exhibit multiple precursor oscillations with mode number m/n = 1/1. The dominant m/n = 1/1 mode oscillates at the plasma rotation frequency. A downshifted mode also appears early in the sawtooth ramp. A normalization of electron cyclotron emission imaging data that removes the contribution of slow electron temperature profile evolution reveals that both modes are consistent with an underlying quasi-interchange plasma displacement