G. Bracco

The International Multi-Tokamak Profile Database

An international multi-tokamak profile database has been assembled, constituting a representative set of reference tokamak discharges for the purpose of testing local transport models against well documented data. In particular, it will allow one to measure the accuracy with which the models can reproduce experiments and draw confidence intervals for the predictions of the models outside the range covered in the database. This database is now available to the fusion community and may be accessed by anonymous ftp to iterphys.naka.go.jp; the purpose of this article is to describe the structure of the database and the discharges contributing to it so that all can take full advantage of this resource. Thus, after an introductory general discussion of the database, there is a more detailed description of its structure, with listings of variables emphasized and how to access the database. There is then a brief description of each contributing tokamak and information on the type of discharges available from that tokamak. This is followed by a more quantitative description of the data, giving the ranges of dimensional and dimensionless variables available. Some typical modelling results to illustrate the use of the database are given in the conclusion.

The 2008 public release of the international multi-tokamak confinement profile database

This paper documents the public release PR08 of the International Tokamak Physics Activity (ITPA) profile database, which should be of particular interest to the magnetic confinement fusion community. Data from a wide variety of interesting discharges from many of the worlds leading tokamak experiments are now made available in PR08, which also includes predictive simulations of an initial set of operating scenarios for ITER. In this paper we describe the discharges that have been included and the tools that are available to the reader who is interested in accessing and working with the data. Most discharge descriptions refer to more detailed previous publications. In addition, we review physics analyses that have already made use of the profile database discharges. Public access to PR08 data is unconditional, but this paper should be cited by any publication that makes use of PR08 data.

Recent progress on the development and analysis of the ITPA global H-mode confinement database

This paper describes the updates to and analysis of the International Tokamak Physics Activity (ITPA) Global H-Mode Confinement Database version 3 (DB3) over the period 1994–2004. Global data, for the energy confinement time and its controlling parameters, have now been collected from 18 machines of different sizes and shapes: ASDEX, ASDEX Upgrade, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M, JT-60U, MAST, NSTX, PBX-M, PDX, START, T-10, TCV, TdeV, TFTR and TUMAN-3M. The database now contains 10382 data entries from 3762 plasma discharges, including data from deuterium–tritium experiments, low-aspect ratio plasmas, dimensionless parameter experiments and plasmas. DB3 also contains an increased amount of data from a range of diverted machines and further data at high triangularity, high density and high current. A wide range of physics studies has been performed on DB3 with particular progress made in the separation of core and edge behaviour, dimensionless parameter analyses and the comparison of the database with one-dimensional transport codes. The errors in the physics variables of the database have also been studied and this has led to the use of errors in variables fits. A key aim of the database has always been to provide a basis for estimating the energy confinement properties of next step machines such as ITER, and so the impact of the database and its analysis on such machines is also discussed.

EXPERIMENTS IN FTU WITH DIFFERENT LIMITER MATERIALS

Over the last few years, a great deal of effort has been devoted to solving the problem of power and particle handling in divertors, which has been recognized as a critical issue for the operation of a magnetic fusion reactor. In particular, the choice of materials for plasma facing components has been examined with a view to developing heat and erosion resistant materials for divertor target plates. A large database on the behaviour of low-Z (carbon or beryllium) materials in tokamaks is available, while for high-Z materials there is little experience in the present generation of magnetic fusion devices. Frascati Tokamak Upgrade (FTU), a high field compact tokamak, has devoted part of its experimental campaign to studying the plasma characteristics when its limiter material is changed from the usual Inconel (nickel) to molybdenum and tungsten. Siliconization of the machine has also allowed the comparison of plasma performance when a relatively low-Z (silicon) ion is the dominant impurity. In this article, results are reported concerning the plasma operation, the differences in plasma characteristics and radiation losses, the impurity generation mechanisms and the relative impurity concentrations in the core plasma. A simulation of the experimental results, made with a self-consistent edge-core coupled model is presented, in order to provide evidence of the main physics mechanisms responsible for the observed behaviour

Gradient length driven electron heat transport study in modulated electron cyclotron heating FTU tokamak

Perturbative and steady-state heat transport of FTU tokamak in current ramp-up discharges are investigated by means of modulated electron cyclotron heating (MECH). Perturbative and steady-state transport experiments are coherent with an electron heat transport which switches from low to high values when electron temperature gradient length reaches a threshold value 1/LTc. The threshold value 1/LTc is shown to be proportional to the ratio s/q. The experimental findings are compared to predictions of an empirical model based on the assumption of a threshold gradient length, LTc (1/LT = |∇Te/Te|), in the electron temperature Te below which electron thermal diffusivity, χe, switches from low to high values. Plasma responses to steady state and MECH are modelled assuming the electron diffusivity as χPB = χ0 + αTe3/2(1/LT-1/LTc)1/2; here Te3/2 reflects the gyro-Bohm assumption, χ0 represents the heat transport for 1/LT 1/LTc, mimics an extra transport possibly due to electron temperature gradient (ETG) modes. In agreement with ETG threshold 1/LTc is shown to be correlated with the magnetic shear s.

Plasma characteristics in FTU with different plasma facing materials

Experiments with different dominant impurities (C, Si, Ni, Mo) have been carried out on the FTU tokamak. Density limit, marfe, Zeff and radiation losses are compared and discussed. An analytical model that couples in a self consistent way the parameters of the SOL and the transport of the impurities generated at the limiter to the bulk plasma parameters has been used to reproduce the experimental data.

Scaling of the energy confinement time with β and collisionality approaching ITER conditions

The condition of the latest version of the ELMy H-mode database has been re-examined. It is shown that there is bias in the ordinary least squares regression for some of the variables. To address these shortcomings three different techniques are employed: (a) principal component regression, (b) an error in variables technique and (c) the selection of a better conditioned dataset with fewer variables. Scalings in terms of the dimensionless physics variables, as well as the standard set of engineering variables, are also derived. The new scalings give a very similar performance for existing scalings for ITER at the standard beta(n) of 1.6, but a much improved performance at higher beta n.

MHD and confinement during pellet injection on FTU

The study sets out to investigate the parts of the FTU operating space over which pellet injection effects performance improvement and to elucidate the phenomena which hinder it. In this respect the behaviour of the q=1 resonance is found to play the most relevant role.

Design and calibration of the JET time of flight neutral particle analyzer with high noise rejection capability

This paper describes the design and calibration of a time‐of‐flight neutral particle analyzer used on the Joint European Torus (JET) tokamak in the diagnostics of plasma close to thermonuclear fusion conditions. The 15 channels of the analyzer can select energy and mass of charge exchange neutrals in the energy range from 0.5 to 250 keV. The analyzer has been calibrated for hydrogen, deuterium, and helium atoms. The utilization of the coincidence technique in the measurement of the time of flight of the detected neutrals also permits discrimination of the particle signal from a random background. Thus, the NPA operates successfully in the presence of the high neutron and gamma fluxes that are characteristic of the high‐performance fusion plasmas produced in large tokamaks. The calibration data processing, the signal analysis, and the signal‐to‐noise evaluation are reported in detail.

Effects of wall boron coating on FTU plasma operations

To achieve good performance on FTU in a large density range (0.3-6.0 × 10 20 m -3 ), boronization with a mixture of He (90%) and B 2 H 6 (10%) (diborane) as the feeding gas has been tested with thermal loads on the limiter surface up to 2.5 MW/m 2 . With boronized limiter (TZM alloy with 98% of Mo) and walls (SS AISI 304), the total radiated power drastically drops from 70-90% down to 35-45% and the Z cn decreases from 6.0 to 2.2 at 0.3-0.4 × 10 20 m -3 related to a strong reduction of heavy-metal concentration and to the getter effect of boron on oxygen (<0.5%). During this phase the action of the boron film as particle reservoir and its quick saturation due to the low temperature of FTU walls makes it difficult to obtain reproducible plasmas. Another consequence of boronization is the large dilution of the plasma with the hydrogen particles released from the B film. All these effects decrease after about 60 discharges when boron is eroded by the limiter but it is still present on the chamber walls. During this phase which lasts for more than 500 discharges, oxygen concentration does not increase at all and metal influx is lower than before boronization because the physical sputtering by oxygen ions and atoms is strongly reduced.