[en] JET was extensively modified in the 1992/93 shutdown. The new pumped divertor and many new systems were brought into operation early in 1994. Operations have progressed to 4MA plasma current and, with substantial additional heating, H-mode confinement results confirm the expected scaling. The high power handling capability of the pumped divertor with sweeping is estimated at 20MW for 20s. H-mode plasmas have large Type I ELMs. With lower hybrid heating alone, 2MA full current drive has been achieved with good efficiency, and with ICRF power, effective heating and direct electron heating have been demonstrated. (Author)

[en] In JET the scientific properties and technical basis of good confinement regimes have been evaluated in the light of the potential extrapolation of such regimes to reactor requirements. In this paper the main experimental H-mode results are discussed highlighting global confinement scaling, low q regimes, the role of the target plate material, the density limit, and finally sawtooth suppression and hot-ion mode. 17 refs., 15 figs

[en] In the operation of JET and of any tokamak many discharges are terminated by a major disruption. The disruptive termination of a discharge is usually an unwanted event which may cause damage to the structure of the vessel. In a reactor disruptions are potentially a very serious problem, hence the importance of studying them and devising methods to avoid disruptions. Statistical information has been collected about the disruptions which have occurred at JET over a long span of operations. The analysis is focused on the operational aspects of the disruptions rather than on the underlining physics. (Author)

[en] A tokamak fusion reactor should ignite at low powers and operate at high powers without thermal instability. The properties of the ignition domain and plasma thermal stability have been analysed with a zero dimensional model whose basic features are those of the POPCON code. A mechanism of stabilizing the reactor operating region, based on the control of energy confinement sensitive parameters such as plasma current, is presented. (author) 5 refs., 3 figs

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[en] The JET vacuum system has been operated routinely since June 1983. It consists of an all metallic fully welded vessel, pumped by four turbomolecular pumps. The pumping speed at the torus is 8500 ls^-1 for hydrogen. The vacuum vessel (volume 200 m³, total surface 1000 m², is bakeable up to 500⁰C, but has been commissioned only to 300⁰C. The base pressure at 250⁰C is 2.5 x 10^-7 mbar for hydrogen and <3 x 10^-10 mbar for impurities. The wall conditioning methods employed are radio frequency assisted glow discharge and pulse discharge cleaning. The resutling z/sub e/sub f/sub f/ values during plasma operation (spitzer resistivity, uncorrected) are between 3 and 5. This paper describes the vacuum system, the operational experience and the modifications required for future tritium operation. The techniques and results of the wall conditioning employed are summarized with respect to plasma performance. Consequences for the choice of limiter and wall material is discussed

[en] The current baseline of ITER foresees 2 Heating Neutral Beam (HNB's) systems based on negative ion technology, each accelerating to 1 MeV 40 A of D^- and capable of delivering 16.5 MW of D⁰ to the ITER plasma, with a 3rd HNB injector foreseen as an upgrade option. In addition a dedicated Diagnostic Neutral Beam (DNB) accelerating 60 A of H^- to 100 keV will inject ≅15 A equivalent of H⁰ for charge exchange recombination spectroscopy and other diagnostics. Recently the RF driven negative ion source developed by IPP Garching has replaced the filamented ion source as the reference ITER design. The RF source developed at IPP, which is approximately a quarter scale of the source needed for ITER, is expected to have reduced caesium consumption compared to the filamented arc driven ion source. The RF driven source has demonstrated adequate accelerated D^- and H^- current densities as well as long-pulse operation. It is foreseen that the HNB's and the DNB will use the same negative ion source. Experiments with a half ITER-size ion source are on-going at IPP and the operation of a full-scale ion source will be demonstrated, at full power and pulse length, in the dedicated Ion Source Test Bed (ISTF), which will be part of the Neutral Beam Test Facility (NBTF), in Padua, Italy. This facility will carry out the necessary R and D for the HNB's for ITER and demonstrate operation of the full-scale HNB beamline. An overview of the current status of the neutral beam (NB) systems and the chosen configuration will be given and the ongoing integration effort into the ITER plant will be highlighted. It will be demonstrated how installation and maintenance logistics have influenced the design, notably the top access scheme facilitating access for maintenance and installation. The impact of the ITER Design Review and recent design change requests (DCRs) will be briefly discussed, including start-up and commissioning issues. The low current hydrogen phase now envisaged for start-up imposed specific requirements for operating the HNB's at full beam power. It has been decided to address the shinethrough issue by installing wall armour protection, which increases the operational space in all scenarios. Other NB related issues identified by the Design Review process will be discussed and the possible changes to the ITER baseline indicated.

[en] In the study of tokamak machines able to sustain plasmas of thermonuclear interest (JIT, IGNITOR, NET, CIT, ET), there is a strong quest for engineering optimization of the circuital components close to the plasma. We have developed a semianalytical axisymmetric MHD equilibrium code based on the technique of the poloidal ψ flux function expansion in toroidal harmonic series. This code is able to optimize the necessary currents in the poloidal circuits in order to sustain a plasma of fixed shape (also x-point configuration), toroidal current and poloidal β. (author) 4 refs., 4 figs

[en] Since the discovery of the H-mode in ASDEX, the term has been used to describe plasma regimes characterized by superior confinement and by specific edge signatures such as reduction of recycling and of high frequency fluctuations. Often H-mode discharges, (without ELM's), have shown the undesirable characteristics of impurity accumulation and uncontrolled rise of plasma density. Moreover the H-mode tends to produce flat-hollow density profiles and broad temperature profiles. Compatibility of basic H-mode characteristics with monster sawtooth, hot-ion mode and pellet peaked density profiles would lead to substantial improvements of plasma performance. In this paper it is shown that the JET H-mode can be compatible with these improved regimes. In the new regimes the basic confinement time is little changed from that the basic H-mode, whilst the central plasma parameters can be substantially improved. (author) 13 refs., 4 figs