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The N2-seeding experiments at JET were carried out in order to develop a type-I ELMy H-mode reference scenario for achieving Q = 10 in ITER compatible with the wall materials (Be main chamber, W divertor) of active operation. One objective of the experiments was to reduce the steady-state divertor-target power load by introducing a radiating extrinsic impurity (N2). An unexpected feature of N2-seeding, in high-triangularity (d) JET with an ITER-Like wall (JET-ILW) plasmas, was the observation of improved pedestal confinement at levels closer to those of JET with Carbon-Fibre-Composite Wall (JET-C). It has been suggested that the level of C and its associated radiation could have been a hidden parameter in the JET-C confinement behaviour of high-d plasmas and that this could partially be recovered by N2 seeding in JET-ILW. Detailed integrated modelling (core and SOL) of these JET-ILW discharges, with and without N2 seeding, may shed some light on the difference in plasma dynamics between high- and low-d configurations and, in particular, the improvement in confinement with high-d and N2-seeding. It is important to compare the simulations with a set of experimental parameters in the core/edge/SOL to ascertain the main physics governing these plasmas has been captured correctly. This paper reports on the present results.

G.P. Maddison, C. Giroud, B. Alper, G. Arnoux, I. Balboa, M.N.A. Beurskens, A. Boboc, S. Brezinsek, M. Brix, M. Clever, R. Coelho, J.W. Coenen, I. Coffey, P.C. da Silva Aresta Belo, S. Devaux, P. Devynck, T. Eich , R.C. Felton, J. Flanagan, L. Frassinetti, L. Garzotti, M. Groth, S. Jachmich, A. Järvinen, E. Joffrin, M.A.H. Kempenaars, U. Kruezi , K.D. Lawson, M. Lehnen, M.J. Leyland, Y. Liu, P.J. Lomas, C.G. Lowry, S. Marsen, G.F. Matthews, G.K. McCormick, A.G. Meigs, A.W. Morris, R. Neu, I.M. Nunes, M. Oberkofler, F.G. Rimini, S. Saarelma, B. Sieglin, A.C.C. Sips, A. Sirinelli , M.F. Stamp , G.J. van Rooij, D.J. Ward , M. Wischmeier and JET EFDA contributors The former all-carbon wall on JET has been replaced with beryllium in the main torus and tungsten in the divertor to mimic the surface materials envisaged for ITER. Comparisons are presented between Type I H-mode characteristics in each design by examining respective scans over deuterium fuelling and impurity seeding, required to ameliorate exhaust loads both in JET at full capability and in ITER. Attention is focused upon a common high-triangularity, single-null divertor configuration at 2.5MA, q95 3.5 yielding the most robust all-C performance. Contrasting results between the alternative linings are found firstly in unseeded plasmas, for which purity is improved and intrinsic radiation reduced in the ITER-like wall (ILW) but normalised energy confinement is 30% lower than in all-C counterparts, owing to a commensurately lower (electron) pedestal temperature. Divertor recycling is also radically altered, with slower, inboard-outboard asymmetric transients at ELMs and spontaneous oscillations in between them. Secondly, nitrogen seeding elicits opposite responses in the ILW to all-C experience, tending to raise plasma density, reduce ELM frequency, and above all to recover (electron) pedestal pressure, hence global confinement, almost back to previous levels. A hitherto unrecognised role of light impurities in pedestal stability and dynamics is consequently suggested. Thirdly, while heat loads on the divertor outboard target between ELMs are successfully reduced in proportion to the radiative cooling and ELM recurrence effects of N in both wall environments, more surprisingly, average power ejected by ELMs and estimated scrape-off-layer pressure ratio during the approach to outboard detachment also both decline in the same proportion for the ILW. Finally, inter-ELM W sources in the ILW divertor tend to fall with N input, although core accumulation possibly due to net ELM erosion still leads to significantly less steady conditions than in all-C plasmas. This limitation of ILW H-modes so far will be readdressed in future campaigns to continue progress towards a fully-integrated scenario suitable for D-T experiments on JET and for "baseline" operation on ITER. The diverse changes in behaviour between all-C and ILW contexts demonstrate essentially the strong impact which boundary conditions and intrinsic impurities can have on tokamak-plasma states.

ICRH was extensively used in the 2015-16 JET-ILW (ITER like wall) experimental campaign; bulk heating together with high-Z impurity chase-out from plasma centre importantly contributed to the good DD fusion performance obtained recently in JET. Power up to 6 MW was launched in H-mode deuterium plasmas and 8 MW during the hydrogen campaign. The ILA was re-installed and contributed positively to the availability of ICRH power. The ILA produces slightly less high-Z impurities than the A2's and the PWI measured via Be line emission on limiters is in the same ballpark. Specific experiments were conducted to optimise ICRH scenarios in preparation for DT in particular the dual frequency scheme, (H)D and (He)D were tested. In addition, it was confirmed that the (D)H scenario is accessible in a ILW environment and the novel 3-ions ICRH scheme was validated experimentally. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053 and from the RCUK Energy Programme [grant number EP/P012450/1]. The views and opinions expressed herein do not necessarily reflect those of the European Commission 'To obtain further information on the data and models underlying this paper please contact PublicationsManager@ukaea.uk.

The 'European Transport Simulator' (ETS) (Coster et al 2010 IEEE Trans. Plasma Sci. 38 2085-92, Kalupin et al 2011 Proc. 38th EPS Conf. on Plasma Physics (Strasbourg, France, 2011) vol 35G (ECA) P. 4.111) is the new modular package for 1D discharge evolution developed within the EFDA Integrated Tokamak Modelling (ITM) Task Force. It consists of precompiled physics modules combined into a workflow through standardized input/output data structures. Ultimately, the ETS will allow for an entire discharge simulation from the start up until the current termination phase, including controllers and sub-systems. The paper presents the current status of the ETS towards this ultimate goal. It discusses the design of the workflow, the validation and verification of its components on the example of impurity solver and demonstrates a proof-of-principles coupling of a local gyrofluid model for turbulent transport to the ETS. It also presents the first results on the application of the ETS to JET tokamak discharges with the ITER like wall. It studies the correlations of the radiation from impurity to the choice of the sources and transport coefficients.

During the installation of the new ITER-like wall (ILW) at JET, the high frequency pellet injector has been further improved. The launching system is now capable of delivering reliable fuelling size pellets from the magnetic outboard side up to 15 Hz repetition rate. Pacing size pellets can be produced at rates up to 50 Hz but pellet trains suffer some losses during the transfer to the plasma. A significant fraction of the pellet train can arrive at the plasma when launched from the outboard, while only a few pellets make it to the vessel inboard launching site. Stable and reliable edge-localized mode (ELM) control was achieved when using outboard fuelling size pellets. This tool was successfully applied for scenario development purposes in the ITER baseline H-mode scenario at 2.5 MA. Employed for ELM sustainment and impurity control, pellets prevented the ELM frequency from becoming so low as to cause a radiative collapse of the discharge. Despite technical limitations, injecting outboard pacing size pellets resulted in a transient enhancement of the initial ELM frequency up to a factor 4.5. This could be achieved in cases where a continuous train of sufficiently large and fast pellets were arriving in the plasma at a frequency of up to 31 Hz. Pacing size pellets were also used to investigate the ELM trigger threshold. Three basic parameters could be identified for outboard pellet launch. The ELM triggering probability increased with (i) the time elapsed since the previous ELM occurred, (ii) pellet mass and (iii) pellet speed. An indication for dependence of the ELM trigger threshold on the poloidal pellet launch position has been found; inboard launched pellets seem to reveal a higher trigger capability than pellets launched from the outboard. Finally, we compared the pellet penetration depth required for ELM triggering in the actual JET configuration with plasma-facing components to similar previous experiments performed with a carbon wall. This comparison indicates that pellet ELM triggering requires deeper penetration in the ILW configuration. © 2013 IAEA, Vienna.

The scan of ion cyclotron resonant heating (ICRH) power has been used to systematically study the pump out effect of central electron heating on impurities such as Ni and Mo in H-mode low collisionality discharges in JET. The transport parameters of Ni and Mo have been measured by introducing a transient perturbation on their densities via the laser blow off technique. Without ICRH Ni and Mo density profiles are typically peaked. The application of ICRH induces on Ni and Mo in the plasma centre (at normalized poloidal flux ρ = 0.2) an outward drift approximately proportional to the amount of injected power. Above a threshold of ICRH power of about 3 MW in the specific case the radial flow of Ni and Mo changes from inwards to outwards and the impurity profiles, extrapolated to stationary conditions, become hollow. At mid-radius the impurity profiles become flat or only slightly hollow. In the plasma centre the variation of the convection-to-diffusivity ratio v/D of Ni is particularly well correlated with the change in the ion temperature gradient in qualitative agreement with the neoclassical theory. However, the experimental radial velocity is larger than the neoclassical one by up to one order of magnitude. Gyrokinetic simulations of the radial impurity fluxes induced by electrostatic turbulence do not foresee a flow reversal in the analysed discharges.

A new pellet injection system was installed at JET designed for both fuelling and ELM pacing. The purpose of the pacing section was to validate pellet ELM pacing as a suitable tool for ELM mitigation in ITER. Pellet pacing was confirmed at the large size scale of JET. The dynamics of triggered ELMs was investigated with respect to their spontaneous counterparts. Triggered ELMs show features also typical for spontaneous ELMs in several operational regimes. Since none of these regimes was unsettled by the pellets this is a strong hint for compatibility with other plasma control tools. Observations and modelling results indicate the ELM triggering occurs by the evolution of the pellet ablation plasmoid into the first ELM filament followed by a poloidal spread of the instability. An ELM obviously can be forced by a pellet due to the strong local perturbation imposed already under unusual onset conditions but then evolves like any ELM typical for the corresponding plasma regime. For tool optimization the pellet mass and hence the convective confinement losses imposed have to be minimized. In our experiments, a lower mass threshold was observed for the first time. It has been found that to reliably trigger an ELM the pellet needs to be sufficiently large (and fast) to penetrate close to the pedestal top. Recent investigations are clear steps forward to validate the pellet pacing approach for ITER.

New FTU ohmic discharges with a liquid lithium limiter at I P = 0.7–0.75 MA, B T = 7 T and n e0 ⩾ 5 × 1020 m−3 confirm the spontaneous transition to an enhanced confinement regime, 1.3–1.4 times ITER-97-L, when the density peaking factor is above a threshold value of 1.7–1.8. The improved confinement derives from a reduction of electron thermal conductivity (χe) as density increases, while ion thermal conductivity (χi) remains close to neoclassical values. Linear microstability reveals the importance of lithium in triggering a turbulent inward flux for electrons and deuterium by changing the growth rates and phase of the ion-driven turbulence, while lithium flux is always directed outwards. A particle diffusion coefficient, D ∼ 0.07 m2 s−1, and an inward pinch velocity, V ∼ 0.27 m s−1, in qualitative agreement with Bohm–gyro-Bohm predictions are inferred in pellet fuelled lithized discharges. Radio frequency heated plasmas benefit from cleaner plasmas with edge optimized conditions. Lower hybrid waves penetration and current drive effects are clearly demonstrated at and above ITER densities thanks to a good control of edge parameters obtained by plasma operations with the external poloidal limiter, lithized walls and pellet fuelling. The electron cyclotron (EC) heating system is extensively exploited in FTU for contributing to ITER-relevant issues such as MHD control: sawtooth crash is actively controlled and density limit disruptions are avoided by central and off-axis deposition of 0.3 MW of EC power at 140 GHz. Fourier analysis shows that the density drop and the temperature rise, stimulated by modulated EC power in low collisionality plasmas are synchronous, implying that the heating method is the common cause of both the electron heating and the density drop. Perpendicularly injected electron cyclotron resonance heating is demonstrated to be more efficient than the obliquely injected one, reducing the minimum electric field required at breakdown by a factor of 3. Theoretical activity further develops the model to interpret high-frequency fishbones on FTU and other experiments as well as to characterize beta-induced Alfvén eigenmodes induced by magnetic islands in ohmic discharges. The theoretical framework of the general fishbone-like dispersion relation is used for implementing an extended version of the HMGC hybrid MHD gyrokinetic code. The upgraded version of HMGC will be able to handle fully compressible non-linear gyrokinetic equations and 3D MHD.

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation. © 2009 IAEA, Vienna.

In the Reversed Field Pinch (RFP) configuration the plasma can access regimes in which a single tearing mode dominates the magnetic perturbations spectrum (Quasi Single Helicity state, QSH) [1]. Thermal measurements performed in the RFX-mod experiment [2] (Padua, Italy) during QSH states have shown clear helical structures, identified by significant temperature gradient in the plasma core, which can reach a considerable size (25-50% of plasma radius) and temporal persistance up to 85% the flat top phase (~200ms) of the discharge [3]. In this work we study some aspects of the transport mechanisms relevant to these conditions: on the one hand the impact of residual magnetic chaos, on the other hand the effect of electrostatic micro-instabilities, the latter have been overlooked in the past when the RFP configurations where frequently dominated by turbulent chaotic regimes (so-called MH regimes). We shall here consider the best perfomance discharges (plasma current Ip ~1.0-1.3MA and electron density ne ~2-4?1019 m-3) characterized by QSH regimes with separatrix expulsion from the helical core [4].

As in Tokamaks RFX-mod discharges are subject to the Greenwald density limit. The similarity between Reversed Field Pinches (RFP) and Tokamaks extends to many properties of the edge transport physics, including the electrostatic nature of turbulence, the presence of highly sheared ExB flows - sustained by the Reynolds' stress - and the presence of coherent structures emerging out of the turbulent background. The origin of the density limit in RFX-mod is sought along the legitimate idea that it should be similar to the Tokamak case. It is proposed that the poloidally symmetric and toroidally asymmetric strong radiation emission that develops where the plasma interacts with the wall, at high densities triggers a thermal instability that cools the plasma core. The process resembles the evolution of a MARFE in a Tokamak, but in the RFP the effect on the plasma core occurs well before the total radiated power approaches the (ohmic) power input and the result is a soft landing of the discharge.

In reversed-field pinch (RFP) devices the magnetic configuration is characterized by a toroidal magnetic field whose direction in the plasma edge is opposite that in the core. With an edge safety factor qa? - 0.02 ( q = aBt/RBp), the toroidal magnetic field in the RFP is much smaller than in the Tokamak and stability is ensured by the large magnetic shear (r/q dq/dr). Since in the RFP the safety factor is everywhere less than one, a large number of MHD modes are resonant and their amplitude largely affect the confinement. In RFX, as usual in RFP devices, MHD modes amplitude had been controlled by the use of a passive thick conductive shell that vanish the radial magnetic field at the surface. In RFX-mod a different approach has been followed: radial magnetic field control is obtained by external magnetic field coils and a close fitting thin conductive shell. In the so-called Virtual Shell operation (VS), radial field zeroing at the thin shell radius is stationary provided by the externally controlled coils. First experiments of RFX-mod proved the capability of the active scheme to control the radial magnetic field at any time. Furthermore it has been found that such edge magnetic field control extends its beneficial effects to the whole plasma, leading to a stationary 2 to 3-fold reduction of the core Br field amplitude. The reduction of field fluctuations positively reflects on confinement. In fact, a strong reduction of the loop voltage is observed and correspondingly a 3-fold increase in pulse length is achieved by using the same poloidal flux swing. Temperature and particle measurements confirm the improved confinement properties of the virtual shell operation. With a lower ohmic input power, higher electron temperature in the plasma core and a steeper profile is measured. Particle and heat transport have been studied by means of a 1-d code. Local power balance was used to compute the heat conductivity profile: for the VS discharges a lower conductivity over a significant region of the plasma is found. The improved properties of RFX-mod VS operation provide a better confinement scaling both in terms of plasma current and density. The results show that compared to the thick shell solution, a significant confinement improvement can be obtained under stationary conditions by actively controlling the plasma magnetic boundary.

Pellet injection is considered the most promising technique to fuel ITER plasmas. However, standard low field side (LFS) pellet injection seems inadequate for an efficient fuelling of the plasma core because it is too demanding in terms of injection velocity. Alternative configurations, as injection from the high field side (HFS) or vertically from the top of the machine (VHFS) are technologically less demanding due to a drift of the ablated material down the magnetic field gradient, which helps its penetration. In the past year LFS and VHFS pellet injection experiments have been performed on the Frascati Tokamak Upgrade (FTU: BT = 8 T, Ip = 1.5 MA, R = 93.5 cm, a = 30 cm), a circular cross-section device with a molybdenum toroidal limiter located on the HFS. The rationale of these experiments was to study the advantages of VHFS injection in a high toroidal field machine, where the value of BT can be considered as representative of that in ITER. The results demonstrate that the main advantage of VHFS injection (with respect to LFS) is that it allows one to obtain the same post-pellet density profile with an injection speed 2–3 times lower than that required in the case of LFS injection. However no significant displacement of the ablated material was observed, possibly due to the small value of the ∇ B-induced drift in these ohmic, low β plasmas. The experimental measurements were compared to the predictions of the drift model (Pegourie B. et al 2007 Nucl. Fusion 47 44) showing a reasonable agreement in both the magnitude of the drift and the main features of the line integrated density during pellet ablution.

Plasma experiments resumed in December 2004 on RFX-mod. The machine now has a thin (3mm) Cu shell with one overlapped poloidal gap and one toroidal gap. Shell penetration time for Bv has been lowered from 450 to 50 ms and shell/plasma proximity from b/a=1.24 to 1.1. Toroidal equilibrium is feedback-controlled and new power supplies provide a better control of the toroidal field. Newly designed graphite tiles protect the vessel from highly localized power deposition. The MHD Control System, MHD-CS, a set of 192 external saddle coils controlled by a digital feedback system, is used to control radial fields due to field errors, MHD modes and Resistive Wall Modes (RWMs). A dramatic improvement of plasma performance was obtained by using the MHD-CS to cancel all of the radial field components, an operational mode dubbed Virtual Shell (VS). The toroidal loop voltage was lowered by more than 40% and the plasma pulse duration tripled. In practice, steady state RFP pulses are now limited only by the applied volt-seconds. Hence RFX-mod initial operation demonstrated the possibility to operate a large RFP without a thick conducting shell, and opened enhanced RFP scenarios. Indeed the improved magnetic boundary in VS mode, which mimics an ideal closely fitting shell, has an effect on the tearing modes underlying the sustainment of the RFP configuration, the so-called dynamo modes, which are also responsible for field line stochastization in the plasma core and confinement limitation. With the VS the amplitude of such modes in the plasma centre was nearly halved. As expected, this led to improved particle and energy confinement. For instance, peak electron temperature in reference pulses at 600 kA was increased from 200 to 300 eV with more peaked profiles, which corresponds to a reduction of the thermal conductivity by a factor 2 in the region r/a < 0.9. The MHD-CS is extremely flexible and can be used for a variety of mode control experiments. The most important result already obtained was the demonstration of the active control of RWMs. We found that full VS control completely inhibits the growth of RWMs, whereas such modes are indeed seen to grow in agreement with the theoretical prediction if the MHD-CS operated in Selective VS mode, i.e. leaving one or more mode helicity uncontrolled.

Carbon and nitrogen impurity transport coefficients are determined from gas puff experiments carried out during repeat L-mode discharges on the Mega-Amp Spherical Tokamak (MAST) and compared against a previous analysis of helium impurity transport on MAST. The impurity density profiles are measured on the low-field side of the plasma, therefore this paper focuses on light impurities where the impact of poloidal asymmetries on impurity transport is predicted to be negligible. A weak screening of carbon and nitrogen is found in the plasma core, whereas the helium density profile is peaked over the entire plasma radius., Comment: 17 pages, 7 figures

Experiments were performed on JET where high-density plasmas with an internal transport barrier (ITB) were created by means of combined use of lower hybrid current drive (LHCD) and pellet injection before the barrier formation. Attempts were also made to use pellets to fuel the plasma and to sustain the density during the ITB phase. It was found that shallow pellets ablating in the region r/a ≥ 0.8 and far from the foot of the barrier did not destroy the ITB, whereas deeper pellets penetrating up to 0.6 ≤ r/a ≤ 0.7 affected the barrier and led to its disappearance. Modelling of these experimental scenarios has been performed with transport and fluid turbulence codes. The codes used in the analysis were: JETTO, a 1.5 dimensional transport code, TRB, a global electrostatic fluid turbulence code and CUTIE, a global electromagnetic fluid turbulence code. The results show that for the shallow pellet case all codes reproduce the general features of the experiment, whereas for the deep pellet case, there are differences in the degree of agreement between the different codes and the experiment. Runs performed varying the pellet penetration depth indicate that not only the pellet penetration, but also the barrier strength plays a key role in the dynamics of the pellet-ITB interaction.

A neutral gas and plasma shielding model is presented that describes the interaction of a pellet with the high energy ions and electrons generated during heating or current drive experiments. The main improvements are the self-consistent calculations of the electrostatic sheath at the cloud–plasma interface and of the extra ablation due to the fast tail of the electron and ion distributions, including heating in the volume of the pellet. With regard to the comparison between the code predictions and the experimental results, realistic three-dimensional (space, energy, pitch angle) distributions have been used for both the ions and electrons. For ohmic discharges, the code has been tested on more than 40 well-documented pellets selected in the International Pellet Ablation DataBASE. For additionally heated plasmas (ion cyclotron resonance heating—minority regime—and lower hybrid current drive, 2–4 MW of injected power), Tore Supra data have been used. In these different cases, the calculations are in good agreement with the experimental penetrations and ablation profiles. A parametric study is also presented, which enlightens the control parameter that governs the pellet penetration. In what concerns the capability of pellet injection to fuel reactor grade plasmas, it is shown that no strong extra ablation due to the α-particles is expected.

EX/P4-13 · The Greenwald density limit in the Reverse Field Pinch M. Valisa1), F. Auriemma, A. Canton, L. Carraro R. Lorenzini, S. Ortolani, R. Paccagnella, M.E. Puiatti, F. Sattin, P. Scarin, G. Spizzo, M. Spolaore, N. Vianello, A. Alfier, V. Antoni, L. Apolloni, M. Bagatin, W. Baker, O. Barana, F. Basso, D. Bettella, P. Bettini, T. Bolzonella, D. Bonfiglio, A. Buffa, S. Cappello, R. Cavazzana, M. Cavinato, G. Chitarin, A. Cravotta, F. D'Angelo, S. Dal Bello, M. Dalla Palma, A. De Lorenzi, L. De Pasqual, D. Desideri, D.F. Escande, P. Fiorentin, P. Franz, L. Frassinetti, G. Gadani, E. Gaio, L. Garzotti, E. Gazza, L. Giudicotti, F. Gnesotto, L. Grando, S.C. Guo, P. Innocente, A. Luchetta, G. Malesani, G. Manduchi, G. Marchiori, D. Marcuzzi, L. Marrelli, P. Martin, E. Martines, S. Martini, A. Masiello, F. Milani, M. Moresco, A. Murari, P. Nielsen, R. Pasqualotto, S. Peruzzo, R. Piovan, P. Piovesan, N. Pomaro, G. Regnoli, G. Rostagni, V. Schmidt, G. Serianni, P. Sonato, E. Spada, C. Taliercio, D. Terranova, V. Toigo, P. Zaccaria, P. Zanca, B. Zaniol, L. Zanotto, E. Zilli, G. Zollino, M. Zuin 1) Consorzio RFX - Associazione EURATOM-ENEA sulla Fusione, Padova, Italy Abstract: In the Reversed Field Pinch RFX the density operating space exhibits an upper boundary that is well described by the Greenwald limit. The basic findings are described and similarities and differences with the Tokamak case are analysed. The high-density limit occurs as a non-disruptive limit. Fast terminations have been observed but in a wide area of the operating space and therefore are unlikely to be related to the density limit. Radiation induced thermal collapse can also be ruled out being radiation losses always a relatively small fraction of the ohmic input power. The occurrence of plasma detachment while approaching the density limit can be excluded. The absence of additional heating makes it difficult to distinguish with certainty between a power balance effect and an intrinsic physical limit but the overlapping of the Greenwald limit with the upper boundary of the density operating space is however remarkable, especially in the case of He plasmas . While in Tokamaks one of the plausible causes for the limit is a transport induced edge thermal instability, in RFX, as density increases the density profile becomes hollow, inside the last closed magnetic surface particle diffusion decreases and the global energy confinement time improves. The normalised density fluctuations measured by Langmuir probes and by the outermost chords of an interferometer do not increase. Simulations of the hydrogen discharges with the RITM code confirm the importance of recycling in determining the edge density gradient and the minor role of radiation losses. Differences between H and He cases are analysed in terms of particle penetration capability and in terms of edge ExB shear. A statistical analysis of the edge density fluctuations is presented, looking for differences arising as density is increased. Finally, the possible role of pressure driven modes is discussed.

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 reference regime (n/n(GW) similar to 0.85, H-98(y,H-2) similar to 1, beta(N) similar to 1.8). ELMy H-mode plasmas have been realized reaching or even exceeding those parameters in steady-state conditions (up to similar to5 s or 12tau(epsilon)) in a reproducible way and only limited by the duration of the additional heating phase. These results have been obtained (a) in highly triangular plasmas, by increasing the average triangularity delta towards the ITER reference value (delta similar to 0.5), and (b) in plasmas at low triangularity (delta similar to 0.2) by seeding of Ar and placing the X-point of the plasma on the top of the septum. Pellet injection from the high field side is a third method yielding high density and high confinement, albeit not yet under steady-state conditions. In highly triangular plasmas the influence of input power, plasma triangularity and impurity seeding with noble gases has been studied. Density profile peaking at high densities has been obtained in (a) impurity seeded low triangularity discharges, (b) ELMy H-modes with low levels of input power and (c) discharges fuelled with pellet injection from the high field side. New ELM behaviour has been observed in high triangularity discharges at high density, opening a possible route to ELM heat load mitigation, which can be further amplified by Ar impurity seeding. Current extrapolations of the ELM heat load to ITER show possibly a window for Type I ELM operation. Confinement scaling studies indicate an increase in confinement with triangularity and density peaking, and a decrease in confinement with the Greenwald number. In addition, experiments in H isotope and He indicate tau(E) proportional to M(0.19)Z(-0.59). The threshold power for the L-H transition in He plasmas shows the same parametric dependence as in D plasmas, but with a 50% higher absolute value.

The results of an extensive investigation of electrostatic and magnetic turbulence in the edge region of two Reversed Field Pinch (RFP) experiments EXTRAP-T2R and RFX are reported . In both experiments particle transport is mostly driven by electrostatic turbulence and a highly sheared ExB flow takes place. Recent results have shown that almost 50% of the particle losses is due to coherent structures. These structures have features reminiscent of monopolar or dipolar vortices and their relative population and preferred vorticity is determined by the local ExB shear. It has been demonstrated that the plasma diffusivity D can be separated in two comparable terms: one due to coherent structures and another one due to background turbulence. The ExB shear results to affect both terms, modifying the populations of vortices and the phase of background density and velocity fluctuations . This effect has been proved by several experiments of transport control based on modification of the ExB shear by insertion of active electrodes or by applying transient modifications of the magnetic field. In order to investigate the process responsible for the observed ExB flow profile, the momentum balance has been recently addressed and it has been found that also transport of momentum is anomalous as experimental kinematic viscosity results consistent with anomalous diffusion. Moreover the balance reveals that plasma flow profile is regulated by turbulence and in particular by the electrostatic component of the Reynolds Stress (RS). These results prove the existence of a dynamic interplay between turbulence properties, anomalous transport and mean profiles. All results are discussed highlighting the similarities with other magnetic configurations.

Low collisionality, low particle source. ELMy H-modes (type-III) with sawteeth are produced in JET in order to address the question of density profile evolution in the reference q(95) = 3 ITER scenario. The paper focuses on particle transport in the core zone around the mid-radius. The pedestal region including ELMs and the region affected by sawteeth are not considered. By replacing a significant part of the neutral beam heating by RF power the beam particle flux at mid-radius has been reduced to Gamma(Beam)/n(e) = 0.07 m s(-1). The additional flux due to wall neutrals is estimated as Gamma(Wall)/n(e) = 0.12 m s(-1). Density profiles are found to be modestly peaked under these conditions with a relative density difference of Deltan/(n) = 0.23 across the zone not affected by sawteeth and ELMs. In a region around the mid-radius the ratio of effective particle diffusivity to electron thermal diffusivity is found to be D-e,D-eff/chi(e) approximate to 0.2, which might indicate an anomalous pinch provided the particle diffusivity D-e is sufficiently lame. The measured values of D-e,D-eff/chi(e) are at the lower end of the range used in ITER models.

In view of the installation of a multiple pellet injector on the RFX experiment, an integrated diagnostic system has been designed to monitor the pellet behavior in the plasma. The evidence in the reversed field pinches (RFPs) of large poloidal and toroidal pellet deflections along magnetic field lines due to asymmetric ablation, calls for a 3D system with good space and time resolution. Two arrays of H? detectors will view the trajectories of the pellets from below and from behind and a digital charged coupled device (CCD) camera will view the injection poloidal section from a tangential viewpoint. The two systems will allow the complete determination of the time-resolved trajectory of each pellet. The space integrated H? emission rate will be measured by a large angle detector and by a space integration of the CCD camera measurement. © 1995 American Institute of Physics.

Over the last two years, several experiments relevant for the study of particle transport and density profile evolution, have been performed at JET. They can be classified as stationary discharges with and without central particle source due to the beams, quasi-stationary discharges with deuterium gas puffing, deep pellet fuelled discharges and discharges perturbed by cold pulses obtained by shallow pellet injection. All these experimental scenarios have been simulated by means of the JETTO transport code, employing different transport models: purely empirical models and the semi-empirical mixed Bohm/gyro-Bohm transport model, both with the addition of different theory-based expressions for the anomalous particle pinch and the first principle Weiland transport model. The coefficients used to scale the pinch velocity in the purely empirical and in the mixed Bohm/gyro-Bohm model have been varied from shot to shot. In this paper, the results of the simulations are presented. The main conclusions are that, for the cases studied in this paper, the sawtooth activity is the main particle transport mechanism in the plasma centre (r/a ≤ 0.5). Nevertheless, to reproduce the density profile in the gradient zone (0.5 ≤ r/a ≤ 0.9), an anomalous pinch seems to be necessary, at least for L-mode plasmas. This anomalous convective flux is well reproduced by the off-diagonal elements of the transport matrix given by the Weiland model.

An overview of the Frascati tokamak upgrade (FTU) results during the period 2000-2002 is presented. Long duration internal transport barriers (ITBs) have been obtained on FTU with combined injection of lower hybrid (LH) and electron cyclotron (EC) waves in 5 T/0.5 MA discharges. The ITB phase lasts about 10 energy confinement times and is characterized by an energy confinement time up to 1.6 times the ITER97 L-mode scaling. Temperatures up to I I keV are measured at 0.9 x 10(20) m(-3) central density. ITB studies using IBW injection have also been continued up to 8 T/0.8 MA. The LH system has operated at full power allowing us to complete the current drive (CD) studies at ITER-relevant densities. At these density values the electrons and ions are coupled and an increase in the ion temperature is clearly observed. A preliminary sign of enhanced CD efficiency has been obtained in the combined injection of EC and LH waves at magnetic field values lower than the resonant field for thermal EC absorption. Pellet optimization studies have been performed in order to test the conditions under which a quasi-steady-state confinement improvement can be obtained and impurity accumulation can be avoided. The confinement time in Ohmic discharges is generally in agreement with the ITER97 L-mode scaling. Transient confinement improvement is observed for a duration less than one energy confinement time. Radiation improved mode studies have been started thanks to the recently inserted boronization system, which has allowed reduction of the radiated power. Confinement improvement with neon injection has been observed in 6 T/0.9 MA discharges. Transport studies on profile stiffness and magnetohydrodynamic (MHD) studies of fast reconnection and snakes will also be presented.

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High field pellet fuelled discharges in FTU have achieved good energy and particle confinement properties in ohmic quasi-steady regime. Peak densities close to 1x1021 m-3 have been obtained, showing a post pellet decay time of several energy confinement times. Experiments aimed at obtaining Radiative Improved Modes (RIM) at high field have also been carried out using Neon injection. We have observed an increase of average density with respect to the reference discharge, which cannot be explained by the contribution of Neon. The neutron yield increases also by a factor 3- 6 and the energy confinement time increases by a factor 1.4

In this paper we analyze the effect of Pulsed Poloidal Current Drive (PPCD) on the profiles of magnetic field, electron and impurity densities and electron temperature in RFX. 3D MHD non-linear simulations and 1D time dependent transport codes show that the external inductive drive pinches and peaks the current profile driving the configuration through a transient phase, where the spontaneous dynamo action is quenched and consequently the magnetic fluctuations and the transport are reduced.

In this paper, we analyse the main features of the pulsed poloidal current drive (PPCD) technique, used in the reversed field pinch configuration to achieve improved confinement conditions. In the RFX experiment, PPCD corresponds to a decrease of the magnetic fluctuations, to a peaking of the temperature profile, and to a reduced transport and plasma-wall interaction. A three-dimensional MHD nonlinear code and one-dimensional time-dependent transport models have been applied to study the effect of PPCD on the magnetic and plasma profiles. The three-dimensional MHD simulations show that the external inductive drive pinches and peaks the current profile driving the configuration through a transient phase, where the spontaneous turbulent dynamo action is quenched. The one-dimensional transport codes indicate that the experimental profile modifications associated with PPCD are consistent with a reduction of the stochastic transport.

The paper highlights the progress made in the modelling and understanding of the mechanisms underlying particle and energy transport in the RFP. A transport analysis has been performed on the perturbed transport coefficients for pellets of different characteristics launched both in standard RFP plasmas and in combination with pulsed poloidal current drive, this latter scenario potentially allowing an efficient fuelling and a good energy confinement. The study has confirmed the importance of large-scale MHD modes in determining the transport properties in the RFP core. As far as confinement in the edge is concerned, the analysis has been focussed on the role played by burst events on the electrostatic particle flux and to the 'intermittent events'. The intermittent events cluster during relaxation processes and such behaviour is more evident close to the magnetic reversal surface. This result confirms an interplay between the small scale edge turbulence and the large scale relaxation events driven by core-resonant tearing modes.

Particle transport in magnetized plasmas is investigated with a fluid model of drift wave turbulence. An analytical calculation shows that magnetic field curvature and thermodiffusion drive an anomalous pinch. The curvature driven pinch velocity is consistent with the prediction of turbulence equipartition theory. The thermodiffusion flux is found to be directed inward for a small ratio of electron to ion pressure gradient, and it reverses its sign when increasing this ratio. Numerical simulations confirm that a turbulent particle pinch exists. It is mainly driven by curvature for equal ion and electron heat sources. The sign and relative weights of the curvature and thermodiffusion pinches are consistent with the analytical calculation.

A pellet penetrating the inner region of a tokamak discharge, where the safety factor drops below unity, triggers an instability analogous to a sawtooth crash. Because of the simultaneity of the crash and pellet crossing, the latter is an appropriate probe for investigating the current distribution during reconnection. In this Letter, pellet deflection is used to characterize the associated electron distribution function. The perturbation compatible with the observed trajectory requires a negative current layer on the q = 1 magnetic surface between 3 and 12 times the equilibrium current density and an expulsion of high energy electrons from the plasma core.

Perturbative transport experiments have been performed in JET low or reverse magnetic shear plasmas either in conditions of fully developed internal transport barrier (ITB) or during a phase where an ITB was not observed. Transient peripheral cooling was induced either by laser ablation or by shallow pellet injection, and the ensuing travelling cold pulse was used to probe the plasma transport in the electron and, for the first time, also in the ion channel. Cold pulses travelling through ITBs are observed to erode the ITB outer part, but, if the inner ITB portion survives, it strongly damps the propagating wave. The result is discussed in the context of proposed possible pictures for ITB formation. In the absence of an ITB, the cold pulse shows a fast propagation in the outer plasma half, which is consistent with a region of stiff transport, while in the inner half it slows down but shows the peculiar feature of amplitude growing while propagating. The data are powerful tests for the validation of theoretical transport models. This article was scheduled to appear in issue 7 of Plasma Phys. Control. Fusion. To access this Special issue please follow this link: http://stacks.iop.org/0741-3335/44/i=7

Pellet injection experiments have been performed on the RFX reversed field pinch (RFP). The main aim of the experimental campaign was the study of the target plasma response to get information both on the possibility of an efficient pellet fuelling of the discharge and on the associated transport modification. In this paper, we present an analysis of different experimental scenarios. The tool employed is a one-dimensional transport code coupled with an ablation calculation that yields the pellet particle source and trajectory. It is seen that, in standard plasmas, the pellets that reach the plasma centre trigger a phase of rapid profile flattening and poorer particle and energy confinement, and an increase of the transport coefficients is required to reproduce the post-pellet behaviour of the profiles. On the other hand, the slow pellets that suffer large deflection and miss the plasma centre do not have such a deleterious effect and, in this case, the unperturbed transport coefficients are able to reproduce the post-pellet evolution of the profiles. Finally, when the pellets are launched in combination with some method that induces an improved confinement, the transport degradation is less evident. In our case, pellet injection has been combined with pulsed poloidal current drive (PPCD), a technique that reduces magnetic field stochasticization and transport in the RFP core. When the effect of the PPCD is strong enough, a synergy between pellet injection and PPCD is observed and plasma fuelling without dramatic confinement reduction is possible.

The results of an extensive study on transport mechanisms and on improved confinement scenarios in RFX are reported. The scaling of the thermal conductivity in the core with the Lundquist number indicates that the magnetic field in this region is not fully stochastic, as proved by the existence of thermal barriers observed in single helicity configurations. The electrostatic transport at the edge has been proved to depend on the highly sheared E × B flow, which has been interpreted using fluid and Monte Carlo models. Regimes of improved confinement have been obtained in the core by poloidal current drive techniques, and the electrostatic transport has been reduced at the edge by biasing experiments. A radiation mantle caused by impurity seeding has been found to successfully reduce the local plasma-wall interaction without causing a significant deterioration in the plasma performance.

The experimental and theoretical activity performed in the RFX device has allowed a deeper insight into the MHD properties of the reversed field pinch (RFP) configuration. A set of successful experiments has demonstrated the possibility of influencing both the amplitude and the spectrum of the magnetic fluctuations which characterize the RFP configuration. A new regime (quasi-single-helicity states) where the dynamo mechanism works in a nearly laminar way and a helical core plasma is produced has been investigated. With these studies a reduction of magnetic chaos has been obtained. The continuous rotation of wall locked resistive tearing modes has been obtained by an m = 0 rotating perturbation. This perturbation induces rotation of m = 1 non-linearly coupled modes.

In reversed-field pinches, two different mechanisms have been proposed to explain the dynamo process which drives the poloidal current needed to sustain the magnetic configuration: the kinetic dynamo theory and the magnetohydrodynamic (MHD) dynamo theory. Experimentally, they can be distinguished by the radial behavior of the electron distribution function. In this Letter the trajectory deflection of frozen hydrogen pellets has been used as a diagnostic of suprathermal electrons in the plasma. The classical Spitzer-Härm distortion of the electron distribution function consistent with the MHD dynamo electric field is found to give a better modeling of the pellet trajectory.

Fast visible imaging of Mega-Ampere Spherical Tokamak plasmas during injection of refueling pellets of frozen deuterium can sometimes vividly reveal their internal magnetic structure. Concentrated deposition of particles around the flux surface with unit value of field-line helicity results in locally increased bremsstrahlung emission which illuminates this whole surface or, occasionally, a single flux tube within it. Analysis of such images could be exploited to check equilibrium reconstruction models.

The particle and energy transport mechanisms have been investigated in RFX. In the core the particle diffusivity and thermal conductivity are consistent with a stochastic magnetic field model. At the edge particle transport is mainly driven by electrostatic fluctuations, whereas radiation, electrostatic and magnetic turbulence account only for 50% of the total energy losses.