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Following the report of Fleischmann and Pons,(1) we (The Bavarian Bubble Bottle Team) have attempted to reproduce their claims of cold nuclear fusion, and failed. We note that our measurements would not be able to detect neutrons at the level of Joneset al.(2) Three electrolytic cell experiments were conducted using palladium cathodes and Platinum anodes, in a 0.1-Molar solution of LiD in heavy water, without any signs of neutrons, tritium, or gammas above backgrounds, and within ±0.3 watt accuracy calorimetry, no excess heating. Excess heating at the levels of F&P would have been easily detected, if present. Intrinsic tritium, differing from each D2O bottle tested, was however observed. The longest duration experiment ran for 21 days, and was an attempt to duplicate the large “melting incident” of F&P. This was terminated on April 28, 1989, by throwing the vacuum-cast 22 gram, deuterium-loaded palladium cathode directly into liquid nitrogen, immediately next to a bare BF3 counter (backed by 25 cm of moderator), in order to attempt one of the Italian ENEA neutron production variants. No neutrons above backgrounds were seen, while counting for 1 hour, and also none while the piece warmed to room temperature over the next hour. Post-mortem analysis of the darkened, hardened Pd piece showed large crystal grains (up to 2 mm × 2 mm), and continuing evolution of gas bubbles at the grain boundaries even days after the experiment was ended. eight weeks after loading, the catalytically-active palladium piece continued to create heavy water (with exposure to oxygen in the air).

The paper summarizes the experiments performed with ion cyclotron resonance heating (ICRH) on ASDEX, from November 1984 until March 1986; the most interesting results are reported and discussed in detail. Heating and confinement studies using the hydrogen second harmonic scheme and the hydrogen minority scheme (PIC < 2.6 MW, tIC < 1.5 s) show a typical L-mode behaviour, i.e. a power dependent confinement degradation, which is rather similar to that found with neutral beam injection (NBI) heating. ICRH is accompanied by a slightly improved particle and energy confinement compared with that of NBI; this is also true for a combined ICRH + NBI scheme, up to Ptot ≈ 4.5 MW, absorbed in the plasma. Particular efforts have been devoted to investigations of the second harmonic regime in H/D plasmas with nH/ne ≈ 0.1 - 1, with a view to heating mixtures in reactor relevant plasmas. The achievement of H-mode transitions with ICRH alone in the hydrogen minority scheme at an absorbed RF power of about 1.1 MW supports the assumption of common confinement properties in auxiliary heated tokamaks, since they appear to be widely independent of the additional heating method. ICRH specific impurity problems, such as the strong release of iron from the vessel walls, have been overcome by applying extensive in situ wall carbonization. The mechanisms responsible for impurity generation have partly been identified and analysed; however, the problem still remains to be solved. Impurities preferentially released from the ICRH antenna do not pose problems.

Pulsator was operated by IPP Garching from 1973 to 1979. It made two important contributions to international tokamak research. 1. In Pulsator, for the first time a tokamak plasma was influenced locally by an external, resonant, helical field. The accompanying formation of islands on magnetic surfaces, which affects transport and stability, provided the starting point for experimental investigation of the disruptive instability and its theoretical description. 2. By gas puffing during the discharge it was possible to attain a high-density regime which dramatically raised the then prevailing nτE values by almost two orders of magnitude. In this high-density regime an inward directed particle drift velocity and, as a consequence, under certain conditions an accumulation of impurities in the plasma core was found.

The highest β-values derived from diamagnetic flux measurements in H-discharges are found to be close to the β-limit due to kink and ideal ballooning modes. Both energy and particle confinement are degraded in these discharges and do not recover during the injection period. The energy confinement time decreases by a typical factor of two, the electron thermal diffusivity being correspondingly enhanced. Electron heat conduction is again found to be the dominant energy loss channel. The degradation of confinement close to the ideal-MHD limit suggests that kink and ballooning instabilities occur and are responsible for the enhanced transport. This conclusion is supported by the correlation with the pressure profile and by the result that global and local transport do not change in a reference discharge which is stable against ideal-MHD modes.

An instrumented scoop limiter probe is being operated on TEXT to test the concept of limiter cooling and improved particle removal efficiencies using an externally-applied resonant magnetic field perturbation (the resonant helical divertor concept). Cooling of the limiter face has been demonstrated for limiter positions ranging from r L = 29.0 cm inward to r L = 25.5 cm (the Text primary poloidal hoop limiter radius r a = 27.0 cm). Pressure rises in the limiter throat of approximatel 40% are observed under optimized conditions. Interchangable limiter heads with thicknesses of 1.0 cm and 0.3 cm have been used to examine particle ducting into the scoop aperture. Experimental results are discussed along with observations of the limiter floating potential, H α recycling emissions, pressure measurements, and edge density and temperature measurements.

Investigations of impurity accumulation phenomena in ASDEX are reviewed. There are four different operating regimes where pronounced accumulation is observed and these regimes are also characterized by improved energy confinement. In particular, medium-Z metallic ions are involved in accumulation processes whereas low-Z ions appear almost unaffected. The rapid accumulation observed in the case of metallic ions may be explained by neoclassical inward drifts if we assume that the anomalous diffusion is sufficiently suppressed, some indication of this being found from laser blow-off studies. The present results, however, can only be partly explained by neoclassical theory, according to which accumulation of low-Z impurities should also occur. The temporal behaviour of accumulation and the retarding effect of proton dilution for collision dominated transport are also discussed. Finally, we conclude that the full benefits of improved energy confinement can be achieved only if the impurity influxes are kept to a sufficiently low level. Expressed in terms of concentrations under low confinement conditions we have to postulate, for ASDEX, concentrations ≲ 10−4 for metals and ≲ 2% for all light impurities.

The H-mode transition can lead to a rapid increase in tokamak plasma confinement. A semiempirical transport model was derived from global OH and L-mode confinement scalings and then applied to simulation of H-mode discharges. The radial diffusivities in the model depend on local density and pressure gradients and satisfy an appropriate dimensional constraint. Examples are shown of the application of this model and of similar models to the detailed simulation of two discharges which exhibit an H-mode transition. The models reproduce essential features of plasma confinement in the Ohmic heating and the low- and highconfinement phases of these discharges. In particular, the evolution of plasma energy content through the H-mode transition can be reproduced without any sudden or ad hoc modification of the plasma transport formulation.

A toroidal large area limiter was installed in the divertor tokamak ASDEX. The limiter basically consists of a plane annular stainless steel sheet which is tangential to the plasma and which is placed in the bottom of the main plasma chamber. In a second series of experiments this limiter was covered with graphite plates. With the steel limiter clean discharges could not be obtained. Energy deposition on the limiter was only 12% of the input. With the graphite limiter Z eff was below 3, energy deposition on the limiter was 25% (up to 40% in low density operation). Energy deposition was toroidally symmetrical. Results, including some experiments with a single null divertor are presented.

Discharges fuelled by stationary pellet injection (PI), gas puffing (GP) or a combination of the two methods are compared with respect to recycling in the divertor and particle confinement. Fuelling by PI yields much better global particle confinement than by GP. This has been found for both low and high recycling. In the low-recycling case this improvement is due to the deeper particle deposition for PI than for GP since the transport in the inner plasma is not reduced. For high recycling the improvement results from both the deeper deposition and a reduction in the transport. The best global particle confinement was found for phases with low or no GP. This, however, can be reached for short times only. Since with PI alone it is impossible to keep the recycling on a high level, GP is unavoidable for sustaining the favourable high-recycling condition.

Neutral particle fluxes onto the walls of ASDEX have been investigated using a time-of-flight (TOF) method. The energy distributions of the neutrals could be determined in the range of 10–1000 eV/amu. Ohmic divertor and limiter discharges with equal plasma currents and densities have been compared for He and D2. The He0 outflux at ∼2000 eV from He discharges is 110 of the corresponding D0 flux in D2 discharges. At lower energies this difference is much smaller. In all cases many more He neutrals were observed than was anticipated from the CX rate-coefficients for He2+. The impurity fluxes due to sputtering by the CX-neutrals show no significant difference for He and D2 discharges. For divertor discharges CX-sputtering can fully account for the Fe impurity content determined spectroscopically.

The fueling efficiency for gas puffing, i.e. the fraction of the external gas flux that is ionized inside the separatrix, is reduced in divertor discharges since part of it is ionized in the scrape-off layer and pumped off by the divertor. The fueling efficiency is determined by switching-off the gas feed during the stationary phase of a discharge and dividing the time derivative of the total number of particles inside the separatrix by the external gas flux. The determination of this time derivative must take into account profile changes. In ASDEX the fueling efficiency ranges from close to 1.0 for discharges with a stainless steel poloidal limiter and decreases to about 0.2 at high densities ( 6 × 10 13 cm −3 line average) for diverted discharges. These results are compared with estimates of the fueling efficiency which include molecular disintegration, plasma albedo for neutral atoms and imperfect wall reflection.

A review is given on investigations in the ASDEX Tokamak on impurities in ohmically, NI and ICRH heated plasmas. For ohmic discharges in H2 and D2 it is found that iron release from the wall can be explained by sputtering due to neutral charge exchange (CX) atoms. In the case of He, however, significant contributions caused by ion sputtering are inferred. Comparing discharges with C limiters in He and D2 suggests that in the case of hydrogen chemical processes are involved in C sputtering. By means of wall carbonization the concentrations of metal ions in the plasma could be substantially reduced. This achievement is of particular importance for NI counter-injection and ICRH, where under non-carbonized conditions severe impurity problems occur. We studied impurity confinement in the case of various heating scenarios by means of the laser injection technique. The poorest confinement is found for the L-phase of NI. Metal injection into the high confinement H-phase generally causes temporary suppression of the edge localized modes (ELM's). With respect to ICRH we conclude that enhanced wall erosion — probably due to the production of high energy ions in the boundary — together with a slightly increased impurity confinement is the dominant reason for the increase of the metallic concentrations. Impurity sputtering as an alternative strong erosion process was experimentally ruled out.

Local transport in the L (low βp) and H (high βp) types of double-null divertor discharges in ASDEX is investigated by computer modelling. For the L type, two subregimes with different local transport behaviour can be distinguished: Homogeneous electron heat diffusivities χe = e or flat χe profiles and diffusion coefficients D = 0.2 χe are found for e < 5.2 x 104 cm2s–1 (L-I), while above this limit, steep χe and flat D profiles are determined (L-II). The observed decrease in energy confinement time with increasing neutral injection power is found to be due to enhancement of the electron heat diffusivity. In the L-II regime, χe strongly increases with radius, while the diffusion coefficient remains flat. The diffusivities show a rather weak dependence on time as in the L-I regime. In both L regimes they rise with increasing injection power and decreasing plasma current, but are independent of density. In the H regime, transport is characterized by flat χe profiles and D = 0.2 χe. The diffusivities are reduced by a factor of about two in the whole plasma compared with the L type. They show the variation with plasma current found in L discharges and are also independent of density. One more common feature of L and H regimes is that the ion heat diffusivity is in the range of 1 to 3 times the neoclassical values.

A series of initial resonant helical divertor (RHD) experiments have been carried out in ohmically and auxiliary heated JIPP T-IIU plasmas. Disruptive and MHD instabilities make the interpretation of the RHD results difficult but an apparent increase in the energy confinement time is observed when the helical magnetic perturbation is applied. This may be due to the suppression of MHD activity or to a reduction in the edge convective heat losses. Magnetic island effects have been observed on the floating potential of a Langmuir probe array and energy scrape-off layer widths have been measured with and without helical perturbations during ICRF operation. Basic pump limiter data is presented including ion temperatures and C4+ impurity profiles. Energy confinement times are reported in ohmically and NBI heated discharges.

Systematic investigations of the scrape-off layer (SOL) in the midplane of ASDEX have been carried out in He, D2 and H2 for diverted ohmic discharges over a wide range of plasma conditions: ne ∼ 0.5−4.7 × 1013 cm−3, Ip = 200–450 kA, BT = 16–23 kG, qa∼ 2.4–4.4 and POH = 200–480 kW. For the first two cm outside the separatix, ne is found to decay exponentially with an e-folding length λn given by λn = kqα (He, k = 1.32 cm, α = 0.52; D2, k =1.29 cm, α = 0.35; H2, k = 1.18 cm, α = 0.4) when from which follows for qa = 3: λn(D2) ∼ λn(H2) ∼ 0.8 λn(He). The qα scaling is roughly predicted by the simple formula λn = D⊥ Lυ∥ under the assumption D⊥ ∝ mi−0.5 (as has been observed on ASDEX for H2 and D2). There appears to be no explicit λn dependence on heating power. λn varies strongly with ne in the range ne ≤ 1 × 1013 cm−3, decreasing for example (D2,H2; qa = 3.0), from λn ≥ 3 cm at ne ∼ 0.5 × 1013 cm−3 to λn ∼ 1.9 cm for ne ≥ 1.5 × 1013 cm−3, ne at the separatrix is primarily a function of ne.

In the first experimental phase the divertor tokamak ASDEX was run with a closed SS-limiter. The cleaning procedure for the vessel consisted of baking to 120°C, carbon removal by glow discharge in hydrogen, and oxygen removal with a continuous low power 50 Hz AC discharge. Wall contact of the plasma was reduced by carefully positioning the plasma with a feedback system. Discharges with plasma currents up to 280 kA and a disruption-free duration of up to 1 s were reliably produced with a filling pressure of 5 × 10 −5 mbar and a programmed current rise. No change in start-up conditions and discharge behaviour was observed in material limiter discharges with superimposed divertor field.

This letter reports on two experiments undertaken to evaluate the retention of gaseous and target produced impurities in the ASDEX divertor. The retention for gaseous impurities was determined by puffing Ar into the main chamber and simulating the time behaviour of the Ar XVI line intensity with a time dependent impurity transport code including a simple divertor model. During Ohmic heating a factor of 3 and 4.5 enhancement of impurity retention if found relative to the vacuum time constant (90 ms) of the divertor chamber, for ne = 2 × 1013 cm−3 and ne = 3.5 × 1013 cm−3, respectively, while a drastic breakdown of the retention occurs during high power NI heating. – To deduce the retention of impurities generated at the divertor plates, a segment (3.5%) of the plates was covered with copper, a metal previously not used in ASDEX. By measuring the Cu influx at the target plates and the line intensity of the Cu XX line (11.38 A) in the core plasma and by using the transport code, it is found that during NI heating (ne ≤ 2 × 1013 cm−3) Cu atoms originating from the target plates have a ≤ 3.5 times higher probability to penetrate into the core plasma than if they had when originating from the main chamber walls.

Recent measurements of plasma boundary and divertor scrape-off parameters for ohmically and neutral injection heated plasmas are presented. For these data the power flow onto the divertor plates and the sputtering rates at the plates are calculated and compared with separate measurements. The impurity behaviour in front of the plates is also discussed.