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Edge localised modes (ELMs) are a serious concern for plasma facing components in next generation tokamaks. Due to the limited capabilities of the commonly used bolometers reliable radiation data are generally only available with a time resolution of a few milliseconds, which is much larger than the power deposition time of type-I ELMs. Therefore an improved evaluation algorithm for bolometric data has been developed which allows to calculate the radiated energy on type-I ELM relevant timescales and to tomographically reconstruct the spatial distribution of the radiated energy during an ELM. This method has been used to investigate radiation during type-I ELMs in various plasma configurations. The spatial distribution of the radiated energy during an ELM shows that radiation losses are mainly located in the divertor. Up to 40% of the ELM energy loss are found as radiation. The energy balance is nearly fulfilled in most cases.

Bolometers are radiation sensors designed to have a spectral response as constant as possible in the region of interest. In high-temperature plasmas, the main radiation output is in the ultraviolet and SXR part of the spectrum and the metal foil bolometers are special detectors developed for this interval. For such sensors, as in general for all bolometers, the absolute calibration is a crucial issue. This problem becomes particularly severe when, like in nuclear fusion, the sensors are not easily accessible. In this article, a detailed description of the in situ calibration methods for the bolometer sensitivity S and the cooling time τc, the two essential parameters characterizing the behavior of the sensor, is provided and an estimate of the uncertainties for both constants is presented. The sensitivity S is determined via an electrical calibration, in which the effect of the cables connecting the bolometers to the powering circuitry is taken into account leading to an effective estimate for S. Experime...

An on-line predictor of the time to disruption has been installed on the ASDEX Upgrade tokamak. It is suitable either for avoidance of disruptions or for mitigation of those that are unavoidable. The prediction uses a neural network trained on eight plasma parameters and their time derivatives extracted from 99 disruptive discharges. The network was tested off-line over 500 discharges and was found to work reliably and to be able to predict the majority of the disruptions. The trained network was installed on-line, tested over 128 discharges and used to inject killer pellets to mitigate the disruption loads.

Radiation distribution and heat load on the target plates of ASDEX Upgrade are investigated over a wide experimental parameter range with discharges up to 20 MW heating power and n e =(0.4–0.8)n GW . Integrating the radiation density over the whole plasma yields a total radiated power of about 80% of the input power, and radiation from the divertor and X-point region is about 40% of the input power. Whereas the radiated power in the main plasma is comparable for the former open divertor I (Div I) and the new closed LYRA divertor (Div II), the radiation fraction in the divertor has nearly doubled for the LYRA divertor. Simultaneous to the increased divertor radiation in the LYRA divertor, the maximum heat flux density to the target plates, measured by thermographic cameras, is reduced by a factor of about two compared to that in Div I for comparable attached conditions. Furthermore, especially in the inner divertor fan, the radiation contributes to a large part of the measured heat flux.

A simpler and faster method for determining the total radiated power emitted from a tokamak plasma in real-time has been developed. This quantity is normally calculated after the discharge by a deconvolution of line integrals from a bolometer camera. This time-consuming algorithm assumes constant emissivity on closed flux surfaces and therefore needs the exact magnetic equilibrium information. Thus, it is highly desirable to have a different, simpler way to determine the total radiated power in real-time without additional magnetic equilibrium information. The real-time calculation of the total radiated power is done by a summation over ten or 18 lines of sight selected out of a bolometer camera with 40 channels. The number of channels is restricted by the summation hardware. A new selection scheme, which uses a singular value decomposition, has been developed to select the required subset of line integrals from the camera. With this subset, a linear regression analysis was done against the radiated power...

A miniaturized low-noise metal-film bolometer detector array capable of recording radiation spanning a wide spectral range is presented. The bolometers consist of structured gold layers coated with a black nickel layer to reduce the reflectivity in the long-wavelength range. Investigation of the device`s photoresponse showed sufficient sensitivity at wavelengths lambda =4 mm to 6 AA. The detector unit meets high-vacuum requirements, is bakeable up to 300 degrees C, can withstand high nuclear radiation doses, is insensitive to thermal and electromagnetic interferences and can be operated in high magnetic fields. The cooling time of the bolometer is determined by its thermal properties, being about 70 ms in air. The device has meanwhile been successfully used in a few fusion plasma experiments.

Puffing of impurities (neon, argon) and deuterium gas in the main chamber is used to feedback control the total radiated power fraction and the divertor neutral particle density simultaneously in the ASDEX Upgrade tokamak. The variation of Psep=Pheat-Prad(core) by impurity radiation during H mode shows a similar effect on the ELM behaviour as that obtained by a change of the heating power. For radiated power fractions above 90%, the ELM amplitude becomes very small and detachment from the divertor plates occurs, whilst no degradation of the global energy confinement is observed (completely detached high confinement mode). Additional deuterium gas puffing is found to increase the radiated power per impurity ion in the plasma core owing to the combined effect of a higher particle recycling rate and a lower core penetration probability. The outer divertor chamber, which is closed for deuterium neutrals, builds up a high neutral pressure, the magnitude of which is determined by the balance of particle sources and pumping. For this particular situation, the effective pumping time of neon and argon is considerably reduced, to less than 0.3 s, mainly owing to an improved divertor retention capability. The radiation characteristics of discharges with a neon driven radiative mantle are modelled using a 1-D radial impurity transport code that has been coupled to a simple divertor model describing particle recycling and pumping. The results of simulations are in good agreement with experiment

We report on an investigation of a photoresponse induced in a novel, miniaturized low-noise metal resistor film bolometer when irradiated by ultrashort laser pulses (duration 100 ps to 200 ns) with wavelengths covering the mid- and far-infrared spectral ranges. Successful operation at room temperature was achieved and almost constant detectivity was found over the whole frequency range investigated (15–1100 cm−1). The minimum detectable laser pulse energy was determined to be about 0.6 μ J (with an electronic bandwidth of 1 kHz). This high detectivity and the miniaturized dimensions afford the possibility of building detector arrays applicable in mid- and far-infrared spectroscopy.

Maximum radiation loss without disruption is essential to extend the maintenance intervals of ITER and following fusion reactors. The strong radiation distribution at the divertor must be known in fine detail to control such discharges. The corresponding reference diagnostic uses metal resistor bolometer cameras for detailed tomographic reconstructions of the radiation distribution2. The radiation hardness of the detector had been identified as the main question concerning the reference design. We report on the results of a Japanese/European collaboration (under ITER task T492) achieving the first in situ neutron irradiation tests of present day bolometers. A neutron dose equivalent to 0.1 dpa was reached, corresponding to the fluence between two maintenance periods at the bolometer locations behind the first wall elements or in the divertor under full load conditions of ITER-FEAT.

A general qualitative and quantitative analysis of the experimental and numerical errors which affect the plasma total radiated power measurements performed in thermonuclear fusion experiments with widely used miniaturized low-noise metal resistor bolometers is presented in this paper. This analysis, which generalizes the results of previous work on the detection limit, is motivated by the fact that, by means of bolometric tomography, detailed reconstruction of the radiated power emissivity distributions, in particular during fast phenomena, requires a high level of precision in the raw measurement and then a thorough control of the error sources. Since these bolometers are widely used in the largest tokamak, RFP and Stellarator experiments and are also considered for the reference design of the bolometric diagnostics in the International Thermonuclear Experimental Reactor (ITER), the study presented here has been accomplished in a machine-independent way. Practical formulae and methods to estimate the experimental errors that do not depend on the magnetic configuration are presented.

Bolometers are indispensable diagnostic instruments for ITER. They allow to control the radiated power fraction and the position of the radiation maximum in the divertor, both key issues for the minimisation of erosion and achievement of high performance. To fulfil the ITER requirement of a reasonably good tomographic reproduction of the divertor radiation distribution one needs not only a dense coverage with lines of sight in the divertor but also a substantial coverage of the main plasma. To achieve this, cameras will have to be placed into a harsh environment. This is, where JET’s experience with the same approach is very useful. The JET detector based on a proprietary [1, 2], compact 4-channel bolometer which has been modified so that active cooling is not required [3, 4, 5, 6] leads the way towards the solution of these problems. Drawing from JET’s experience with this new system a concept for ITER is outlined with the intention to take all relevant aspects into account.

A new multichord bolometer array has been installed for the first time in the RFX experiment, a large machine designed to study plasma confinement at high current in the reversed field pinch configuration. The main purpose of the diagnostic is to provide emissivity spatial profiles with good time resolution for the measurement of the total radiated power. The detectors used are miniaturized metal foil bolometers which are mounted in a water cooled pinhole camera covered by a graphite shield. The bolometers can be absolutely calibrated in situ according to a technique originally developed for the Asdex tokamak. In order to prove the electrical calibration technique a blackened bolometer was optically irradiated with a monochromatic light over a wide spectral range with wave numbers from 2.7 up to 47 000 cm-1. In the visible and UV spectral range an absorptivity of 80% was measured while in the IR and far-infrared regions only 7% of the incoming light was absorbed by the bolometer foil. In the mm wavelength range less than 1% of the radiation was absorbed.

Bolometric measurements are performed in the RFX reversed field pinch experiment by means of a low noise miniaturized multichannel metal resistor bolometer module. The first tests on RFX, an experiment with geometrical dimensions comparable with those of large tokamaks but characterized by a much faster plasma dynamics, show that this thin film bolometer can be operated with high bandwidth and high time resolution in very different fusion experiments. A new data analysis technique, which allows the derivation of the power absorbed by the bolometer at a high time resolution (Δt=3×10−5 s) for the study of fast phenomena is presented.

Tokamak discharges with repetitive pellet fuelling were investigated in the ASDEX divertor device. The importance of sufficiently high divertor recycling for a high density at the separatrix and for successful density buildup in the bulk plasma was demonstrated. In contrast to low recycling discharges where no permanent improvement of the energy confinement time was achieved, in OH-heated discharges with high recycling an energy confinement time of 160 ms was reached, the normal value being 80 ms in the rollover region. The peaked density profiles in this case were accompanied by reduced or suppressed sawtooth activity and finally ended in a phase of strong central impurity accumulation. The particle transport was characterized by strong, non-classical inward drift, while the improved energy transport can be explained by the following alternatives: (la) a local model which assumes neo-Alcator χe for the electrons and χi= 3χneocl for the ions in the gas puff cases, reducing to χi= χneocl for the optimum pellet cases; (1b) the assumption χi= χneocl under all conditions and an electron energy confinement worse than neo-Alcator in the rollover region in gas-puff-discharges; (2) a profile consistency picture where Te(a) determines the energy confinement. Low power, NI heated discharges with pellet fuelling behave like Ohmic discharges, while for high power in the L-mode no successful density buildup was possible, and τE was not improved. The H-regime was extended from a density maximum e = 0.8 × 1020 m-3 without pellets to e = 1.2 × 1020 m-3 by the injection of pellets. In this case a density buildup takes place, but further density profile peaking could not be observed.

In JET, both high density and low-q operation are limited by disruptions. The density limit disruptions are caused initially by impurity radiation. This causes a contraction of the plasma temperature profile and leads to an MHD unstable configuration. There is evidence of magnetic island formation resulting in minor disruptions. After several minor disruptions, a major disruption with a rapid energy quench occurs. This event takes place in two stages. In the first stage there is a loss of energy from the central region. In the second stage there is a more rapid drop to a very low temperature, apparently due to a dramatic increase in impurity radiation. The final current decay takes place in the resulting cold plasma. During the growth of the MHD instability the initially rotating mode is brought to rest. This mode locking is believed to be due to an electromagnetic interaction with the vacuum vessel and external magnetic field asymmetries. The low-q disruptions are remarkable for the precision with which they occur at qψ = 2. These disruptions do not have extended precursors or minor disruptions. The instability grows and locks rapidly. The energy quench and current decay are generally similar to those of the density limit.

Radiation losses from the JET plasma are measured (temporally and spatially resolved) with three pin‐hole cameras which observe one poloidal plasma cross section from underneath (one vertical camera with 14 bolometers) and from the side (two horizontal cameras with 10 bolometers each). In addition, eight collimated single bolometers monitor toroidal symmetry of plasma emissivity. An improved metal resistor bolometer capable of surviving at least 5000 D–T pulses in JET is composed of two identical foils (4‐μm‐thick gold absorber, 7.5‐μm‐thick Kapton substrate, and 0.1‐μm‐thick gold resistor), mounted one behind another in one housing. The rear foil serves as a reference bolometer in a bridge circuit. With an integration time constant of 20 ms a detection limit of 70 μW/cm2 was measured in JET. Examples of measuring results in JET are presented: strong poloidal asymmetries of the radiation density as precursor signals of a density limit disruption and the formation of a poloidally symmetric, cold plasma man...

Densities achievable in ASDEX Upgrade discharges are restricted by a disruptive limit in the L-mode caused by an edge-power imbalance which is linking divertor detachment, Marfe formation and the separatrix density. The attainable average densities depend then on the internal particle sources and the core transport and can exceed the empirical Greenwald density. In H-mode an upper density limit is found which represents a non-disruptive H - L back transition, which is preceded by the occurrence of type-III ELMs. Close to the Greenwald limit this H - L transition cannot be avoided at any power flux across the separatrix and - at high external neutral gas fluxes - confinement compared with ITER H-92P scaling degrades even before the back transition. The H-mode operational window is determined by local edge-barrier parameters and their gradients, respectively. The boundaries are represented by the L - H transition-temperature threshold, the ideal ballooning edge-pressure gradient limit, the upper temperature limit for type-III ELMs and an upper H-mode barrier density limitation. The cause for the last limitation is not yet identified; it may be due to resistive ballooning modes or the separatrix density limit. Despite the limited edge densities the Greenwald density could be surpassed by a factor of three with pellet refuelling from the low magnetic-field side. Pellet injection from the high-field side gains from the strong increase of fuelling efficiency due to the assisting toroidal outward drift of the formed high- ablatant. Higher densities are achievable in H-mode compared with low-field side injection and diminished convective losses avoid confinement degradation up to the Greenwald density. In gas-puffed type-I ELMy H-modes the plasma thermal energy and the edge-pressure gradients, which are limited by ballooning stability, are linked via a robust temperature-profile stiffness and the flat density profiles resulting from dominant edge refuelling at high densities. Their confinement does not improve with increasing density (and neutral gas fluxes) and may even slightly degrade. Therefore, the superior confinement of type-I ELMy H-modes compared with type-III ELMy ones at medium densities is actually offset at densities close to the Greenwald density. In contrast to the temperature-profile resilience density profiles can be changed both by deep refuelling (with pellets) and intrinsic transport improvements connected with density peaking (observed in CDH-modes), which offers the combination of high confinement and high density operation. The possible alliance with radiation cooling, divertor detachment and divertor compatible type-III ELMs could solve the power exhaust problem.

Recent experiments at ASDEX Upgrade have achieved advanced scenarios with high β N (>3) and confinement enhancement over ITER98(y, 2) scaling, H H98y2 = 1.1-1.5, in steady state. These discharges have been obtained in a modified divertor configuration for ASDEX Upgrade, allowing operation at higher triangularity, and with a changed neutral beam injection (NBI) system, for a more tangential, off-axis beam deposition. The figure of merit, β N H ITER89-P , reaches up to 7.5 for several seconds in plasmas approaching stationary conditions. These advanced tokamak discharges have low magnetic shear in the centre, with q on-axis near 1, and edge safety factor, q 95 in the range 3.3-4.5. This q-profile is sustained by the bootstrap current, NBI-driven current and fishbone activity in the core. The off-axis heating leads to a strong peaking of the density profile and impurity accumulation in the core. This can be avoided by adding some central heating from ion cyclotron resonance heating or electron cyclotron resonance heating, since the temperature profiles are stiff in this advanced scenario (no internal transport barrier). Using a combination of NBI and gas fuelling line, average densities up to 80-90% of the Greenwald density are achieved, maintaining good confinement. The best integrated results in terms of confinement, stability and ability to operate at high density are obtained in highly shaped configurations, near double null, with δ = 0.43. At the highest densities, a strong reduction of the edge localized mode activity similar to type II activity is observed, providing a steady power load on the divertor, in the range of 6 MW m -2 , despite the high input power used (> 10 MW).